Page 1
Page 2
Page 3
Page 4
Page 5
Page 6
Page 7
Page 8
Page 9
Page 10
Page 11
Page 12
Page 13
Page 14
Page 15
Page 16
Page 17
Page 18
Page 19
Page 20
Page 21
Page 22
Page 23
Page 24
Page 25
Page 26
Page 27
Page 28
Page 29
Page 30
Page 31
Page 32
Page 33
Page 34
Page 35
Page 36
Page 37
Page 38
Page 39
Page 40
Page 41
Page 42
Page 43
Page 44
Page 45
Page 46
Page 47
Page 48
Page 49
Page 50
Page 51
Page 52
Page 53
Page 54
Page 55
Page 56
Page 57
Page 58
Page 59
Page 60
Page 61
Page 62
Page 63
Page 64
Page 65
Page 66
Page 67
Page 68
Page 69
Page 70
Page 71
Page 72
Page 73
Page 74
Page 75
Page 76
Page 77
Page 78
Page 79
Page 80
Page 81
Page 82
Page 83
Page 84
Page 85
Page 86
Page 87
Page 88
Page 89
Page 90
Page 91
Page 92
Page 93
Page 94
Page 95
Page 96
Page 97
Page 98
Page 99
Page 100
Page 101
Page 102
Page 103
Page 104
Page 105
Page 106
Page 107
Page 108
Page 109
Page 110
Page 111
Page 112
Page 113
Page 114
Page 115
Page 116
Page 117
Page 118
Page 119
Page 120
Page 121
Page 122
Page 123
Page 124
Page 125
Page 126
Page 127
Page 128
Page 129
Page 130
Page 131
Page 132
Page 133
Page 134
Page 135
Page 136
Page 137
Page 138
Page 139
Page 140
Page 141
Page 142
Page 143
Page 144
Page 145
Page 146
Page 147
Page 148
Page 149
Page 150
Page 151
Page 152
Nuclear Science User Facilities 2 Nuclear Science User Facilities 995 University Boulevard Idaho Falls ID 83401-3553 www.nsuf.inl.gov Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof nor any of their employees makes any warranty expressed or implied or assumes any legal liability or responsibility for the accuracy completeness or usefulness of any information apparatus product or process disclosed or represents that its use would not infringe privately owned rights. References herein to any specific commercial product process or service by trade name trade mark manufacturer or otherwise does not necessarily constitute or imply its endorsement recommendation or favoring by the U.S. Government or any agency thereof.The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof. INLEXT 14-2057-R1 Prepared for the U.S. Department of Energy Office of Nuclear Energy under DOE Idaho Operations Office Contract DE-AC07-051D14517. This report covers the period beginning October 12013through September 302014 2014 ANNUAL REPORT Jeff Benson Program Administrator 208 526-3841 jeff.bensoninl.gov Dan Ogden Deputy Director 208 526-4400 dan.ogdeninl.gov John Jackson Industry Program Lead 208 526-0293 john.jacksoninl.gov Collin Knight Project Manager Post-irradiation Examination 208 533-7707 collin.knightinl.gov Sarah Robertson Communications Liaison 208 526-0490 sarah.robertsoninl.gov Brenden Heidrich Capabilities Coordinator 208 533-8210 brenden.heidrichinl.gov Renae Soelberg Administrative Assistant 208 526-6918 renae.soelberginl.gov Sebastien Teysseyre Research Scientist 208 526-8263 sebastien.teysseyreinl.gov Jim Cole Chief Scientist 208 526-8101 james.coleinl.gov J. Rory Kennedy Director 208 526-5522 rory.kennedyinl.gov OUR NSUF TEAM Nuclear Science User Facilities 4 2014 ANNUAL REPORT Nuclear Science User Facilities From the NSUF Director..................................................................................................................... 7 Researcher Profiles............................................................................................................................. 9 A Passion for Diversity - PNNL......................................................................................................... 16 Scientific Review Board.................................................................................................................... 20 Interview with the Director.............................................................................................................. 24 NSUF Overview Program Overview........................................................................................................................... 31 Reactor capabilities........................................................................................................................... 32 PIE capabilities.................................................................................................................................. 34 Beamline capabilities........................................................................................................................ 38 Calls for Proposals............................................................................................................................ 40 Users Meeting.................................................................................................................................. 42 Distributive Partnerships Map Distributed Partnerships at a Glance................................................................................................ 44 NSUF Projects Awarded reports.............................................................................................................................. 48 Industry Program reports.............................................................................................................. 136 Resources Acronyms...................................................................................................................................... 143 Index............................................................................................................................................. 148 TABLE OF CONTENTS Nuclear Science User Facilities 6 2014 ANNUAL REPORT 7 FROMTHE NSUF DIRECTOR J. Rory Kennedy Director 208 526-5522 rory.kennedyinl.gov This marks my first annual report letter to you as director of the AdvancedTest Reactor National Scientific User Facility. It is also the last letter I will write to you as director of the AdvancedTest Reactor National Scientific User Facility ATR NSUF. Later this year the official name of the ATR NSUF will change to Nuclear Science User Facilities.This new designation more accurately reflects just how much the entire public-sector nuclear research effort in the United States has grown and matured.While our former name has served us well since the user facility was established in 2007 it referred to only the test reactor and its associated facilities at Idaho National Laboratory INL. Our new designation reflects what we are today a network of facilities throughout the United States that are available to researchers for the study and advancement of nuclear science and engineering. Today our 11 partner facilities including the Westinghouse Materials Center of Excellence our first member from the private sectorprovide their cutting-edge equipment and scientific expertise to researchers seeking to develop nuclear energy as a safe and reliable resource. the partner program to increase utilization efficiencies.We envision the sample library and capabilities database will become invaluable references for both NE and researchers in the future. These are all exciting endeavors intended to better support our users and advance nuclear research not only for the United States but the entire world.And while I will help lead the effort to fulfill them going forward the foundation for their implementa- tion has been laid by my predeces- sors Dr. Mitch Meyer Professor Todd Allen Professor JeffTerry and Frances Marshall who served as interim director during my transition.Their innovative thinking hard work and dedication to the job and to the ATR NSUF have enhanced the organization the partnership program and indeed the entire nuclear industry further than we could have anticipated. I am grateful to them and hope I can continue the impressive legacy they have established. Sincerely Dr. J. Rory Kennedy Director As exciting as our name change is and what it represents there are a number of other tasks the ATR NSUF has undertaken or will undertake either under our own initiative or at the behest of the U.S. Department of Energy DOE.They include project- forward funding in which an entire research project is funded at the time it is first approved consolidating our calls for proposals into the Consolidated Innovative Nuclear Research CINR Funding Opportunity Announcement FOA which includes those of the Nuclear Energy University Programs NEUP and Nuclear Energy Enabling Technologies NEET to give users the opportunity to apply for RD and NSUF no-cost access in one application opening project principal investiga- tors to others besides just university researchers including those from industry national laboratories and small businesses increasing our partner facilities interaction and promotion by simulcasting the Users Meeting to and from all the partner facilities as well as updating our conference and meeting exhibits continuing to enhance the sample library of irradiated materials and establishing a searchable database of all Nuclear Energy NE-accessible capabilities available to users through Nuclear Science User Facilities 8 2014 ANNUAL REPORT 9 RESEARCHER PROFILES After graduating from high school in her native India in 2006 Mahima Gupta moved to the U.S. to attend the University of Michigan with every intention of studying genetic engineering. But even the best plans often dont go as expected. I was always interested in physics says Guptaso I was taking an honors class. My teacher was this crazy German guy very intelligent and his class just blew my mind. She spent the following summer working with that professor on one of the first programs in the U.S. investi- gating Bose-Einstein condensates.When she resumed her undergraduate studies in the fall her advisor suggested she look into nuclear engineering. It was instant love Gupta says.I loved that you could use nuclear power to create carbon-free electricity. It was clean renewablethe whole deal. During her undergraduate time at Michigan she served as president of the localAmerican Nuclear Society chapter. At one of the societys conferences she met Dr.ToddAllen. She mentioned she wasnt sure where she wanted to go to graduate school and he asked what she wanted to study.When she said fuels he suggested she come to the University of Wisconsin Madison UW and work with him. She agreed and received her Ph.D. inApril 2015 after spending the last two years finishing her thesis at INL. I was finished with all my classes in the summer of 2013 says Gupta so I moved to Idaho. I worked at the Materials and Fuels Complex MFC for a year and then started an internship with ATR NSUF. Its been a great experience. She wasted no time taking advantage of the NSUFs proposal process.With two proposals already accepted the first of which is included in this annual report and a third submitted she is already making her mark as an up-and-coming researcher.The initial proposal a first-of-its-kind experi- ment examined the effects of irradia- tion on uranium-dioxide UO2 at the atomic level. Most studies on irradiation damage in UO2 are usually done on a larger scale explains Gupta. Were looking at angstrom and sub-angstrom scales so the physics is totally different. One of my goals is to see if there is any relationship between these really tiny scales and slightly larger scales. She spent the first two years of the project irradiating her samples in the Ion Beam Laboratory at UW one of the NSUFs partner institutions. It was one of the first ion accelerator irradiations completed on UO2 at the UW lab. She also helped write the protocols for those irradiations and as a result UW now has a prepara- tion facility that can handle irradiated Mahima Gupta Nuclear Science User Facilities 10 samples for future research. Next the samples were prepared and examined at CAES using a focused ion beam FIB and transmission electron microscopy TEM. One thing were seeing is that oxygen defect clustering increased with ion beam irradiation on a larger scale that is seen inTEM micrographs says Gupta. This leads us to believe that there is a whole different class of defect clusters larger than individual point defects but smaller than what can be detected in theTEM. To explore that possibility a systematic series of experiments was conducted utilizing the Stanford Synchrotron Radiation Laboratory SSRL. Samples were studied using X-ray absorption fine-structure spectroscopy XAFS to see how the atomic structure changes in UO2 after irradiation.After creating minimally radioactive samples using ion accelerators Gupta utilized a technique called micro-focusing XAFS. It was the first time the process was used on irradiated specimens. Using the instruments at CAES she and her colleagues picked out samples about 10 x 15 microns in size and sent them to SSRL. By micro-focusing the X-ray beam another first for nuclear fuel samples they were able to raster a micron-sized beam across the samples to find the UO2 signals and perform the XAFS analysis. Gupta is quick to point out the advantages of these rapid turnaround experiments RTEs. I submitted the proposal right when I started my internship at INL she says. It was approved a couple of months later and we were able to start work instantly.You get results right away with state-of-the-art instru- ments and with the full support of the CAES staff. Its a great way to work on your Ph.D. Having finished work on the second proposal Gupta is poised to start the third phase of her project but now that shes graduated a job search has taken priority. Im trying to figure out how to make the transition. I can still continue my work here but I feel like if there were other Ph.D. students who were interested someone could write their entire thesis on these samples. You get results right away with state-of-the-art instrumentsand with the full support of the CAES staff.Its a great way to work on your Ph.D. 2014 ANNUAL REPORT 11 After five years as an early career researcher at the South Africa Nuclear Energy Corporation Dr. Isabella van Rooyen went to work on her countrys pebble bed modular reactor program.When the govern- ment withdrew funding from that program in 2011 she decided the U.S. offered the best opportunities for her to pursue her passion for studying nuclear materials particularly tristructural isotropic TRISO-coated fuel particles. Partly because of its involvement with the Next Generation Nuclear Plant NGNP program and partly because of her broad respected background in the nuclear field INL welcomed her with open arms. I had also gained a lot of experience working in the industrial sector van Rooyen says. That turned out to be a big advantage because now I not only think about what Im working on at that moment I think about what the application of that product will be for the end user. When she first came to INL she split her time between studying the viability ofTRISO-coated particles and working on the light water reactor LWR sustainability program. Her success withTRISO fuels soon won out and this year she and her colleagues Dr.Thomas Lillo and Dr.Yaqiao Wu received an INL Outstanding Paper award for Identification of Silver and Palladium in IrradiatedTRISO-Coated Particles which was published in the Journal of Nuclear Materials in 2014. I had started trying to solve the silver transport mechanism problem in South Africa says van Rooyen and I learned a lot there but I could never go to the next stage because we couldnt study irradiated particles. That hurdle was overcome at INL and resulted in a breakthrough discovery. We always knew from gamma spectrometry that under irradiation silver is released outside these tiny silicon-carbide-coated particles van Rooyen explains but we never knew where and how. Now that we know the path it follows we can work toward a solution to prevent that release from happening. While they have not yet tracked the full mechanism they feel very close to gaining a new set of knowledge. Using transmission electron microcopy we are currently looking systematically at the orientation and grain boundary characteristics of each grain that is adjacent to the silver we have found says van Rooyen. Once we find the grain characteristics we can manipulate the release process in such a way that does not favor the driving transport mechanisms. Isabella van Rooyen Nuclear Science User Facilities 12 The question of silver transport has been a nagging one for the Nuclear Regulatory Commission because of its impact on the licensing of high- temperature gas reactors. Being able to change the microstructure of the silicon-carbide layer will be a huge step forward.Van Rooyens study has also provided valuable information on a question that researchers in the LWR accident-tolerant fuel program have been longing to answer in their investigation ofTRISO-coated particles as a potential fuel for that program. Van Rooyen is quick to point out that their discovery was made possible by the excellent microscopes avail- able at Center for Advanced Energy Studies CAES as well as the excellent scientists who operate them. This project was really innovative on the part of our whole team van Rooyen said. Many people said we couldnt accomplish it because of the high radioactivity in the parent mate- rial but we persisted. It shows other researchers that they must not stop. It is no longer an excuse for nuclear material scientists to say they dont have the availability of equipment. Because of the brilliant partnership program at ATR NSUF we can all use the newest stuff. Already scientists as far away as the United Kingdom France China and South Korea have begun using the nanostructural and microstructural information on silicon-carbide provided by this breakthrough study to pursue further research. Every time I present my work at a conference I realize the impact the NSUF has had in my role as a scientist here at INL said van Rooyen. One of my presentations at Users Week two years ago stimulated collaboration with a university and I became a part of their research team. Helping us to build our networks is one of the missions of the NSUF and I can truly say I am living proof that it is working. Those networks offer benefits beyond any one particular project. One of the students that worked with us as an intern on this latest project is now applying for a perma- nent position at INL van Rooyen said says with satisfaction. So by funding this research the NSUF has not only helped me along with other researchers and INL it is also helping to build the next generation of researchers. I hope we can continue that as we move forward. So by funding this researchthe NSUF has not only helped mealong with other researchers and INLit is also helping to build the next generation of researchers. I hope we can continue that as we move forward. 2014 ANNUAL REPORT 13 Nuclear Science User Facilities 14 Every student has that fork-in-the- road moment that defines the rest of their life and more often than not its a good teacher who provides it. Boise State University Assistant Professor JanelleWharry always wanted to be an engineer she just didnt know what kind.The summer after her junior year in high school she left home in Hawaii to attend a two-week program at the University of Illinois and all that changed. It was like a camp saidWharry They had kids from all over and we all got to learn a little bit about the different engineering majors the university offered.They had a nuclear engineering program that really got me interested in the medical applications especially cancer research. She ended up going to the University of Michigan UM where her focus shifted to nuclear energy and finished her masters degree in 2005. She didnt really know where she wanted to go from there so she took a job with Duke Energy in North Carolina for two years as a core design engineer. What I learned there was that the neutronics of a nuclear system are reli- able and very well understood but the main issues we were always having with the plants were the effects of radiation on structural materials. So I decided thats where I would focus my thesis. She went back to UM and received her Ph.D. in 2012.While doing a six-month postdoctoral study there she started looking for a job. She knew academia was where she wanted to be and after all the moving around she had done over the years she just wanted to settle down somewhere and make a life for herself.When Boise State made her an offer it seemed like the perfect fit. I love it hereWharry said.My background is mostly irradiation effects in materials but since I came here and started my own program Ive expanded into metallurgy and mechanical proper- ties.The main focus of my research group is to study how mechanical properties evolve in correlation with microstructure under irradiation. Materials science and engineering is a relatively new area of study at Boise State.The undergraduate program began barely 10 years ago and shortly after that with the help of Micron Technology a masters curriculum was added. In 2012 thanks to a 13 million grant from the Micron Foundation a Ph.D. program was instituted.The first year just over 20 students were seeking their doctorates and today there are more than twice that number along with 50 undergraduates and more than 40 masters students. Boise State is not yet an NSUF partner but along with the University of Idaho and Idaho State University it is a partner in the state-funded CAES. One of my long-term goals for BSU is to establish a partnership withATR NSUF as well Janelle says.Were in the process of buying some pretty cool equipment for in-situ mechanical testing experiments in the transmission electron microscope TEM that Im hoping will bring a unique capability to CAES. Hopefully that will be a major step toward that goal. Janelle Wharry Were in the process of buying some pretty cool equipment for in-situ mechanical testing experiments in the transmission electron microscope TEM that Im hoping will bring a unique capability to CAES.Hopefully that will be a major step toward that goal. 2014 ANNUAL REPORT 15 Adding a Hysitron PI-95 picoindenter and new thermal-aging capabilities to her BSU laboratory this year will be a big help in one of her main research efforts looking at the role solid-solution strengthening plays in the mechanical strength of advanced ferritic-martensitic steels and oxide- strengthened dispersion steels. My real interest is in trying to improve steels says Janelle not just in irradiation environments but in general. Can we harness properties that are desirable for radiation tolerance and use them to develop better steels in nonirradiation environments I think its important to broaden our horizons and look at what people are developing for non-nuclear applications as well and how we can use those advances in the nuclear environment. One National Science Foundation project shes currently working on is a prime example of that theory. Were looking at titanium-dioxide TiO2 as a potential lithium-ion battery anode material explainsWharry.Irra- diation creates a high concentration of defects in a material which we believe will provide more space for lithium ions to intercalate. So if we can harness the power of irradiation to create vacancies in theTiO2 we may be able to improve the functionality and efficiency of batteries. Irradiation is always talked about as damage. I guess its the optimist in me but I dont believe it can be 100 percent bad.There has to be something useful for everyday life that can come from irradiating materials. Nuclear Science User Facilities 16 Collaboration and teamwork are the keys to achieving success in almost any endeavor and in scientific research and development theyre critical.This spirit typifies research at INL and is in turn carried out to other nuclear research institu- tions through a distributed partnership program overseen by NSUF. Begun in 2009 the program now comprises 10 national laboratories and universi- ties along with one industry partner Westinghouse. One of the first to join this comprehensive national network of scientific research capabilities was Pacific Northwest National Laboratory PNNL in RichlandWashington. Since its inception PNNL has been operated by the Battelle Memorial Institute. It is one of five U.S. laboratories stewarded by DOEs Office of Science A PASSION FOR DIVERSITY PNNL but it serves all of DOEs missions science energy environment and national security.As such the passion of its scientists and researchers for innova- tive solutions to problems faced by the U.S. as well as other countries extends to the U.S. Department of Homeland Security the National Nuclear Security Administration and a host of other government agencies universities and industry customers. It was born in 1965 as the Pacific Northwest Laboratory PNL to conduct research and development for the nearby Hanford nuclear site.There its scientists and engineers designed and conducted research at the Fast Flux Test Facility testing fuels and materials for the Liquid Metal Fast Breeder Reactor program to expand the nations nuclear energy capabilities. Research 2014 ANNUAL REPORT 17 soon expanded into areas as diverse as holography and heart transplants cybersecurity and environmental studies. PNNL the second N for National was added in 1995 researchers pioneered compact disc technology invented the first portable blood irradiator for leukemia treatments and were chosen to help study lunar materials retrieved during NationalAeronautics and SpaceAdministrations NASAApollo program to name just a few from a long list of scientific accomplishments.Today more than 4300 scientists engineers and nontechnical staff are responsible for a legacy that includes 2342 patents 104 of which were awarded in 2014 alone and a total of 81 Federal Labora- tory Consortium awards for technology transfer since 1984. The lab is divided into four major technical directorates Energy and Environment National Security Fundamental and Computational Sciences and the Environmental Molecular Sciences Laboratory. Two major clients the Nuclear Regulatory Commission NRC and DOEs Office of Nuclear Energy NE of which NSUF is a part comprise the core of PNNLs work in nuclear energy while a third area nonconventional helps keep the U.S. engaged in commercial as well as international programs and gathers insight from those collaborations. We have a very broad portfolio of science and technology interests says Steve Unwin PNNLs Nuclear Market Sector Manager.Just looking within those three big blocks in nuclear power we tend to be very eclectic. For example were heavily involved in helping implement the new generation of risk- informed regulations and we were very proud to have led the environmental reviews on behalf of the NRC for the first two plants to receive construction permits in the U.S. in over 30 years in South Carolina and Georgia.Weve also been a leader in materials science and nondestructive examination a very important field given the aging fleet of reactors that must be kept vital to meet the nations energy needs. On the NE side the labs biggest effort goes toward fuel cycle research and development. In addition to a long history of expertise in spent fuel quali- fication and being involved in used fuel reprocessing research its scientists are also studying new fuels that are more accident-tolerant and more economi- cally attractive. NE is looking at the next generation of reactors Unwin saysand weve been developing technologies that will help make them safe and economically attrac- tive.Also small modular light water reactors are showing great promise as a viable solution to energy needs and we helped the NRC develop the regulatory structure to license them. Part of the value of all this expertise is the creative thinking that goes along with it and the experience accumulated through the years50 years to be exact. In 2015 PNNL celebrates its GoldenAnniversary a milestone that brings both the past and the future into sharper focus. I dont think we have fulfilled our potential in the nuclear arena yet says Unwin.Our vision is to contribute to that potential by ensuring the success of our clients missions as well as that of the nation and we want to bring all our ideas innovations and technical expertise to the table. Another area where PNNL scientists have particular depth is working towards a nuclear spent fuel repository. We have the right types of expertise to help streamline that process in areas such as waste form qualification and risk assessment Unwin says. We do a lot of work for example around ensuring the integrity of spent fuel systems from storage and transporta- tion to final disposition. Advanced reactor technology is another equally important part of the diversity of research that is a hallmark of PNNL. The beauty of some of these advanced reactor designs is that they actually generate their own fuel Unwin explainsso we can minimize both the consumption of fuel resources and the creation of waste.The data and expertise we gained from the operation of the Hanford Fast FluxTest Facility is allowing us to bring valuable insight to the new generation of reactors.All the analysis shows is that if were going to meet national energy needs it really has to be an all-of-the-above policy of which nuclear powerclean reliable and carbon-freeis a critical compo- nent.At the end of the day its going to be the commercial sector building these reactors so our work with industrial clients opens up a conduit for new technology developed by NE. Remaining at the forefront of all this cutting-edge research requires state-of- the-art equipment and that in turn requires investing heavily in the future of the nuclear industry. Nuclear Science User Facilities 18 I would say that stock in the nuclear portfolio is higher than it has been in recent years says Steve Schlahta director of the Nuclear Science Project Management Office at PNNL.Were committed to a significant internal investment leveraging our materials science specifically our spectroscopy capabilities which we feel are hallmarks in this area.We recently invested over 2 million in a JEOL aberration- corrected transmission electron microscope one of the few in the world.Were also putting a FIB in the Radiochemical Processing Lab RPL as well as several other instruments and adapting a 30-ton INSTRON load frame to work inside a shielded hot cell all of which make the RPL one of the gems of the Office of Science stable. In the past the RPLs mission was largely Hanford-related.Today PNNL manage- ment assesses it as a billion-dollar asset. And in order to make sure it remains such a vital part of their portfolio its important to make sure the lab is being used. One of the ways they do that is by taking part in NSUFs partnership program. Staff Engineer David Senor has been PNNLs chief liaison with the NSUF over the past five years and will complete his term as chair of the Users Organization the group that oversees that program in June 2015. Its tough to pin down PNNLs specialties because were so diverse says Senorbut I think the biggest benefit we bring to NSUF is extra post-irradiation examination PIE capacity. If you look at these types of experiments the bottleneck usually comes after the irradiation is done. So the more PIE capacity a user facility can put together the more responsive they can be to the PIs in getting the results of irradiation experiments out in a timely fashion.Another major piece of that is our irradiation experiment design and fabrication capability.Very few places in the country retain that expertise these days. So besides playing a role in preparing for PIE if there are limitations at the user facility on the front end of irradiation experiments we can help alleviate a log jam there as well. Were pretty nimble facility-wise too adds Schlahta.We have a lot of flexibility with what we can put into our modular hot cells especially with the new INSTRON load frame and thats not the case with some of the other NSUF partners. On the other side of the equation Schlahta agrees that being an NSUF partner also helps PNNL. I think this is a great example of how the DOE system should work he says. We have some very specific capabilities that give us an entre and partnering with NSUF gives us some excellent opportunities to actually demonstrate what we can do. One such opportunity currently on PNNLs PIE plate is a fuel project headed by Dr. Mehdi Balooch at the University of California Berkeley UCB.The objective is to develop a hydride fuel that would potentially replace the uranium-dioxide UO2 fuel currently used in LWRs and to explore the use of a liquid metal as a replacement for helium to fill the pellet cladding gap. The problem with UO2 is that it has a low conductivity explains Balooch so if you want to get the most energy out of it you have to go close to 2000C.The hydride fuel we are proposing has a conductivity five times higher than UO2 so in order to get the same amount of energy out of it you only have to go up to around 600C. It also has better properties in terms of neutronics and safety. Beginning in 2010 the fuel elements for the experiment were made at UCB. They were put into the Massachusetts Institute ofTechnology MIT reactor in 2011 and in the spring of 2014 three capsules of irradiated samples were sent to PNNL.All of these institutions are NSUF partner facilities. Only one capsule has been disassembled for PIE.The other two will be shipped to INLs sample library where all that material will become available to other researchers. The multifacility track followed by this experiment is a perfect example of the kind of collaboration fostered by NSUFs partnership program. We received the capsules at PNNL in March 2014 Senor saidthen we had to build the fixturing and figure out the best way to take them apart. Disassembly was done in the fall and now weve started cutting up the samples and getting into the examination phase. Experiment Manager Andy Casella organized groups of PNNL scientists with specific areas of expertise to examine the samples. Mehdi has a list of questions he would like to get answered with these samples said Casellaso were trying to figure out the best way to use the facilities we have to generate those results.That way we can tie the begin- ning of the experiment to the conclu- sion so it can be presented in a way that is easily digested. This is not the end of the story said Balooch. Were hoping to get to the point where a decision can be made as to whether its worth it to continue this research or not and 90 percent of that takes place here in the PIE work. Its 2014 ANNUAL REPORT 19 not going to answer all of the questions as to whether we can replace the UO2 with this new fuel thats just not going to happen with this small project and small budget. And in the course of answering these initial questions Senor saidweve already started to generate a whole bunch of new questions which will lay the foundation for follow-on studies if someone decides to take this to the next step.To introduce a new fuel or a new cladding material youre looking at a 15 to 20 year time frame to accomplish all the development and testing thats going to be required to license that new material with NRC.This project is really just step one of that process. Step two according to Balooch and the PNNL team would likely be further irradiation testing perhaps on slightly larger rodlets and at higher tempera- tures. But first theyll collaborate on a series of papers and reports outlining all the data collected from this experi- ment. From there its up to someone else to pick up the ball and run with it perhaps through an NSUF proposal that would utilize the unique capabilities and expertise available through its partner- ship network. I graduated from MIT said Balooch but I never thought that at some point MIT and UCB would work hand-in- hand together.This is one project where it actually happened. UCB got the ball rolling MIT got involved PNNL is involved INL is involved and one big advantage is that a lot of graduate students have benefitted from it.They come together work together and that has helped this project quite a bit. Its a great idea. There are really very limited opportu- nities for professors and their students to interact with a test reactor and to do an experiment and follow it all the way through Senor said.The NSUF creates those opportunities and were proud to be one of their partners. Nuclear Science User Facilities 20 Every year during the summer nine distinguished nuclear scientists and engineers each of them an expert in a particular field of nuclear science or education gather at the CAES in Idaho Falls Idaho. Together they comprise the Scientific Review Board SRB of what until recently was known as the Advanced Test Reactor National Scientific User Facility ATR NSUF. Dr. John Sackett is a former associate laboratory director from Argonne National Laboratory. Now retired he acts as the SRBs chairman. The NSUF mission he says is to support the advancement of nuclear science and technology in the United States by providing nuclear energy researchers access to world-class testing facilities. The SRB is an advisory group that helps the NSUF fulfill that mission in the best way possible. The board brings together experts in customer service materials irradiation post-irradiation examina- tion and fuels who want to make the NSUF as useful to its users as possible explains Dr. Sean OKelly who until recently served as deputy director of the National Institute of Standards andTechnology NIST Center for Neutron Research NCNR in Gaithersburg Maryland. SCIENTIFIC REVIEW BOARD A Guiding Hand for a Maturing Program OKelly is a good example of one of the experts he refers to.The NCNR has a users program that is considered one of the best in the world due to the reliability of its facility and the high-quality support it provides users. The users the ATR NSUF wants to attract he points out are similar to the users the NIST reactor provides its own services to so I can offer insights on whats been successful for us. As part of his duties in the NIST users program OKelly served as chief of reactor operations and engineering so he also brought a deep understanding of reactors and how they operate to the board. Basically I kept the NIST reactor operating efficiently for the users program which used the reactor as a source of neutrons for experiments.We maintained and operated the reactor at NIST 24 hours a day and the NCNR has a very good reputation for the reliability of operations and the predictability of schedule which users appreciate when they travel long distances to perform experiments. This past February OKelly accepted a position with INL as associate labora- tory director for the Advanced Test Reactor meaning he is no longer an outside expert so he has relinquished his seat on the board. Seans input 2014 ANNUAL REPORT 21 on what makes a very good users program even better was invaluable to us said Sackett. He will be hard to replace. On the other handAndrew Klein professor of nuclear engineering and health sciences at Oregon State University OSU is not going anywhere. He brings to the SRB a unique perspective on the world of higher education especially with regard to one of the NSUFs primary user groups university researchers and their graduate-level nuclear science and materials students.Take a quick look at his curriculum vitae and youll see hes the head of the Depart- ment of Nuclear Engineering and Radiation Health Physics at OSU the director of educational partnerships at INL and a member of the board of directors for the Foundation for Nuclear Studies. He also brings a certain amount of self-interest to his position. Im a university researcher myself so Im a potential user of the NSUF he explains. I have a vested interest in being the champion of its customers and finding ways that the program can serve them as completely as possible. To that end while the members of the SRB support the NSUF and the role it plays in developing nuclear energy as a safe reliable source of energy theyre not cheerleaders. Were an independent advisory board says OKelly so we do criticize.We dont hold anything back. Nuclear Science User Facilities 22 Our recommendations are only that recommendations Sackett said. We dont have the authority to require the NSUF to take any specific action. But those recommendations do carry the weight of a higher authority as the SRB actually represents the mission of the United States DOE the govern- ment agency that oversees the NSUF. We report to the DOE on our view of the quality of the technical and administrative work being done at the NSUF Sackett continues. And every year weve given it very high marks. As an example of a recent recom- mendation the board made that the NSUF acted upon OKelly points to an improvement suggested to Rory Kennedy the new director of the NSUF. The board noticed that getting samples irradiated at the ATR and then shipping them to graduate students working on projects was taking far too long he says. After all graduate students only have so many years to find a project they feel comfortable working on.And the longer they have to wait for samples the less time they have to complete their theses. So the review board recommended that the ATR NSUF improve the throughput of its samples. The user facility did better than that. It increased student access to the sample library of already irradiated material leftover from previous experiments. The board agreed this was an excel- lent idea said Klein. Students can now get their irradiated samples much faster than if they had to wait their turn on the reactor. Of course there is no dearth of proj- ects still going through the reactor but for students anxious to start their training having test materials readily in hand is a big advantage. New Voices. New Perspectives. Although the NSUF was initially primarily accessed by universities and national laboratories now the SRB and NSUF are seeking more participa- tion from the private sector and the international community. 2014 ANNUAL REPORT 23 The board includes a member from the United Kingdom. International participation on the board has always been very impor- tant Sackett said. In fact when the board was originally formed it included a member from Japan. Today Simon Pimblott professor of radiation chemistry at the University of Manchester in England serves on the SRB. Pimblotts presence is significant for two reasons. First it promotes a cooperative effort between the NSUF and the United Kingdoms newly founded National Nuclear User Facility NNUF. Secondly building a new nuclear test facility is a very expensive proposition. The United States government could pursue a truly robust nuclear power development program on its own but it is more cost-effective to leverage resources from national laboratories universities and international partners. At present Sackett said efforts to develop relationships with partners overseas are constrained by limited budgets on both sides but we all hope that funding increases from our respective governments are in the works because there is so much value in sharing our test facility capabilities. By the same token having represen- tatives from the private sector on the SRB not only provides valuable resources from a research point of view it also lends a pragmatic real-world perspective to the NSUFs work. For example one of the major concerns private industry has is that the research and work at the NSUF not only be relevant to commercial needs but that its done on budget and on schedule. In the nuclear science field the interests of private industry and academia are closely intertwined Klein said. For example one of the things our industry partner has stressed is balancing the costs and effectiveness of the technical support provided during experiments. One of the ways weve sought that balance is by optimizing our research processes on projects that begin at one facility and then must be moved to another when substantial demands on time exist at both facilities. The SRB has helped establish lines of communication through which industry or universities can say to the NSUFWe have a need. Can someone do a project on this OKelly said. So now you have university profes- sors working on problems private industry has identified that before no one at the national labs or universities knew were out there. Now we do. The SRB hopes that the NSUF can attract more members from the private sector as well as from the international community.This seeming expansion of the NSUFs mission has led to a subtle yet dramatic shift in how it presents itself. Whats in a name Everything. If you didnt notice it on the cover of this annual report the name of AdvancedTest Reactor National Scientific User Facility has been changed to the Nuclear Science User Facilities.As we said its not a major shift. Its hardly noticeable in fact. Perhaps more apparent is the logo. The graphic of the United States with circles emanating from the ATR NSUF in southern Idaho has been replaced by a world globe. Both the new name and new logo signify the logical transition from our original focus on the AdvancedTest Reactor to what we have grown into says Sackett pointing out that the reference to ATR has been removed. We are an extended family of facili- ties providing research and technical expertise to a broad spectrum of scientific needs that go beyond just the materials and fuels sciences but embrace other fields as well. The name change better represents the broadand growingset of capabilities available to researchers Klein said. Its a reflection of the value not only of the NSUF but the SRB itself OKelly said. Board members want the NSUF to achieve its highest poten- tial.They want it to grow to become significantly larger than it is today and to provide broad researcher access to the national nuclear research infra- structure but theyre not biased about how it gets there they want to do it right.And the new name shows that were heading in the right direction. Nuclear Science User Facilities 24 INTERVIEWWITHTHE DIRECTOR Dr. Rory Kennedy was appointed director of the Advanced Test Reactor National Scientific User Facility in January 2014 replacing Jeff Terry who returned to his faculty position at the Illinois Institute of Technology. He is the first full-time director in the ATR NSUFs eight-year history an indication of how the organiza- tion and the job itself have grown. Recently Kennedy sat down with Sarah Robertson NSUFs commu- nications specialist to talk about where the NSUF is now where its going and how it is re-emphasizing the resources skills and strengths available through the NSUFs nuclear research partnerships. QFirst of allcan you briefly give us a little background on yourselfWhat did you do before you became director of the user facility A Before I took this position I was the national technical lead for Metallic FuelTechnology Development within the Advanced Fuels Campaign of DOEs Fuel Cycle Research and Devel- opment Program. QYouve been director for just over a year now. What have you learned in that time A Ive learned how important it is to maintain a good relationship with the DOE to work very closely with them and to communicate with them on a regular basis.They act as our primary sponsor and interface with federal government policy makers. So that was a big learning experience since I didnt have quite the degree of contact with the DOE before as I do now. I also learned how useful and valu- able the user facility as a whole is to the advancement of nuclear research. Before I concentrated on the individual program I was working. I had a very directed scope that commanded my full attention. Now as director of the user facility I am responsible for a much broader scope of research.There are many different types of investigations being conducted through the ATR NSUF and I have to have a significant degree of understanding of each of them. Its given me a greater appre- ciation for the set of diverse talents required to reach our common goal. Its also made me more aware of the gaps in the data and the fundamental knowledge that I didnt realize existed in the field of nuclear fuels and mate- rials that need to be addressed. Q What sort of gaps A Well lets take cladding as an example.There are cladding mate- rials that have undergone systematic studies over the last 60-plus years so youd expect to be able to find data on the irradiation behavior of these materials at pretty much any temperature neutron flux and time in reactor. But in many cases the studies that would provide the specific data youre looking for have never been performed. I find that J. Rory Kennedy Director 208 526-5522 rory.kennedyinl.gov An Interview with J. Rory Kennedy Ph.D. 2014 ANNUAL REPORT 25 Nuclear Science User Facilities 26 rather surprising because over that period of time I would have assumed that all that information would have been collected.The answer likely is that the scientists involved limited their research to fairly constrained sets of reactor conditions.As we move forward with advanced reactor designs we need to understand the effects of irradiation over a broader range of conditionsfilling in the gaps in the needed data if you will to really gain a better understanding of how these types of materials can be used in future reactors. Q I know were in the process of changing the name of the user facility.Can you tell us something about that AThis was under consideration before I became director and I think its a good idea because if you look at the user facility and how it has grown since 2007 the current name does not adequately reflect that growth. Of course the original concept of the user facility centered around INL and its AdvancedTest Reactor along with our PIE capabilities and the instrumentation that we were devel- oping. But then its project load started expanding fairly quickly as the interest in the ATR and INL PIE facilities increased so that we couldnt handle it all and we started adding other existing facilities that could assume some of the workload as partners. Now along with ATR we can assign projects to the High Flux Isotope Reactor HFIR at Oak Ridge the MIT reactor at Massachusetts Institute of Technology or the PULSTAR reactor at North Carolina State University.And as we get more projects other reactors 2014 ANNUAL REPORT 27 of value might be added as part of our user facilities.The same can be said about the PIE facilities So we have evolved. Rather than just one single entity user facility we are now a network of facilities stretching from one coast to the other all dedi- cated to the study and advancement of nuclear science. And so because were expanding our resources our capabilities and the community we engage we started the conversation about changing our name to reflect this growing organization that had taken on a national scope. After a great deal of discussion we decided to change the name from the ATR National Scientific User Facility to the Nuclear Science User Facili- ties which expresses more accurately what we really are a broader national program that encompasses partner facilities around the country.And thats what we want to reflect.This also keeps the NSUF abbreviation that everyone is accustomed to. Q Another significant change in policy is project- forward funding.Can you tell us about that AYes.This was a very good sugges- tion from DOE because it alleviates the potential for fluctuations in funding from one year to the next. Up until now each project was funded one year at a time. So the possibility existedand it has happenedthat a project would have to be interrupted because there was not enough funding for it in the budget for that year. Now we get a cost estimate for the full scope of the project and the full cost of the project comes out of the funding for the year the project is approved.That money is then distrib- uted as needed over the life of the project. So instead of going year- to-year with the funding the entire amount of support is there to carry the project to its completion. Not surprisingly the facilities look very favorably on this since it allows them to plan their facilities usage and budgets better sometimes years in advance. Another thing weve done that affects forward-funding is weve expanded the maximum length of time well support large irradiation and PIE projects up to seven years. Here we envision this schedule to cover one year for design and fabrication three years irradiation and three years PIE. This year is our first experience with forward funding and were learning a lot. But we have a schedule we have the funding we have the people and we have the facilities. So even with its challenges were excited about the process improvements forward- funding can provide us. Q Can you tell me about some other changes that youre working on A Were learning and were insti- tuting other practices that will help us improve our processes even more. For example this year weve set up a number of metrics that will help us evaluate the effectiveness of these new processes and how the partnership program as a whole is progressing. Another thing we changed in 2014 was designating who can lead a project. Previously the policy was that only a university could submit an application to the NSUF.This year weve opened the submission process up to everyone. So now university national laboratory private industry and small business can all apply as principal investigator. And this is true not only for entities in the U.S. but foreign researchers as well.The only stipulation is that they have a U.S. entity as a co-lead.We want to make the benefits of the partnership program available to the broadest possible number of users and the broadest possible number of ideas. We want to encourage as many good ideas as we can no matter where they come from. Q It sounds like youre looking to expand the partnership program. AYes we are and on two fronts.Weve always been open to increasing the number of partners that offer unique capabilities but were also trying to increase our use of the partners to engage them more and to allow them to be more involved as significant contributors to the user community. Its all part of why were changing the name to be more inclusive.We want the user facility as an organization to support the partners and we want the partners to support the user facility as an organization. Q In your letter at the beginning of this annual reportyou mention two resources youre developing to help researchers in their work.Can you talk about those A Were creating a database of all the capabilities available to DOE NE-supported researchers through this effort.There are a whole slew of analyses well be doing with this database cost to maintain facilities cost to replace equipment facilities utilization cost of utilization equip- ment condition anticipated remaining life and lots more. Connected to that is something were calling a gap analysis which is basi- cally compiling a list of the capabilities we may want to invest in according to Nuclear Science User Facilities 28 a cost-benefit analysis. For example if we identify a required or desirable post-irradiation examination tech- nology thats not currently accessible in the U.S. we do an assessment of how we might establish it in the U.S. and where. The other side of the gap analysis is to identify if there are capabilities with a high degree of redundancy or under utilization. Is the same PIE technique duplicated at several facilities but is only being used 20 percent of the time So if a facility asks to add the same technology we may suggest that the institution consider using the already established yet under-utilized capability. Its a big job to create this database manage it and then employ it to help DOE use its capabilities more cost-efficiently. This capability database will be made available to the public so users can come in and determine for themselves if the technology or area of expertise they need is already available either in the U.S. or internationally or if it has potential as an area of opportunity for their own institution.Also individual researchers can use the database to help develop their research projects and proposals. The second resource were continuing to develop is the NSUF sample library. It has proven to be very valuable from cost-efficiency and time-saving perspectives and so we are actively moving to enhance it.There are a lot of irradiated materials that have been sitting around for quite some time. Theres no reason for us to do an irradiation test that could cost upwards of a million dollars if the irradiated samples already exist.And there may already be a knowledge base associated with those samples that we havent identified. It is very important though that the samples have a very good pedigree documented. As we go forward we will continue to add to the sample library with new and unique samples as we conduct additional irradiation tests. In addi- tion we are engaging the international community in the hopes that there could eventually be an international sample library of materials available to nuclear researchers. DOE is very interested in pursuing this kind of cooperative effort and the NSUF is more than willing to be the tool to facilitate it. In fact were trying to establish a collaboration with the United Kingdom which just recently established its own user facility. One of the conversations weve had with them dealt with the sample library.They have a large number of archived samples and agree that we should work toward collaborating in some way.Were hoping that once the momentum is achieved with the UK we can use it to forge relationships with other countries as well. Q Finallywhats your view of the future of nuclear energyboth in the U.S.and abroad A Im very optimistic both on the domestic and international fronts. The fact is nuclear energy has a very high value as an inexpensive source of energy over the long term.Technolo- gies are being investigated that can allow us to build safer more efficient reactors that can last perhaps 100 years or more. So while their initial construction costs can be high the cost of the power it produces over the life of that reactor is actually very low. At the same time nuclear power helps reduce the carbon emissions produced by fossil-fuel power generation. Its so gratifying that the NSUF and the researchers and staff members at our various partner facilities are making a significant contribution to making nuclear energy a safer more efficient more environmentally responsible energy resource for the entire world. 2014 ANNUAL REPORT 29 Nuclear Science User Facilities 30 2014 ANNUAL REPORT 31 NSUF A Model for Collaboration NSUF and its partner facilities repre- sent a prototype laboratory for the future.This unique model utilizes a distributed partnership with each facility bringing exceptional capa- bilities to the relationship including reactors beamlines state-of-the-art instruments hot cells and most importantly expert mentors.Together these capabilities and people create a nationwide infrastructure that allows the best ideas to be proven using the most advanced capabilities.Through NSUF university researchers and their collaborators are building on current knowledge to better understand the complex behavior of materials and fuels in a nuclear reactor. In 2014 NSUFs partnership program had eight universities two national laboratories and added one industry partner.The partner facility capabilities greatly expand the types of research offered to users.The avenues opened through these partnerships facilitate cooperative research across the country matching people with capa- bilities and students with mentors. In 2014 NSUF included INL and the following institutions Illinois Institute ofTechnology Massachusetts Institute ofTechnology North Carolina State University Oak Ridge National Laboratory Pacific Northwest National Laboratory Purdue University University of California Berkeley University of Michigan University of Nevada LasVegas University of Wisconsin Westinghouse The pages that follow contain specific details on the capabilities of NSUF its partners and how to access these capabilities through the calls for proposals.There is also informa- tion on the Users Meeting a yearly event hosted by NSUF designed to instruct and inform.This event is free of charge to interested persons and a number of scholarships for travel and hotel are offered to students and faculty.Take time to familiarize yourself with the many opportunities offered by NSUF and consider submit- ting a proposal or two. NSUF Research Supports DOE-NE Missions The U.S. DOE-NE organizes its research and development activities based on four main objectives that address challenges to expanding the use of nuclear power Develop technologies and other solutions that can improve the reli- ability sustain the safety and extend the life of current reactors. Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the administrations energy security and climate change goals. Develop sustainable nuclear fuel cycles. Understand and minimize the risks of nuclear proliferation and terrorism. NSUF research addresses a number of these mission needs. Most of the research contained in this report looks at either understanding the mecha- nisms of radiation on materials and fuels to address the challenges of the aging current fleet or looks at mate- rials and fuels for the next generation of reactors.To be eligible as an NSUF research project the research must support at least one of the DOE-NE missions. For specifc information on DOE missions go to httpwww. energy.govnemission. To learn more about proposing a research project visit the NSUF website httpnsuf.inl.gov. PROGRAM OVERVIEW Nuclear Science User Facilities 32 each of which can contain multiple experiments. Experiment positions vary in size from 0.5 to 5 inches in diameter 1.27 to 12.7 centimeters and all are 48 inches 121.92 centi- meters long.The peak thermal flux is 1x1015 ncm2 -sec and fast flux is 5x1014 ncm -sec when operating at full power of 250 MW. There is a hydraulic shuttle irradiation system which allows experiments to be inserted and removed during reactor operation and pressurized water reactor PWR loops which enable tests to be performed at prototypical PWR operating conditions. Idaho National Laboratory Advanced Test Reactor Critical Facility ATRC is a low-power version same size and geometry of the higher- powered ATR core. It is operated at power levels less than 5 KW with typical operating power levels of 600 W or less.ATRC is primarily used to provide data for the design and safe operation of experiments for ATR.ATRC is also used to supply core performance data for the restart of ATR after periodic core internals replacement. Occasionally ATRC is used to perform low-power irradia- tion of experiments. REACTOR Capabilities OS-1 OS-3 OS-8 OS-13 OS-18 OS-4 OS-9 OS-14 OS-19 OS-5 OS-10 OS-15 OS-20 OS-6 OS-11 OS-18 OS-21 OS-7 OS-12 OS-17 OS-22 OS-2 ON-8 ON-3 ON-9 ON-4 ON-10 ON-5 ON-11 ON-6 ON-12 ON-7 ON-1 ON-2 13-GA50331 ATR NSUF offers access to a number of reactors.ATR is located at the ATR Complex ATR on the INL Site and has been operating continuously since 1967. In recent years the reactor has been used for a wide variety of government and privately sponsored research.The ATRC reactor is low-power version of ATR. The MIT reactor is a 5-MW reactor with positions for in-core fuels and materials experiments. Oak Ridge National Laboratorys ORNL HFIR is an 85-MW reactor offering steady-state neutron flux and a variety of experiment positions.The PULSTAR reactor at North Carolina State University is a pool-type reactor that offers response charac- teristics similar to commercial light water power reactors. Idaho National Laboratory Advanced Test Reactor ATR is a water-cooled high-flux test reactor with a unique serpentine design that allows large power varia- tions among its flux traps.The reac- tors curved fuel arrangement places fuel closer on all sides of the flux trap positions than is possible in a rect- angular grid.The reactor has nine of these high-intensity neutron flux traps and 68 additional irradiation positions inside the reactor core reflector tank ATRs serpentine design allows a variety of experiment configurations. 2014 ANNUAL REPORT 33 Top of the HFIR reactor. Aerial view of the ATRC reactor core and bridge. Oak Ridge National Laboratory High Flux Isotope Reactor HFIR is a versatile 85-MW research reactor offering the highest steady- state neutron flux in the western world.With a peak thermal flux of 2.5x1015 ncm2 -s and a peak fast flux of 1.1x1015 ncm2 -s HFIR is able to quickly generate isotopes that require multiple neutron captures and perform materials irradiations that simulate lifetimes of power reactor use in a fraction of the time. HFIR typi- cally operates seven cycles per year each cycle lasting between 23 and 26 days.Associated irradiation processing facilities include the HydraulicTube Facility PneumaticTube Facilities for Neutron Activation Analysis NAA and Gamma Irradiation Facility. Massachusetts Institute of Technology Reactor Massachusetts Institute ofTechnology MITR is a 5-MW tank-type research reactor. It has three positions available for in-core fuel and materials experi- ments over a wide range of condi- tions.Water loops at pressurized water reactorboiling water reactor PWR BWR conditions high-temperature gas reactor environments at tempera- tures up to 1400C and fuel tests at light water LWR temperatures have Annular fuel rig in the MITR core. Downward view of the PULSTAR reactor pool. been operated and custom conditions can also be provided.A variety of instrumentation and support facilities are available. Fast and thermal neutron fluxes are up to 1014 and 5x1014 n cm2 -s. MITR has received approval from the Nuclear Regulatory Commis- sion for a power increase to 6 MW which will enhance the neutron fluxes by 20 percent. North Carolina State University PULSTAR Reactor The PULSTAR reactor is a 1-MW pool- type nuclear research reactor located in North Carolina State Universitys NCSU Burlington Engineering Laboratories.The reactor one of two PULSTAR reactors built and the only one still in operation uses 4 percent enriched pin-type fuel consisting of uranium dioxide pellets in zircaloy cladding.The fuel provides response characteristics that are very similar to commercial light water power reactors.These characteristics allow teaching experiments to measure moderator temperature and power reactivity coefcients including Doppler feedback. In 2007 the PULSTAR reactor produced the most intense low-energy positron beam with the highest positron rate of any comparable facility worldwide. Nuclear Science User Facilities 34 POST-IRRADIATION EXAMINATION Capabilities ATR NSUF offers researchers access to a broad range of PIE facilities. These include capabilities at INLs MFC the Microscopy and Charac- terization Suite MaCS at the Center for Advanced Energy Studies the Nuclear Services Laboratories at North Carolina State University hot cells radiological laboratories and the Low Activation Materials Development and Analysis LAMDA facility at Oak Ridge National Laboratory the Radiochemistry and Materials Science andTechnology Laboratories at Pacific Northwest National Laboratory the Interaction of Materials with Particles and ComponentsTesting IMPACT Hot Fuel Examination Facility located at the Materials and Fuels Complex at DOEs INL Site in Idaho. Transmission electron microscope one of many PIE capabilities in the Microscopy Characterization Suite MaCS at the Center for Advanced Energy Studies in Idaho Falls Idaho. facility at Purdue University several instruments from the Nuclear Materials Laboratory at University of California Berkeley the Irradiated Materials Complex at the University of Michigan the Harry Reid Center Radiochemistry Laboratories at the University of Nevada LasVegas and the Characterization Laboratory for Irradiated Materials at the University of Wisconsin. Idaho National Laboratory Hot Fuel Examination Facility Analytical Laboratory Electron Microscopy Laboratory Hot Fuel Examination Facility HFEF is a large alpha-gamma hot cell facility dedicated to remote examina- tion of highly irradiated fuel and structural materials. Its capabilities include nondestructive and destruc- tive examinations.The facility also offers a 250-kWth Training Research Isotope General Atomics TRIGA reactor used for neutron radiography to examine internal features of fuel elements and assemblies. The Analytical Laboratory is dedi- cated to analytical chemistry of irradiated and radioactive materials. It offers NIST-traceable chemical and isotopic analysis of irradiated fuel and material via a wide range of spectrometric techniques. 2014 ANNUAL REPORT 35 The Positronium Annihilation Lifetime Spectrometer located in the PULSTAR reactor facility on the NC State North Campus in Raleigh N.C. The Scanning Electron Microscope in Oak Ridge National Laboratorys LAMDA facility. A hot cell in the Radiochemistry Processing Laboratory at Pacific Northwest National Laboratory. The Electron Microscopy Labora- tory EML is dedicated to materials characterization primarily using transmission electron scanning electron and optical microscopy.The EML also houses a dual-beam FIB that allows examination and small-sample preparation of radioactive materials. Center for Advanced Energy Studies Microscopy and Characterization Suite The MaCS is equipped to handle low- level radiological samples as well as non-radiological samples. MaCS offers several high-end pieces of equipment including a local electrode atom probe LEAP automated hardness tester scanning electron microscope SEM nano indenter and atomic force microscopeTEM and focused ion beam. North Carolina State University Nuclear Services Laboratories Post-irradiation examination capa- bilities at NCSUs Nuclear Services Laboratories include neutron activa- tion analysis radiography and imaging capabilities and positron spectrometry. Oak Ridge National Laboratory Hot Cells Radiological Laboratories LAMDA Facility ORNL hot cells and radiological laboratories offer a wide variety of research and development and production capabilities from radiochemistry and isotope packaging to materials testing to irradiated fuels examination. Facilities include the Irradiated Materials Examination and Testing IMET facility Irradiated Fuels Examination Laboratory IFEL and Radiochemical Engineering Development Center REDC. The Low Activation Materials Development and Analysis LAMDA Laboratory added in 2012 offers post- irradiation examination capabilities including refractory element test stands for tensile testing optical and scanning electron microscopes and thermal diffusivity and density measurement equipment. Pacific Northwest National Laboratory Radiochemistry Processing Laboratory Materials Science and Technology Laboratory The RPL and the Materials Science andTechnology Laboratory MSTL offer a wide range of specialized equipment for handling and testing fuels and materials. Capabilities include experiment hardware design fabrication and assembly testing facilities for both nonradioactive and radioactive structural materials and the advanced characterization of unirradiated and irradiated fuels and materials using instruments including TEM SEM and optical microscopy. Nuclear Science User Facilities 36 Capability at the Irradiated Materials Complex on the UM campus at Ann Arbor Michigan. The IMPACT facility at Purdue University. Purdue University IMPACT Facility The IMPACT facility offers a wide range of spectroscopy techniques to study the surface of materials.The IMPACT facility houses a variety of examination instruments including low-energy scattering spectroscopy LEISS X-ray photoelectron spectros- copy XPS auger electron spectros- copy AES extreme ultraviolet reflec- tometry EUVR extreme ultraviolet EUV photoelectron spectroscopy and mass spectrometry. University of California Berkeley Nuclear Materials Laboratory The Nuclear Materials Labora- tory provides several capabilities for examining irradiated material samples including a nano-indenta- tion system for nano and microscale hardness testing at ambient and elevated temperature and inert environments positron annihilation spectroscopy and warm sample preparation polishing cutting grinding and mounting. University of Michigan Irradiated Materials Complex The Irradiated Materials Complex provides laboratories and hot cells with capabilities for conducting high- temperature mechanical properties and corrosion and stress corrosion cracking experiments on neutron- irradiated materials in an aqueous environment including supercritical water and for characterizing the fracture surfaces after failure. UC Berkeley nano- indentation system. 2014 ANNUAL REPORT 37 PIE capabilities at the Harry Reid Center Radiochemistry Laboratories located on the UNLV campus in Las Vegas Nevada. Operators use manipulators to perform work at the Westinghouse Hot Cell Lab. University of Nevada Las Vegas Harry Reid Center Radiochemistry Laboratories Post-irradiation examination capabilities at the Radiochemistry Laboratories include metallographic microscopy X-ray powder diffrac- tion Rietveld analysis SEM and TEM electron probe microanalysis and X-ray uorescence spectrometry. University of Wisconsin Characterization Laboratory for Irradiated Materials The Characterization Laboratory for Irradiated Materials offers PIE capa- bilities including SEM andTEM on neutron-irradiated materials. Westinghouse Materials Center of Excellence Laboratories Westinghouse offers its Materials Center of Excellence Laboratories MCOE Hot Cell Facility and accompanying laboratories to provide experimental support to ATR-related nuclear energy materials research programs.TheWestinghouse facilities in Churchill Pennsylvania are housed in four cells that provide a broad range of testing evaluation and characterization capabilities for both unirradiated and irradiated materials. In-place capabilities include the ability to test under a variety of environments an extensive mechanical testing laboratory a specialized corrosion and stress corrosion cracking lab and materials microstructure and chemical characterization instruments. Specialized facilities are also available to measure the radioactivity properties of materials under investigation as well as neutron and gamma sources facilities which can be employed to assess materials response to in-situ radiation. A JEOL 200CX TEM equipped with EDS and scanning system and an electro-polisher and dimpler at the Characterization Laboratory for Irradiated Materials located on the UW campus in Madison Wisconsin. Nuclear Science User Facilities 38 Aerial view of the Advanced Photon Source at Argonne National Laboratory located in Argonne Illinois. ATR NSUF offers researchers access to a broad range of facilities with beamlines including accelerator facilities for radiation damage experi- ments synchrotron radiation studies neutron diffraction and imaging as well as positron and neutron activa- tion analysis. In 2014 the ATR NSUF program offered researchers access to four university partner beamline facilities. These include the Illinois Institute of Technology Materials Research Collab- orative AccessTeam MRCAT beamline at Argonnes Advanced Photon Source the PULSTAR reactor facility at North BEAMLINE Capabilities Carolina State University the University of Michigan Ion Beam Laboratory and the University ofWisconsinTandem Accelerator Ion Beam. Illinois Institute of Technology IIT MRCAT at Argonne National Laboratorys Advanced Photon Source The MRCAT beamline offers a wide array of synchrotron radiation experiment capabilities including X-ray diffraction X-ray absorption X-ray uorescence and 5-m-spot size uorescence microscopy. 2014 ANNUAL REPORT 39 North Carolina State University PULSTAR Reactor Facility The PULSTAR reactor facility offers a selection of dedicated irradiation beam port facilitiesneutron powder diffraction neutron imaging intense positron source and ultra-cold neutron source.An intense positron source has been developed to supply a high- rate positron beam to two different positronpositronium annihilation lifetime spectrometers. University of Michigan Michigan Ion Beam Laboratory The 1.7-MVTandetron accelerator in the Michigan Ion Beam Laboratory offers controlled temperature proton irradia- tion capabilities with energies up to 3.4 MeV as well as heavy ion irradiation. University of Wisconsin Tandem Accelerator Ion Beam A 1.7-MV terminal voltage tandem ion accelerator Model 5SDH-4 National Electrostatics Corporation Pelletron accelerator installed at UW features dual ion sources for producing nega- tive ions with a sputtering source or using a radio frequency RF plasma source.The analysis beamline is capable of elastic recoil detection and nuclear reaction analysis. Positron beam cave containing magnetic switchyards and transport solenoids located in the PULSTAR reactor facility on the NC State North Campus in Raleigh NC. Michigan Ion Beam Laboratory for Surface Modication and Analysis located on the UM campus in Ann Arbor Michigan. Tandem Ion Beam Accelerator located on the UW campus in Madison Wisconsin. Nuclear Science User Facilities 40 The NSUF mission is to provide nuclear energy researchers access to world-class capabilities to facilitate the advancement of nuclear science and technology.This mission is supported by providing cost-free access to state-of-the-art experimental irradiation testing and PIE facilities as well as technical assistance in design and analysis of reactor experiments. Access is granted through a competitive proposal process. NSUF offers research proposal options through an online submittal system that helps prospective researchers develop edit review and submit their proposals. NSUF staff is available to help any researcher who desires to submit a proposal. Submitted proposals should be consis- tent with the DOE-NE mission and its programmatic interests.These include the Light Water Reactor Sustainability Fuel Cycle Research and Development Advanced Modeling and Simulation Next Generation Nuclear Plant and the Generation IV Nuclear Energy Systems Initiative programs. All proposals are subject to a peer- review process before selection.All NSUF research must be non-propri- etary and results are expected to be published. Collaborations with other national laboratories federal agencies non-U.S. universities and industries are encouraged.Any U.S.-based enti- ties including universities national laboratories and industry can propose research that would utilize the MRCAT beamline at the Advanced Photon Source or would be conducted as a rapid turnaround experiment. Calls for Irradiation Post-irradiation Examination and Synchrotron Radiation Experiments Applications are submitted annually through the Consolidated Innovative Nuclear Research Funding Opportu- nity Announcement. More informa- tion is available on the NEUP website www.neup.gov. While priority will be given to proposals that further the direction of DOEs nuclear energy research programs the NSUF will consider all technical feasible proposals for scientific merit and selection. Irradiationpost-irradiation exami- nation of materials or fuels. PIE of previously irradiated materials or fuels from the NSUF sample library. Research that requires the unique capabilities of the Advanced Photon Source through the MRCAT beamline operated by the Illinois Institute ofTechnology. All proposals submitted to the open calls undergo thorough reviews for feasibility technical merit relevance CALL FOR PROPOSALS Jeff Benson Program Administrator 2014 ANNUAL REPORT 41 to the DOE-NE missions and cost.The results are compiled and provided to a panel committee who performs a final review and ranks the proposals.The ranking is given to the NSUF director. Awards are announced within two to three months of the calls closing date generally in January and June. Awards allow users cost-free access to specific NSUF and partner capabilities as determined by the program. Calls for Rapid Turnaround Experiments Rapid turnaround experiments are experiments that can be performed quicklytypically in two months or lessand include but are not limited to PIE requiring use of an instrument FIBTEM SEM etc. irradiations in the PULSTAR reactor ion beam irradiation and neutron scattering experiments. Proposals for rapid turnaround experiments are reviewed on a quarterly basis in JanuaryApril July and October and awarded based on the following rankings High PriorityProposal is awarded immediately upon review if funding is available. RecommendedProposal is placed in a queue from which awards are made approximately every other month if funding is available. Not RecommendedProposal is not awarded but the project investigators are offered an opportunity to read the review comments and then resubmit the proposal for the next call. For more information visit the NSUF website www.nsuf.inl.gov NSUF Sample Library NSUF has established a sample library as an additional pathway for research. The library contains irradiated and unirradiated samples in a wide range of material types from steel samples irradiated in fast reactors to ceramic materials irradiated in the Advanced Test Reactor. Many samples are from previous DOE-funded material and fuel development programs. Univer- sity researchers can propose to analyze these samples in a PIE-only experi- ment. Samples from the library may be used for proposals for open calls and rapid turnaround experiments. As the NSUF program continues to grow so will the sample library. To review an online list of available specimens visit the NSUF electronic system at the address above. Nuclear Science User Facilities 42 USERS MEETING The annualATR NSUF Users Meeting offers researchers days of workshops tours discussions and classes.The focus is on providing an understanding of key nuclear technology gaps capabilities required for addressing those gaps recent or emerging advances and techniques for conducting reactor experiments and PIE. Users Meeting is not just a way to learn more about ATR NSUF its capabilities and ongoing research it is also a great opportunity to meet other students scientists and engineers who are interested in responding to ATR NSUFs call for proposals. Users Meeting supports ATR NSUF as a model for the laboratory of the future where collaborative research and shared resources among universities and national laboratories will help prepare a new generation of nuclear energy professionals. The events are free of charge for students faculty and post docs as well as researchers from industry and national laboratories who are interested in materials fuels PIE and reactor-based technology devel- opment. In the six years since its inceptionATR NSUF Users Meeting has hosted 633 participants from 30 countries and 37 U.S. universities. Support to help defray travel hotel and meal expenses is offered to university faculty and students on a competitive basis. What to expect at Users Meeting Users Meeting kicks off with an introductory workshop to ATR NSUF which includes a description of current and upcoming research capabilities offered by INL and its university partners a briefing on the solicitation process and a welcome from DOE. Each year Users Week offers a number of workshops and courses for students to participate.These may vary from year to year but courses generally focus on a variety of topic-specific areas such as in-reactor instrumenta- tion fuels and materials or how to conduct radiation experiments. Participants are always offered an opportunity to tour ATR as well as INLs MFC where many PIE facilities are housed. Users who are not able to attend the Users Meeting in person now have the ability to participate in the meeting online. For more information about the Users Meeting visit our website nsuf.inl.gov. 2014 ANNUAL REPORT 43 Nuclear Science User Facilities 44 Pacific Northwest National Laboratory University of Nevada Las Vegas University of Wisconsin Madison Purdue University University of Michigan Oak Ridge National Laboratory Westinghouse North Carolina State University Massachusetts Institute of Technology University of California Berkeley Illinois Institute of Technology INL PIsUser Institutions Partners As of 2014 DISTRIBUTED PARTNERSHIP 2014 ANNUAL REPORT 45 Nuclear Science User Facilities 46 2014 ANNUAL REPORT 47 NSUF AWARDED PROJECTS Awarded Reports.......................... 48-135 Industry Program Reports......... 136-141 Nuclear Science User Facilities 48 The ability to conduct fast neutron irradiation tests is essential to meeting fuels and materials development requirements for future nuclear reactors.At the same time the lack of domestic fast neutron testing capabilities hinders the development of advanced reactors. The concept behind this project is to equip one of the corner lobes of the Advanced Test Reactor ATR with a thermal neutron filter.This material comprised of hafnium-aluminide Al3Hf particles 23 by volume in an aluminum matrix Al3Hf- Al will absorb thermal neutrons and booster fuel augmenting the neutron flux and heat transfer from the experiment to pressurized water cooling channels. Thermal analyses conducted on a candidate configuration confirmed that the design of the water-cooled Al3Hf-Al absorber block is capable of keeping the temperature of all system components below their maximum allowable limits. However the thermo- physical properties of Al3Hf have never been measured nor have the effects of irradiation on these properties ever been determined. It is essential therefore to obtain data on the effects of irradiation including corrosion behavior and radioactive decay products on the thermophysical and mechanical properties of both Al3Hf intermetallic and Al3Hf-Al composite before we can proceed with the design and optimization of the filter. Project Description The purpose of this project is to evaluate the properties and behavior of this new material. Specific objectives are to determine 1. The thermophysical and mechanical properties of Al3Hf in- termetallic and Al3Hf-Al compos- ite at different temperatures. Irradiation Effects on Thermophysical Properties of Hafnium- Aluminide Composite A Concept for Fast Neutron Testing at ATR Heng Ban Utah State University USU heng.banusu.edu Figure 1. Calculated dpa values and total fluence for experiment. 2014 ANNUAL REPORT 49 2. The effects of irradiation on the thermophysical and material prop- erties of Al3Hf intermetallic and Al3Hf-Al composite and physical morphological metallurgical and microstructural changes of Al3Hf- Al composite after different cycles of irradiation. 3. The decay products of hafnium Hf-179m1 vs. Hf-179m2 and corrosion behavior of the Al3 composite. Successful completion of the project will 1. Provide the necessary data for the development of fast neutron test capabilities at ATR. 2. Fill a knowledge gap on the basic properties of the Al3Hf intermetal- lic and Al3Hf-Al composite. 3. Advance the scientific understand- ing of the irradiation effects on these materials. The end result in terms of the data and fundamental understanding obtained will directly support DOEs mission and benefit the science community in general. Accomplishments During FY 2014 Zilong Hua Utah State University USU performed 3D microstructural reconstruction of unirradiated specimens of the HfAl3- Al metal matrix composite material developed for this project. Focused ion beam FIB milling and electron backscatter diffraction EBSD was performed using the FEI Quanta 3D field emission gun FEG located at CAES.The gallium ions from the FIB were found to be very damaging to the HFAl3-Al so a new procedure was developed to enable the acquisi- tion of acceptable Kikuchi patterns. The serial scans were reconstructed using Dream.3D software and visual- ized using ParaView.This work is pioneering in that 3D microstructural reconstruction has never before been attempted on this material.The procedure was developed using an unirradiated specimen in preparation for work with an irradiated specimen. Completed post-irradiation examina- tion PIE of specimens irradiated in the ATR includes 1. Gamma scans of 18 specimens 2. X-ray diffraction XRD on four specimens 3. Scanning electron microscopy SEM on one specimen 4. Differential scanning calorimeter DSC testing on nine specimens 5. Density measurements on nine specimens DSC results show a marked exotherm on the first heating cycle a mani- festation of radiation damage in the material. DSC was performed on irradiated specimens with 20.0 28.4 and 36.5 vol HfAl3 and compared to similar measurements for the unir- radiated material.The specific heat of the irradiated material was more than 50 higher than that of the unirradi- ated material. Results of the flux monitor analysis were interpreted and published.The performance of the material was evalu- ated by placing neutron fluence moni- tors within shrouded and unshrouded holders and irradiating them in the ATR for up to four cycles.The irradiation assembly consisted of eight capsules containing flux monitors placed in holders fabricated from this new mate- rial referred to as shrouded or in 6061 aluminum alloy holders referred to as unshrouded.The adjusted neutron fluences were calculated and grouped into three binsthermal epithermal and fastto evaluate the spectral shift created by the new material.A comparison of shrouded vs. unshrouded fluence monitors showed a thermal fluence decrease of approxi- mately 11 for the shrouded monitors. For all capsules the fast-to-thermal neutron ratio was higher for the flux monitors shrouded with the HfAl3- Al composite material whereas the ratio is nearly uniform for the wires shrouded by the 6061 aluminum material.The fast-to-thermal ratio appears to be fairly consistent for the unshrouded flux wires regardless of irradiation position e.g. height in the reactor or total fluence e.g. MWd. Nuclear Science User Facilities 50 Future Activities Research yet to be completed on the irradiated materials includes 1. Measurements of thermal conduc- tivity using the laser flash method on nine specimens 2. Thermal expansion measurements on three specimens 3. 3D EBSD of one specimen 4. Transmission electron microscopy TEM utilizing local electrode atom probe LEAP of one specimen 5. Tensile and hardness tests on 11 specimens Zilong Hua and Donna Guillen of INL are developing a model of the thermal conductivity in both the irradiated and unirradiated materials using the EBSD data that has been reconstructed to 3D. Using INLs MARMOT code they will further investigate the effects of radiation by comparing the response of irradiated vs. unirradiated material to a heat source applied at the boundaries of the material.Thermal conductivity is a key parameter because the material is being developed as a conduction- cooled neutron absorber block. Publications and Presentations 1. D.P. Guillen L.R. Greenwood and J.R. Parry 2014 High Conduc- tion Neutron Absorber to Simu- late Fast Reactor Environment in an ExistingTest Reactor Journal of Radioanalytical and Nuclear Chemistry Vol. 302 No. 1 pp. 413424. 2. D.P. Guillen Z. Hua and H. Ban 2014 Procedure For 3D Mi- crostructure Reconstruction of a Heterogeneous Metal Matrix Composite Material 3D Materi- als Science 2014 AnnecyFranceJune 29July 22014. 3. D.P. Guillen J. Burns Z. Hua and H. Ban 2014Microstructure of Aluminum Matrix in Composite Absorber Block Material2014TMS MeetingSan DiegoCAFeb.16-202014. See additional publications from other years in the Media Library on the NSUF website. The opportunity to use state- of-the-art instruments and work with top level researchers in the CAES Microscopy Suite is a great experience that has provided me with valuable skills to kickstart my career. Zilong HuaUSU Postdoctoral Researcher Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Idaho National Laboratory AdvancedTest Reactor Collaborators Idaho National Laboratory Donna Post Guillen principal investigator University of Nevada LasVegas Thomas Hartmann co-principal investigator Utah State University Heng Ban principal investigator Zilong Hua postdoctoral researcherAdam Zabriskie graduate student Kurt Harris undergraduate student Heather Wampler undergraduate student 2014 ANNUAL REPORT 51 This experiment will provide the necessary data for the development of a new material that will enable a robust fast neutron test capability at ATR. Figure 2. Material microstructure reconstructed from EBSD images. Nuclear Science User Facilities 52 The critical need for accurate nuclear data has been pointed out in recent studies devoted to Generation IV systems.The very high mass actinides can play a significant role in the feasibility assessment of innovative fuel cycles. For example the build-up of 252 californium Cf when recycling all transuranic waste TRU in a light water reactor LWR leads to increased neutron emissions that could impact the fuel fabrication process.This experiment will generate enough data to validate models and provide better neutron cross-sections for advanced reactor designers for years to come.As a consequence nuclear data of higher mass transura- nics should be significantly improved. Project Description The principal goal of this experiment is to irradiate very pure actinide samples in ATR and determine the amounts of resulting transmuta- tion products. Determining nuclide densities before and after neutron irradiation will allow the inference of energy-integrated neutron cross- sections.This type of information together with neutron cross-section differential measurements is ulti- mately used by physicists in charge of nuclear data evaluations for the Evaluated Nuclear Data File ENDF. In order to obtain effective neutron capture cross-sections corresponding to different neutron spectra ranging from fast to epithermal three sets of actinide samples were irradiated.The first one was filtered with cadmium Cd and the other two were filtered with 5-mm and 10-mm thicknesses of boron B.The determination of atom densities before and after irradiation will be carried out using inductively- coupled plasma mass spectrometry ICPMS at the INLs Analytical Lab and the accelerator mass spectrometry AMS at the ArgonneTandem Linac Accelerator System ATLAS facility. Measurement of Actinide Neutronic Transmutation Rates with Accelerator Mass Spectroscopy MANTRA George Imel Idaho State University gimelisu.edu FIGURE 1. OM RD technician Crystal Poole prepares a non-rad test sample to run through the multicollector left. INL chemist Jeffrey Berg presses test samples right. 2014 ANNUAL REPORT 53 Creating two different sets of inde- pendent measurements will increase confidence in the results. This project has been funded in part by the DOE Office of Science in addition to the ATR NSUF and has been given the name MANTRA. It became an official ATR NSUF project in January 2010. Accomplishments The first two irradiations were completed in January 2013.The third and last sample irradiation was completed in January 2014 after two cycles in ATR. Under neutron irradia- tion these isotopes transmute into other isotopes and even though the number of transmutation products at the end of the irradiation is relatively small it is sufficient to infer the neutron capture cross-section if the measurements are precise enough. Measurements of isotopic ratios in most of the samples were finalized in 2014 using the Multi-Collector Induc- tively Coupled Plasma Mass Spectrom- eter MS-ICP-MS.This project was very successful with about 65 samples initial irradiated characterized in a relatively short period of time.All the isotopic ratios of interest have an associated 2-sigma uncertainty of less than 1. The measurements showed that after irradiation less than 1 of the initial material was transmuted when the 5-mm B filter was used whereas up to 25 was transmuted when the Cd filter was used.This was expected as the capture reaction rates in the resonance region seen by the Cd-filtered samples are higher than those in the faster neutron region seen by the 5-mm B-filtered samples.The detailed ATR as-run calculations using the Monte Carlo N-Particle MCNP code proved more challenging than originally thought however signifi- cant progress was made in 2014. In particular Jim Sterbentz the analyst in charge as well as Idaho State University ISU Ph.D. student Jyothier Kumar Nimmagadda showed that a detailed modeling of the neutron self-shielding found in some of the samples is crucial to reproducing the physical phenomena during irradia- tion.This will complicate the calcula- tions and will require a more detailed model than was originally planned. These analyses will be completed in 2015 and will allow us to start interacting with the nuclear data community and in particular the evaluators in charge of the nuclear data files. In April 2014 based on the recommendations of INL Fellow Pino Palmiotti this work was presented to the Organisaton for Economic Co-operation and Developments Nuclear Energy Agency OECDNEA Expert Group on Improvement of Integral Experiment Data for Minor This experiment is unique in the sense that it will provide a consistent set of neutron cross-sections in fast and epithermal neutron spectra for most isotopes of interest to reactor physics. GillesYouinou INL Principal Investigator Nuclear Science User Facilities 54 Actinides Management where it received very positive feedback. More information was provided at another OECDNEA meeting in October 2014. A status-of-the-experiment report was also presented at the ATR NSUF Users Meeting in June 2014. Future Activities Goals for 2015 include 1 finalizing the MC-ICP-MS measurements 2 finalizing ATR as-run calculations with FIGURE 2. Details of the ATR MCNP model. Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor Collaborators Idaho State University George Imel principal investigator Idaho National Laboratory GillesYouinou principal investigator Giuseppe Pino Palmiotti collaborator Tom Maddock collaborator Jeff Giglio collaboratorTheresa Giglio collaborator Jeff Berg collaborator Jim Sterbentz collaborator MCNP and 3 preparing samples for researchers to run complementary measurements using the AMS at the ATLAS facility. Publications and Presentations See additional publications from other years in the Media Library on the NSUF website. 2014 ANNUAL REPORT 55 FIGURE 3. Illustration of the cadmium-filtered irradiation set-up. This experiment will generate enough data to validate models and provide better neutron cross-sections for advanced reactor designers for years to come. Nuclear Science User Facilities 56 Robust materials are critical to meet evolving advanced reactor and fuel designs.These mate- rials need to operate in extreme envi- ronments of elevated temperatures corrosive media and high radiation fluences with lifetime expectation of greater than 60 years. Full under- standing of a materials response to irradiation is paramount to long-term reliable service.The layered ternary carbides and nitrides known as MAX phases have the potential to be used in the next-generation nuclear reactors. All MAX phases are fully machinable even though some of them such as Ti3SiC2 andTi2AlC2 are similar to titanium metal in density but are three times as stiff.The thermal and electrical conductivities are high and metal-like.They have relatively high fracture toughness values and some are chemically stable in corrosive environments.They also have shown irradiation damage tolerance in heavy ion studies. The aim of this project is to inves- tigate the damage in Ti3SiC2Ti3AlC2 and chemical vapor deposition CVD SiC for comparison after exposure to a spectrum of neutron irradiations consistent with conditions found in light water nuclear reactors. Advanced Damage-Tolerant Ceramics Candidates for Nuclear Structural Applications Michel W. Barsoum Drexel University barsoumwdrexel.edu Figure 1. a Brightfield TEM images taken near the 0001 zone axis of the ATR-Ti3SiC2 sample irradiated to 0.1 dpa at 100C reveals perturbation of the surface and long basal plane dislocations. The mottled surface was likely due to improper FIB cleaning. b TEM micrograph of ATR-Ti3SiC2 sample irradiated to 0.1 dpa at 500C taken in 2 beam condition near the 11-20 zone axis shows dislocation arrays parallel to the basal plane and stacking faults. Small defects can be seen throughout but were not confirmed as loops. c Brightfield TEM micrograph taken on the 11-20 zone of the ATR-Ti3SiC2 sample irradiated to 0.1 dpa at 1000C showing a preexisting TiC particle that was highly damaged - with dislocation loops and black spots - not present in the surrounding MAX matrix which remains clear of irradiation induced defects. 2014 ANNUAL REPORT 57 The facilities and capabilities available to me at CAES have been incredibly helpful in advancing my researchand I am looking forward to my future visits. Darin J.Tallman Ph.D.Candidate observed that in several instances the samples had either fused together or to the capsule making retrieval diffi- cult.The most troublesome capsules were the ones held at 1000C for the longest times.What is believed to have happened is that the materials swelled more than anticipated for the designed capsule volume.As many samples were recovered as possible and all samples have been separated and organized into storage vessels KGTs. Throughout FY 2014 DarinTallman scheduled several trips to INL for PIE characterization using equipment at CAES. Most capsules remain at HFEF where they are stored and await characterization. PIE characterization commenced in FY 2014 with capsules KGT1367 KGT1369 and KGT1371 corresponding to resistivity bars irradiated to 0.1 dpa at 100 500 and 1000C respectively.These were the most readily available capsules for decontamination and preparation as they had already been sent to EML for preliminary analysis. Samples were expertly cleaned and prepared by Karen Wright and Collin Knight at EML for shipment to CAES. Due to their high activity readings the resistivity bars were mounted in The carbides are exposed to a series of neutron fluence levels 0.1 1 and 9 dpa at moderate to high irradia- tion temperatures 100 500 and 1000C in the Advanced Test Reactor ATR at Idaho National Laboratory INL.The damage to the microstruc- tures and the effects of the radiation on the mechanical and electrical properties of the materials will be characterized during post-irradiation examinations.The results will provide an initial database that can be used to assess the microstructural responses and mechanical performances of these ternaries. Accomplishments This project is a collaborative effort between the INL Savannah River National Lab and Drexel University and was initiated in 2009.As noted above the team seeks to characterize the effect of neutron irradiation of select MAX phases for use in nuclear reactor applications. During the programs second year all irradiated samples were shipped to the PIE facility at INL. Unfortunately unavoid- able delays within INL throughout the FY 2013 continued to delay the project.The receipt cask unloading experiment disassembly and cata- loging of specimens have finally commenced led by Collin Knight. Upon opening the capsules it was Nuclear Science User Facilities 58 grains at 1000C Figure 1c. From Fig. 1c alone it is clear that the Si present inTi3SiC2 provides significant irradia- tion resistance compared to its binary counterpartTiC.TEM micrographs of theTi3AlC2 samples revealed stacking faults and possible dislocation loops at both 100 Figure 2a and 500C Figure 2b. More extensiveTEM work is necessary for these samples to collect high-resolutionTEM micrographs of the defect microstructures and confirm the presence of loops. Resistivity measurements were collected using a 4-pt probe technique while applying a constant current of 100 mA.Voltages were collected every 5 seconds for 10 minutes to reach steady state.The resistivity values show more than an order of magnitude increase after irradiation at 100C but epoxy to limit worker exposure and the resistivity jig was redesigned. During three-week trips in June and September Darin collected prelimi- naryTEM results Figures 1 and 2 and resistivity measurements Figure 3 forTi3SiC2 andTi3AlC2 from these three capsules.Ti3AlC2 irradiated at 1000C did not survive extraction and is absent from these results. FIB prep for these initial samples was poorly executed and Darin has since been improving on his techniques. TEM micrographs ofTi3SiC2 reveal 1 black spots within the basal plane at 100C Fig. 1a 2 dislocation arrays and stacking faults at 500C Fig 1b and 3 highly damaged preexisting TiC grains next to mostly cleanTi3SiC2 Figure 2. a Brightfield TEM image near the 11-20 zone axis of ATR-Ti3AlC2 sample irradiated to 0.1 dpa at 100C reveals basal dislocations throughout the sample some of which may possibly be short dislocation loops. b TEM micrograph of same sample irradiated to 0.1 dpa at 500C taken in 2 beam condition tilted away from the 11-20 zone axis shows dislocation arrays parallel to the basal planes as well as stacking faults. Several of the dislocations do not appear straight due to the tilt of the sample they are most likely curved within the basal planes. Further TEM investigation is needed to confirm the presence of dislocation loops or other irradiation induced defects within these materials. The MAX phases a class of machinable layered ternary carbides and nitrides have great promise for use in the next-generation of nuclear reactors. This is the first time the MAX phases have been neutron irradiated at temperatures as high as those carried out here. 2014 ANNUAL REPORT 59 after irradiation at 500C the resis- tivity values recovered to values closer to their values for the pristine samples Fig. 3.These results are consis- tent with those obtained from our Drexel-MITR NEUP project recently published in Acta Materialia. The 9 dpa samples were deemed as high-priority samples to see signs of irradiation damage. However unrelated laboratory shutdowns led to further delays throughout FY 2014 including limitations on sample ship- ping within the INL complex.The 9 dpa samples were also mostly broken during retrieval and were unable to be separated and identified easily. Resources and funding were directed toward the readily available capsules. These issues prevented the high-dose samples from being available for characterization at CAES in 2014. With the extensive work stoppages preventing ATR sample access at CAES several TEM samples from our parallel Drexel-MITR NEUP irradiation project were shipped to INL for PIE utilizing ATR funds while Darin was onsite.With the excellent assistance of Lingfeng He a recently hired research scientist at INL MITR samples of fine-grained Ti3SiC2 and Ti2AlC were characterized. Results from this work are being prepared for publication in Acta Materialia which revealed the formation of black spots and defect clusters in Ti2AlC after irradiation to 0.1 and 0.4 dpa at 360C Figs. 4a and b and dislocation loops and stacking fault formation in both Ti3SiC2 and Ti2AlC Figure 3. a Room temperature resistivity plotted on a log scale as a function of neutron irradiation temperature at 0.1 dpa. After irradiation at 100C the resistivities of both Ti3SiC2 and Ti3AlC2 increase more than an order of magnitude. After irradiation at 500 and 1000C the resistivity of the Ti3SiC2 recovered to values near those of the pristine values. Some recovery is observed in Ti3AlC2 with only a 2-fold increase after irradiation at 500C. Nuclear Science User Facilities 60 after irradiation up to 0.1 dpa at 700C Fig. 4c and d respectively. Characterization of both sample sets is ongoing. Research to be completed Based on the results obtained thus far with the MAX phase samples irradi- ated at the MIT reactor it is anticipated that most of the irradiation damage would have been rapidly annealed out at 1000C. Comparison with other temperatures is needed to explore the irradiation behavior of these materials. Results to date suggest that Ti3SiC2 remains a strong candidate for irradiation application. DarinTallman is collaborating with Lingfeng He at INL to continue theTEM characteriza- tion of both the ATR and MITR sample sets. Darin anticipates graduation in the summer of 2015 thus remaining PIE trips are limited.The high dose 9 dpaATR samples are anticipated to be available by April of 2015 and will be the focus of future characterization. XRD diffractograms of the samples will be obtained once the remaining samples are removed from the capsules and shipping of samples is allowed within INL. Publications and Presentations 1. D. J.Tallman E. N. Hoffman E. N. Caspi B. L. Garcia-Diaz G. Kohse R. L. Sindelar M.W. Barsoum. 2014 Effect of Neutron Irra- diation on Mn1AXn Phases ICACC14Daytona BeachFLJanuary 292014. 2. D. J.Tallman E. N. Hoffman E. N. Caspi B. L. Garcia-Diaz G. Kohse R. L. Sindelar M.W. Barsoum 2014 Effect of Neutron Irra- diation on Mn1AXn phases TMS2014San DiegoCAFebruary 17 2014. 3. D. J.Tallman E. N. Hoffman E. N. Caspi B. L. Garcia-Diaz G. Kohse R. L. Sindelar M.W. Barsoum 2014 Effect of Neutron Irradia- tion on Select Mn1AXn phases CIMTECMontecatiniTermeItalyJune 112014. 4. D. J.Tallman E. N. Hoffman E. N. Caspi B. L. Garcia-Diaz G. Kohse R. L. Sindelar M.W. Barsoum 2015 Effect of neutron irradia- tion on select MAX phases Acta MaterialiaVol.85pp.132143. Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Idaho National Laboratory AdvancedTest Reactor Hot Fuel Examination FacilityAnalytical Laboratory or Electron Microscopy Laboratory Irradiation Assisted Stress Corrosion Cracking Facility Massachusetts Institute ofTechnology Reactor Collaborators Drexel University Darin J.Tallman collaborator Idaho National Laboratory Lingfeng He collaborator Savannah River National Laboratory Elizabeth N. Hoffman collaborator Brenda L. Garcia-Diaz collaborator 2014 ANNUAL REPORT 61 Figure 4 a Brightfield high resolution transmission electron microscopy HRTEM micrograph of an edge on grain of a MITR-Ti2AlC sample irradiated to 0.1 dpa at 360C reveals perturbation of the basal planes and the presence of defect clusters andor black spots circled. b TEM micrograph of a MITR-Ti2AlC sample irradiated to 0.4 dpa at 360C taken in 2 beam condition near the 0001 zone axis shows a higher density of small black spots circled. c Brightfield TEM micrograph of a fine-grained MITR-Ti3SiC2 sample irradiated to 0.1 dpa at 695C shows dislocation loops imaged near the 11 -20 zone axis parallel to the basal plane agglomerating near a stacking fault black arrow.The loops are seen as black and white lobes due to strain induced in the surrounding lattice with an average defect density of 3.94 x 1021 loopsm3. d Brightfield TEM micrographs of MITR-Ti2AlC irradiated to 0.1 dpa at 695C taken near the 11 -20 zone axis reveal dislocation loops edge on within the basal planes with a higher average density of 1.1 x 1023 loopsm 3 . Nuclear Science User Facilities 62 Low fluence experiments in metallic fuels specifically uranium-zirconium U-Zr and uranium-molybdenum U-Mo types have a relevance to both the Advanced Fuel Cycle Initiative AFCI and the Reduced Enrichment for Research andTest Reactors RERTR program. Quantified findings for the low fluence behavior of metallic fuels will help us understand fuel performance in thermal and irradiation fields. Project Description The objectives of this program are to understand 1 the microstructural evolution of these fuels as a func- tion of temperature fluence and composition and 2 diffusion-related phenomena in the fuels and cladding also as a function of temperature fluence and composition. Findings from these experiments will explain early microstructural development and mechanisms in detail as well as provide critical data for models under development in both programs. Near-term critical results from this project will support AFCI modeling work on constituent redistribution in irradiated uranium-prasecoymium- zirconium U-Pr-Zr fuels that is currently being undertaken by collaborators on this team.These results will also be used to improve the accuracy of computer models in the RERTR program that predict the overall swelling behavior of the fuel. Accomplishments The design of these experiments along with the necessary quality control documents from University of Central Florida UCF researchers and INL scientists was finalized at a meeting at UCF during FY 2014. Meanwhile UCF researchers have continued to pave the way to a better understanding of thermal behavior without irradiation see graphics. Published technical accomplishments from this indepen- dent work are listed below in Publica- tions and Presentations. Future Activities During 2015 work on specimen preparation with alloy casting is planned at INL. Upon receiving the alloys UCF will produce the samples to be inserted into the ATR. UCF will also continue to document the thermal behavior of alloys and diffu- sion couples so that upon completion of ATR experiments the effects of radiation can be elucidated. Low Fluence Behavior of Metallic Fuels Yongho Sohn University of Central Florida UCF yongho.sohnucf.edu Quantified findings of low fluence behavior in metallic fuels will help us understand fuel performance under thermal and irradiation fields. 2014 ANNUAL REPORT 63 Figure 1. High angle annular dark field STEM micrographs from the water-quenched U-10 wt. Mo vs. Zr diffusion couple annealed at 650C for 720 hours. The three squares indicate sampling positions of the 1 -Zr 2 Mo2Zr -U and -Zr and 3 U6Zr3Mo by selected area electron diffraction and high-resolution TEM with fast Fourier transformation analyses. Nuclear Science User Facilities 64 Publications and Presentations 1. Y. Park D.D. Keiser Jr.Y.H. Sohn 2015 Interdiffusion and reac- tion between UMo And Zr at 650C as a function of time Journal of Nuclear MaterialsVol. 456 pp. 351-358. 2. Y. Park K. HuangA. Paz y Puente H.S. Lee B.H. Sencer J.R. Kennedy Y.H. Sohn 2015 Diffusional Interaction Between U10wt.Zr and Fe at 903K 923K and 953K 630C 650C and 680C Metallurgical and MaterialsTransactions A Vol. 46A pp. 7282. 3. K. HuangY. Park L. Zhou K.R. CoffeyY.H. Sohn B.H. Sencer J.R. Kennedy 2014 Effects of Cr and Ni on interdiffusion and reaction between U and FeCrNi alloys Journal of Nuclear MaterialsVol. 451 pp. 372378. 4. A. Paz y Puente J. Dickson D.D. Keiser Jr.Y.H. Sohn 2014 Investigation of interdiffusion behavior in the MoZr binary system via diffusion couple stud- ies International Journal of Refractory Metals and Hard MaterialsVol. 43 pp. 317321. 5. J. Dickson L. ZhouA. Paz y Puente M. Fu D.D. Keiser Jr. Y.H. Sohn 2014 Interdiffusion and reaction between Zr And Al Al-2wt.SiAl-5wt.Si or 6061 from 425 to 625C Intermetallics Vol. 49 pp. 154162. 6. Y. Park J.Yoo K. Huang D.D. Keiser Jr. J.F. Jue B. Rabin G. MooreY.H. Sohn 2014 Growth kinetics and microstructural evo- lution during hot isostatic press- ing of U-10wt.Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier Journal of Nuclear MaterialsVol. 447 pp. 215224. 7. K. Huang C. Kammerer D.D. Keiser Jr.Y.H. Sohn 2014 Dif- fusion Barrier Selection from Refractory Metals Zr Mo and NbVia Interdiffusion Investiga- tion For U-Mo RERTR Fuel Alloy Journal of Phase Equilibria and Diffusion Vol. 35 pp. 146156. See additional publications from other years in the Media Library on the NSUF website. My interaction has been truly rewardinggiving me an opportunity to strengthen my fundamentals of sciencehone my hands-on laboratory skills and gain greater perspective on energy research with respect to application and societal impact.It is a great program in which graduate students can cover the entire spectrum of RD. Ashley Paz y Puente formerly Ashley Ewhgraduate student University of Central Florida 2014 ANNUAL REPORT 65 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor Collaborators University of Central Florida Yongho Sohn principal investigatorYoungjoo Park graduate student Ryan Newell graduate student Esin Geller graduate student Nicholas Eriksson graduate student Felipe Betanco undergraduate student Idaho National Laboratory Maria Okuniewski co-principal investigator Dennis D. Keiser Jr. co-principal investigator Nuclear Science User Facilities 66 Radiation-Induced SegregationDepletion at Grain Boundaries in Neutron-Irradiated 304SS at Low Dose Rates Emmanuelle Marquis University of Michigan emarqumich.edu Radiation-induced segregation depletion and its deleterious impact on properties in austenitic stainless steels have been studied extensively particularly at high dose rates typically 10-4dpas. However validation of life-extension plans for light water reactors and future applications in advanced fission and fusion reactors require input data on the effects of high fluence obtained at low dose rates on the microstruc- ture and mechanical properties of these steels. Project Description Irradiating the 304 stainless steel 304SS hex-blocks in the Experi- mental Breeder Reactor EBR-II fast reactor to relatively high doses at low dose rates provides an opportunity to investigate the irradiation-induced microstructures.This was performed using transmission electron micros- copy TEM and atom probe tomog- raphy APT both at CAES and at the University of Michigan.The objectives of the work were to Understand the synergy or competi- tion between radiation-induced segregation carbide formation and swelling as function of dose rate dose and temperature. Understand the microstructural changes induced near the sodium- wetted surfaces and their con- sequences on radiation-induced segregation at grain boundaries. Benchmark the APT measurements by comparing them to informa- tion obtained by analytical electron microscopy. Develop a mechanistic under- standing for the observed changes. Accomplishments After a delayed start of over two years due to technical issues regarding materials selection availability and preparation the project progressed significantly thanks to the involvement of Bulent Sencer at INL.The micro- structures of two hex-blocks irradiated at different dose rates and doses were characterized in great detail including voids nickel Ni-silicon Si clusters dislocation loop density loop chem- istry phosphide precipitates grain boundary chemistry grain boundary carbides and surface chemistry. Selected examples of these observa- tions are illustrated in Figure 1. Future Activities All the objectives have now been addressed. University of Michigan graduate studentYan Dong will be presenting the results atThe Minerals Metals and Materials Science TMS conference in 2015 and two publica- tions are in preparation for submission to the Journal of Nuclear Materials. Within the light water reactor susceptibility program understanding the microstructures developing in austenitic stainless steels under very low dose rates is essential to ensuring reliable predictions. 2014 ANNUAL REPORT 67 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators University of Michigan Emmanuelle Marquis principal investigator Yan Dong collaborator Idaho National Laboratory Bulent Sencer collaborator Consultant Frank Garner Figure 1. 3D reconstructions highlighting a the network of dislocation loops segregated with Si and Ni as indicated by the concentration profile underneath. b Phosphide plates with Ni and Si segregating to the plate interface and c small carbide and Ni3Si precipitates with Ni and Si segregating to the carbide interface. Nuclear Science User Facilities 68 Multiscale Investigation of the Influence of Grain Boundary Character on Radiation-induced Segregation and Mechanical Behavior in Steels Used in Light Water Reactors Mitra Taheri Drexel University mtahericoe.drexel.edu Project Description This project is centered on under- standing the behavior of 304 and 316 stainless steels 304SS 316SS in light water reactors LWR.The primary objective is to understand and quantify any microstructural evolution taking place related to long-term aging thermal effects and any irradiation-induced or enhanced solute segregation and precipitation occurring in-grain or at the grain boundaries. A second objective is to understand the role of grain boundary character in low-stacking-fault face-centered- cubic fcc stainless steels including any dependence they exhibit on irradiation-enhanced or -induced grain boundary solute segregation under long-term irradiation. The research provides through advanced electron and atom probe microscopy direct quantification of grain boundary character- dependent stainless steel irradiation responses during long-term aging in current LWRs. Accomplishments Drexel Universitys technical objective in the project was to investigate the role of grain boundary character in radiation-induced segregation and precipitation in fcc austenitic stain- less steel.A 304SS from the Experi- mental Breeder Reactor II EBR-II outer-blanket assembly Row 13 was previously irradiated to a peak fluence of 4.5x1021 ncm2 . High-quality electro-polished bulk samples were prepared at INLs Hot Fuel Examina- tion Facility HFEF for end-user examination at CAES using advanced microscopy techniques.These tech- niques included electron backscatter diffraction EBSD transmission electron microscopy bright field TEM-BF scanning transmission elec- tron microscopy energy dispersive x-ray spectroscopy STEM-EDS and atom probe tomography APT all of which are capable of probing any grain boundary character dependence on multiple-length scales from mesoscale grain boundary structure to atomic scale solute segregation. The 304SS U1302 sample was a hexagonal duct strip from the outer-blanket assembly irradiated The research provides direct quantification of the responses of grain boundary character- dependent stainless steel under LWR irradiation. 2014 ANNUAL REPORT 69 Figure 1. STEM and STEM-EDS results from a twin 3 grain boundary highlighting characteristic Cr rich M23C6 carbides along the examined GB length. Figure 2. Highlight of EBSD map indicating a 45-degree grain boundary that is then thinned for a TEM and APT correlated study. Nuclear Science User Facilities 70 in Row 13 Position 13D4. Over the course of the EBR-II irradiations Run 1 to Run 170B the 304SS was exposed to extended periods of high operating temperatures 450460C. Researchers analyzed samples using a using multi-length scale grain-boundary site-specific process optimized to obtain detailed chemical and structural information. Specifically they performed EBSD to determine particular grain boundary misorientations and used an FIB to extract specific grain boundaries. These samples were analyzed using bothTEM and APT. The study focused on the following grain boundary characters coinci- dence site lattice CSL three coherent twin three incoherent twin and random high-angle misorientation grain boundaries. Researchers focused on the three system due to the differ- ences observed between the atomic structures of the coherent twins having a 111 symmetric tilt grain boundary plane and the incoherent twin having a 112 symmetric tilt grain boundary plane.This grain boundary examination system allows the careful study of the effects of the grain boundary plane inclination angle with respect to irradiation- induced precipitation and segregation. The goal was accomplished and all three grain boundary types were examined using both APT and STEM- EDS.The results indicated that each grain boundary type had extensive carbide M23 C6 precipitation.The presence of carbide nucleation and growth on the coherent twin grain boundary Figure 1 highlights the fact that even low grain-boundary- energy and high atomic-fit grain boundary characters can be susceptible to carbide growth when subjected to extensive irradiation at relatively high temperatures 450460C.The APT examination showed that depletion of Cr to less than 10 at. existed in regions adjacent to the grain boundary carbide.These Cr-depleted zones are typical of grain boundary carbide formations and indicate regions that are potentially susceptible to increases in localized corrosion of aging stain- less steel alloys. The second sample condition the project focused on was a 316 hex-duct stainless steel from location U9027 Row 9 in the EBR-II. Data collection on the exact reactor temperature and fluence is ongoing but they are esti- mated to be 370500C and 2025 displacements per atom dpa respectively. Unlike the 304SS detailed aboveAPT revealed that the 316SS had extensive nickel-silicon Ni-Si clus- ters throughout the examined regions. The grain-boundary microchemistry independent of grain-boundary char- acter showed areas of both carbides and extensive W-shaped Cr depletion. Figures 2-4 show the steps researchers performed on the grain-boundary structures at CAES to fully characterize a particular grain boundary EBSD hummingbird half-grid TEM holder STEM andTEM atom probe. Overall like the 304SS the 316SS has grain-boundary-depleted regions of Cr and other minor solute elements. Unlike the examined outer blanket assembly in 304SS the 316SS has extensive Ni-Si clusters at levels consistent with the presence of Ni3Si precipitates andor possible Si enrich- ment in dislocation arrays. Future Activities The project was completed in 2014. Figure 3. Sequence of three BF-TEM diffraction conditions A-C and STEM-HAADF D of the same GB structure shown in Figure 2. 2014 ANNUAL REPORT 71 Figure 4. A Sequence of atom probe tomography atom map for Fe Cr Ni Mo B highlighted GB microchemistry including Cr depletion with W profile and C concentration profile for Ni-Si enriched clusters. Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory PIE facilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators Drexel University MitraTaheri principal investigator Christopher Barr co-principal inverstigaator Idaho National Laboratory Jim Cole principal investigator Nuclear Science User Facilities 72 Transducers for In-Pile Ultrasonic Measurement of the Evolution of Fuels and Materials Bernhard Tittmann Pennsylvania State University brt4psu.edu Project Description Pennsylvania State University PSU was awarded anATR NSUF program to insert both magnetostrictive and piezoelectric transducers into the Massachusetts Institute ofTechnology Research Reactor MITR for radiation at fluences up to 1021 ncm2 . Ultrasonic measurements of the transducers are being taken in-situ as the transducers are subjected to the neutron flux.The goals of the experiment are to develop a test design including selection criteria for candidate materials and opti- mizing test assembly parameters obtain data from both out-of-pile and in-pile tests conducted at elevated temperatures and compare the accrued data with the expected performance of ultrasonic devices under irradiation conditions. These experiments will enable researchers to design ultrasonic sensors for characterizing the evolution of fuel materials and other nuclear reactor components.The sensors will allow in-situ measurements of material prop- erties and will lead to the development of tools that will monitor the structural health of reactors. Accomplishments In 2014 the following progress was made on the project. The following piezoelectric transducer materials were selected aluminum nitride AlN zinc oxide ZnO and bismuth titanate BiT. Remendur and Galfenol were selected as magnetostric- tive transducers. Assemblies for the magnetostrictive transducers were designed built and tested Figures 1 and 2.Transducer performance was characterized at design temperatures Figures 35. Ground lead weld to cable sheath Power lead weld to cable center conductor Figure 1. Magnetostrictive transducer fabrication. a Silver-palladium is wrapped around an alumina bobbin b the wire is coated with a standoff insulation and heat treated c the wire is coated with an alumina cement and heat-treated a second time d leads are laser-welded to the coax cable and e welded into a pre-fabricated housing. 2014 ANNUAL REPORT 73 The ZnO transducer did not perform as expected and will only operate successfully during reactor shutdowns Figure 5. One of ZnO transducers was replaced with an AlN transducer which operated successfully at the design temperature. Graduate students Brian Reinhardt Ph.D. and Andy Suprock M.S. were trained on high-temperature ultrasonic nondestructive testing transducer fabrication principles and the design of high-temperature radiation-tolerant transducers. Results of first power cycle The first power cycle commenced when the reactor had been operating for 180 hours and lasted until 980 hours of operation. During this time the reactor was operating at 5 MW. To allow the sensor temperature to rise slowly the reactor power had been brought to 5 MW incrementally over a period of one week Figure 6. The ZnO transducer experienced an electrical malfunction during the inser- tion process. One of theAlN transducers developed an electrical short during reactor startup when the temperature reached 400C. During the first power cycleA-scans or amplitude time serieswere periodi- cally collected from the remainingAlN and BiT transducers as well as from the two magnetostrictive transducers.The pulse echo-amplitude was determined by windowing the first returned echo and measuring the amplitude of the fundamental frequency component Figure 2. Piezoelectric transducer fabrication. Top Transducer components from left to right stainless steel cap alumina insulation nickel plunger alumina insulation carbon-carbon backing Kovar waveguide with ZnO sensor on top stainless steel outer casing. Bottom A fully assembled transducer housing with cable and strain-relief sleeve to support the cable connection. Figure 3. AlN ultrasonic pulse-echo amplitudes recorded at room temperature and at the irradiation temperature. Nuclear Science User Facilities 74 Figure 4. Temperature-dependent performance of an AlN piezoelectric transducer. The transducers operated at their highest efficiencies at the design temperature. Figure 5. Temperature-dependent performance of the ZnO transducer. The transducer did not perform well at the design temperature however it recovered its original amplitude upon cooling. This indicates that the sensor will operate satisfactorily during reactor shutdown enabling us to measure performance degradation at those times. using fast Fourier transform FFT. The results of this analysis are shown in Figures 79. Figure 7 shows the pulse-echo amplitude measured from the survivingAlN sensor.The plot was normalized to the pre-irradiation ampli- tude.The pulse-echo amplitude varies by - 20 during irradiation. Figure 8 shows the pulse-echo ampli- tude measured from the BiT sensor. During the first cycle the pulse-echo amplitude decreased by approximately 65. Figure 9 shows the pulse-echo amplitude of the magnetostrictive trans- ducers.These plots were also normalized to the pre-irradiated pulse-echo ampli- tude.The transient behavior in these sensors seems to be less pronounced as the pulse-echo amplitude only varies by about - 10. To gauge the materials practical use in harsh radiation environments the selec- tion criteria of piezoelectric materials The in-pile use of ultrasonic transducers during irradiations at MITR is extremely important because they could provide more accurate higher-resolution data on the performance of candidate fuels and materials exposed to the harsh conditions of irradiation testing. for nondestructive evaluation NDE and material characterization were analyzed. PiezoelectricAlN was observed to be a viable candidate material.Test results on transducers based onAlN BiT Remendur and Galfenol operating in a nuclear reactor within a 40-day window at a fast-neutron flux of 4.05 1013 ncm2 and a gamma dose rate of 1 109 rhr were also evaluated. In each case clearA-Scan measurements were taken at the end of the power cycle. Remendur Galfenol andAlN seemed to maintain their initial transduction efficiencies.The pulse-echo amplitude of theAlN sensors varied by - 20 while that of the Remendur and Galfenol varied by 10. Conversely by the end of the first power cycle the BiT pulse-echo amplitude had decreased by 65. The data shows both piezoelectric and magnetostrictive transducers hold promise for use in high-neutron-flux environments. Each shows potential for improving reactor safety and furthering the understanding of the effects of radiation on materials by enabling researchers to monitor a materials structural health and NDE even in the 2014 ANNUAL REPORT 75 These experiments will enable researchers to design ultrasonic sensors for characterizing the evolution of fuel materials and other nuclear reactor components. The sensors will allow in-situ measurements of material properties and will lead to tools for structural health monitoring. BernhardTittmannSchell Professor and Professor of Engi- neering Science and Mechanics Pennsylvania State University Figure 6. The blue curve indicates the integrated neutron flux fluence for neutrons with energy greater than 1 MeV. The green curve represents the reactor power. Figure 7. A plot of the pulse-echo amplitude measured by the remaining AlN transducer. The amplitude was normalized to the first measurement made at 0 fluence. The plotted data red applies only to the first power cycle and is comparable to the data collected by Parks and Tittmann 2014 black. Nuclear Science User Facilities 76 presence of high levels of radiation and high temperatures that would destroy typical commercial ultrasonic transducers. Future Activities During the second year of the experi- ment PSU personnel shipped signal- processing equipment to MITR and then travelled to MITR to assist with the insertion and setup of the signal- processing equipment.The irradiation started in February 2014. PSU is moni- toring and interpreting the data from the irradiation including supporting laboratory evaluations. Figure 8. Pulse-echo amplitude measured for the BiT transducer during the first power cycle. The projects third year will witness the completion of 18 months of irradiation. PSU and INL researchers will travel to MITR to assist with removal of samples from the reactor final measurements and post-irradiation examination PIE during which the effects of irradiation on the piezoelectric candidate materials will be quantified. System components will be examined at the MIT Hot Box and specific parameters including ultrasonic velocity and attenuation transduction efficiency of piezoelectric electrical resistivity color crystallinity and physical and electrical robustness will be evaluated. The scope of the PIE will be based on the activation level of the transducers. It is likely that the coaxial cables can be left intact as the experiment is moved to the MIT Hot Box. If this is the case it will be possible to reconnect these cables to non-irradiated cables and interrogate the transducers after they have been placed in the Hot Box. Detailed analysis of the magnetostrictive transducer material will probably not be possible due to the cobalt-bearing materials with which they are made. At the conclusion of the PIE activi- ties PSU scientists will prepare a final report summarizing the results of this project and MIT will dispose of the irradiated materials. Publications and Presentations 1. J. Daw J. Palmer P. Ramuhalli P. Keller R. Montgomery H-T. Chien B.Tittmann B. Reinhardt G. Kohse J. Rempe 2015 UltrasonicTrans- ducer IrradiationTest Results. NPIC HMIT at CharlotteNCFebruary 23262015. 2. B. Reinhardt B.Tittmann J. Rempe J. Daw G. Kohse D. Carpenter M. AmesY. Ostrovsky P. Ramuhalli R. Montgomery H.T. Chien and B.Wernsman 2014Progress to- wards developing neutron tolerant magnetostrictive and piezoelectric transducers 41st Annual Review of Prog- ress in Quantitative Nondestructive Evaluation ConferenceBoiseIDJuly 20252014. 3. D.A. Parks and B.R.Tittmann 2014 Radiation tolerance of piezolec- tric bulk single crystal aluminum nitride IEEETransactions on Ultrasonics Ferroelectrics and Frequency ControlVol. 61 No 7 pp. 12161222. See additional publications from other years in the Media Library on the NSUF website. 2014 ANNUAL REPORT 77 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Massachusetts Institute ofTechnology Research Reactor PIE facilities Idaho National Laboratory PIE facilities Collaborators Pennsylvania State University BernhardTittmann principal investigator Brian Reinhardt graduate studentAndy Suprock graduate student Idaho National Laboratory Joy Rempe co-principal investigator Joshua Daw collaborator Joseph Palmer collaborator Massachusetts Institute ofTechnology Gordon Kohse collaborator Pacific Northwest National Laboratory Pradeep Ramuhalli collaborator Argonne National Laboratory H.T. Chien collaborator Bettis Atomic Power Laboratory Ben Wernsman collaborator Figure 9. Pulse- echo amplitude of the Galfenol and Remendur magnetostrictive transducers. Nuclear Science User Facilities 78 Electron Backscatter Diffraction and Atom Probe Tomography to Study Krypton Segregation Behavior in Uranium Dioxide Michele Manuel University of Florida mmanuelmse.ufl.edu Project Description The objective of this research was to investigate the structure of a uranium oxide thin film produced by pulsed dc magnetron sputtering. This technique allows films of specific microstructures to be grown for analysis. Uranium dioxide is a pervasive material in nuclear energy and the ability to produce defined microstructures for testing makes this an exciting technique espe- cially given the difficulty in sample preparation of nuclear fuel.This specific project was to conduct atom probe tomography on the inter- face between the yttria-stabilized zirconia substrate and UO2 thin film to look for possible inter-diffusion that had occurred during processing. This research supports efforts to make the microstructural investiga- tion and testing of nuclear fuels more insightful and cost-effective. Accomplishments Samples were fabricated on the focused ion beam at the Center for Advanced Energy Studies CAES in Idaho Falls Idaho and then analyzed with atom probe tomorgraphy.The goal of the project to visualize the interface between film and substrate was successful and incorporated into a publication in the journal Applied Surface Science.This research was The ability to produce specific fuel microstructures by Pulsed DC Magnetron Sputtering is a powerful sample preparation technique and the research conducted at CAES pushes this method forward to full validation. Figure 1. Representative EBSD maps of 0.7 MeV and 1.8 MeV Kr-irradiated UO2 and the following post-irradiation anneal. 2014 ANNUAL REPORT 79 Access to the CAES facility has provided unprecedented insight into the behavior of nuclear fuels. Michele Manuel Associate Professor Department of Materials Science and Engineering University of Florida conducted primarily by Billy Valderrama and facilitated in large part by staff at CAES including Jatuporn Burns Dr.Yaqaio Wu and Joanna Taylor. Research to be completed This research is complete as it was a single study of one material for a publication. Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators Idaho National Laboratory Jian Gan collaborator University of Florida Michele Manuel principal investigator Billy Valderrama collaborator Hunter Henderson collaborator Figure 2. Kr concentration across a high angle grain boundary in UO2 annealed at 1600C showing significant segregation. The horizontal dashed line represents the bulk composition of Kr in the material. Publications and Presentations 1. Lin I Dahan B Valderrama M. V. Manuel 2014 Structure and properties of uranium oxide thin films deposited by pulsed dc magnetron sputtering Applied Surface ScienceVol. 301 pp. 475480. Nuclear Science User Facilities 80 Irradiation Effects in Aged Cast Duplex Stainless Steels YongYang University of Florida yongyangufl.edu By correlating microstructural studies with associated mechan- ical tests this research provides a fundamental understanding of the aging behavior of cast stainless steels during the 40 years of as-designed reactor life and more importantly during the 60 years and beyond of a reactors anticipated extended life. Project Description Researchers used transmission electron microscopy TEM and atom probe tomography APT to systemically characterize the neutron-irradiated cast stainless steels under non-aged and aged conditions focusing on precipita- tions and elemental segregation.The results of the study will provide some of the first knowledge of the synergistic effects of thermal aging and neutron irradiation on microstructural evolu- tion in reactor components made of cast stainless steel. Accomplishments Microstructural changes in the ferrite atoms of thermally aged only neutron- irradiated only and neutron-irradiated after thermal aging cast austenitic stainless steels CASS were investigated using APT. It was found that low-dose neutron irradiation could effectively induce spinodal decomposition in the ferrite atoms of non-aged CASS while the neutron irradiation of aged CASS would further enhance the spinodal decomposition Figure 1. Determining the combined effects of thermal aging and low-dose neutron irradiation on duplex stainless steels cast stainless and austenitic stainless steel welds is going to have major implications for the sustainability of light water reactors LWR when their service lives are extended to 80 years. Figure 1. Spinodal decomposition of ferrite phase in cast stainless steels where the red color represents Fe atoms and the blue color is the background. Aged As-cast Aged IrradiatedAs-cast Irradiated 2014 ANNUAL REPORT 81 Neutron irradiation not only dramati- cally increases the size of G-phase precipitates in the stainless steel samples but increases the amount of chromium Cr phosphorous P and molybdenum Mo and decreases the iron Fe and manganese Mn Figure 2.This study proves that the effects of neutron irradiation on ferrite degradation are highly dependent on the irradiation dose rate at an LWRs operational temperatures. Researchers postulate that a synergistic effect exists in duplex stainless steel Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators University of Florida YongYang principal investigatorWei-yang Lo graduate student collaborator Zhangbo Li graduate student collaborator Argonne National Laboratory Yiren Chen co-principal investigator University of Wisconsin - Madison Janne Pakarinen co-principal investigator Aged Aged IrradiatedAs-cast Irradiated Figure 2. G-phase precipitates in aged cast-irradiated and aged-irradiated CF-3. Image sizes reflect identical scales and the blue dots represent Cr atoms. The concentrations of the iso-surface plots of Mn nickel Ni and silicon Si are 5 20 and 15 at. respectively. Understanding the effects of long-term thermal aging and neutron irradiation on duplex stainless steels in LWRs will provide a strong scientific basis for managing LWRs as they age beyond their designed-for lifetimes. YongYangAssistant Professor Nuclear EngineeringUniversity of Florida components between thermal aging and neutron irradiation even at very low doses. Future Activities The project was completed in 2014. Publications and Presentations 1. W-Y LoY. Chen J. PakarinenY. WuT. Allen andY.Yang Irra- diation response of delta ferrite in as-cast and thermally aged cast stainless steel submitted to Journal of Nuclear Materials 2014 under revision. Nuclear Science User Facilities 82 Scanning Transmission Electron MicroscopyLocal Electrode Atom Probe Study of Fission Product Transportation in Neutron Irradiated Tri-structural Isotropic Fuel Particles Izabela Szlufarska University of Wisconsin izabelaengr.wisc.edu Tristructural-isotropic TRISO- coated particle fuel consisting of a spherical fuel kernel encapsulated in successive layers of pyrolytic graphite PyC and silicon- carbide SiC is being considered as fuel for both the Generation IV Very High Temperature Reactor VHTR and the Fluoride Salt-Cooled High-Temperature Reactor FHR. The SiC layer is intended to be the primary barrier to the release of radioactive metallic fission products FP. However the release of FP particularly silver Ag from seem- ingly intact TRISO particles has been observed.Therefore understanding the mechanisms of FP Ag transport through SiC is necessary to limit its release and consequently promote the safe operation of the reactor. Because research has been limited by the inability of nanoscale analysis equipment to handle irradiated mate- rials little is known about how FPs are transported through the SiC layer. In Figure 1. Distribution of FP precipitates in the SiC layer 5 m from SiCIPOyC interface of a neutron- irradiated TRISO fuel particle from AGR-1 experiments. 2014 ANNUAL REPORT 83 factAg was first identified in the SiC layer of a neutron-irradiatedTRISO fuel particle only in 2013 by a Boise State University research team at INL using scanning transmission electron microscopy STEM 1. Project Description As a continuation of the 2013 preliminary STEM study this project utilized STEM and the local elec- trode atom probe LEAP to further investigate the morphology composi- tion and distribution of FPs in the SiC layer of the same neutron-irradiated TRISO fuel particles.The goal was to gain further insight into the FPs transport mechanism particularly that of Ag.The various fission prod- ucts found have improved our under- standing of FP release mechanisms in this type of fuel. Accomplishments STEM-EDS AnalysisTwo transmission electron microscopy TEM lamellae lifted from the SiC layer close to the SiCinnerPyC interface were examined in detail using STEM and energy dispersive spectroscopy EDS.The excellent Z contrast observed under STEM clearly revealed the existence of various types of FP precipitates Figure 1 and EDS aided in the determination of the composition of these precipitates.This study confirmed previous reports of Ag-enriched grain boundary and triple-junction precipitates and identified for the first time the Ag-palladium Pd intragranular precipitates as well as a network of small uranium-rich precipitates.These findings advance our understanding of FP transport through the SiC layer in TRISO fuel particles. LEAP Analysis A batch of 15 LEAP tips from the regions where theTEM lamellae were fabricated and exam- ined in 2013 was prepared.The first phase of this work resulted in some fracturing of the tips primarily due to the weakness of theTEM sample grid on which the samples were placed. The largest data set obtained was only from depths of about 30 m. In the second phase of this study another batch of five tips was fabri- cated and tested in 2014.These tips were made from material taken from the same location as in the first phase but this time they were mounted in a much sturdier copper focused ion beam FIB grid.With the optimized equipment parameters laser energy of 80 to 100 petajoules pJ laser pulse of 160 to 200 kilohertz kHz temperature of 3060 K better grid data could be obtained from deeper in the material than was possible in the first phase of the project.The largest data set obtained is shown in Figure 2. Although the results were superior to those obtained from Phase one of this study the analysis depth was The cutting-edge analytical equipment in the Microscopy and Characterization Suite at CAES can shed new light on thermodynamically and kineti- cally driven nanoscale physical and compositional changes in materials. Dr.Bin LengResearch AssociateUniversity ofWisconsin - Madison Nuclear Science User Facilities 84 Figure 2. LEAP results showing the distribution of elements in a tip fabricated from a neutron-irradiated TRISO fuel particle. still limited and no FP were detected. Therefore in the third phase of this study ten tips were fabricated at the University ofWisconsin Madison from a surrogate Ag ion-implanted SiC sample and shipped to CAES for LEAP examinations. Compared with the neutron-irradiated samples the surrogate tips were far less prone to fracture and data was obtained from depths of up to 240 nm. However no Ag was detected Figure 3. It was concluded that LEAP examina- tion of SiC is limited by the relatively low electrical and thermal conductivity Various fission products found in the SiC layer of a neutron-irradiated TRISO fuel particle have improved our understanding of the fission product release mechanisms in this type of fuel. of this material as well as its brittle- ness which is exacerbated by radia- tion-induced defects. More extensive research is needed to exploit the full benefits of the LEAP technique for the examination of irradiated SiC samples. Future Activities This project has been completed. References 1. I. J. van RooyenY. Q.WuT. M. Lillo 2014 Identification of silver and palladium in irradi- atedTRISO coated particles of the AGR-1 experiment Journal of Nuclear Materials 446 pp. 178186. Publications and Presentations 1. I.van Rooyen B. LengY.WuT. Lillo I. Szlufarska K. Sridharan T. Gerczak J. Madden 2014 Identification of Fission Products in irradiated SiC Using Scanning Transmission Electron Microscopy and Atom ProbeTomography ATR NSUF UsersWeek. 2. B. Leng I. van RooyenY.Wu I. Szlufarska K. Sridharan 2015 STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment submitted to the Journal of Nuclear Materials. 2014 ANNUAL REPORT 85 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor Center for Advanced Energy Studies Microscopy and Characterization Suite University of Wisconsin Madison PIE facilities Collaborators University of Wisconsin Madison Izabela Szlufarska principal investigator Bin Leng collaboratorTyler Gerezak collaborator Kumar Sridharan collaborator Idaho National Laboratory Isabella van Rooyen principal investigator James Madden collaboratorTim Lillo collaborator Center for Advanced Energy Studies Yaqiao Wu collaborator Figure 3. LEAP results showing the distribution of elements in a tip fabricated from surrogate Ag ion- implanted SiC. Nuclear Science User Facilities 86 Irradiation-Assisted Diffusion of Uranium-Iron Diffusion Couples Lin Shao Texas AM University lshaotamu.edu Irradiation of nuclear fuel results in fuel swelling and the production and transport of fission products to the cladding. This contact may allow chemical and mechanical interactions with the cladding. Since the chemical compatibility between the fuel and the cladding is imperative for safe opera- tion of a reactor it is essential to limit the interdiffusion between the two. Project Description This project is investigating fuel- cladding interactions with prototypic material systems that have traditionally been difficult to study these interac- tions under extreme conditions. Accomplishments The objective of this research is to provide an understanding of thermally activated and irradiation-enhanced multicomponent-multiphase diffu- sion and microstructural evolution in transition and rare earth metals relevant to fuel-cladding interactions. Studies to date have validated equi- librium phases predicted by phase diagrams of uranium-iron U-Fe uranium-nickel U-Ni and uranium- zirconium U-Zr binary systems. In addition researchers observed unusual microstructures in several phases which were not reported before. Discrepancies with previous publica- tions were documented for crystalline structures and lattice parameters of certain phases. Integrated diffusion coefficients at different temperatures and their activation energies were extracted for each diffusion couple. In the U-Fe system researchers found that integrated diffusion coefficients for diffusion couples prepared using polycrystalline Fe and single-crystal- line Fe are largely different. Figure 1 shows that diffusion coefficients for the polycrystalline Fe sample are systematically larger and the diffusion activation energy is lower than that of the single crystalline Fe sample. This is because atoms can use grain boundaries as quicker diffusion paths to broaden the widths of interfacial phases.This finding also points to the necessity of using single crystal diffu- sion couples for better comparisons between modeling predictions and actual results. Figure 2a-d plot the U and Fe elemental distributions in U-Fe single crystal diffusion couples obtained at different annealing temperatures. Figure 2e-f shows typical back-scattered-electron BSE images of diffusion annealed at 625C and 650C respectively. Future Activities All the focused ion beam FIB time allocated for this project has been used. Researchers plan to write another proposal for continuing support. Publications and Presentations 1. T. ChenT. Smith J. Gigax D. Chen R. Balerio B. H. Sencer J. R. Kennedy L. Shao Diffusion ki- netics and grain boundary effects in interface reactions of UFe diffusion couples Journal of Nuclear Materialsin press. 2. L. Shao D. Chen C-C.Wei M. Martin X.WangY. Park E. Dein K. R. CoffeyY. Sohn B. H. Sencer and J. R. Kennedy 2015 Radia- tion effects on interface reactions of UFe UFeCr and U FeCrNi Journal of Nuclear Materials vol. 456 pp. 302. See additional publications from other years in the Media Library on the NSUF website. Since the chemical compatibility between the fuel and the cladding is imperative for safe operation of a reactor it is essential to limit the interdiffusion between the two. Figure 1. Comparisons of integrated diffusivities forming UFe2 and U6 Fe phases in U-Fe diffusion couples using polycrystalline Fe and single crystal Fe respectively. 2014 ANNUAL REPORT 87 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor Center for Advanced Energy Studies Microscopy and Characterization Suite University of Wisconsin Madison PIE facilities Collaborators University of Wisconsin Madison Mahima Gupta principal investigator Janne Parkarinen collaborator Idaho National Laboratory Todd Allen co-principal investigator Jian Gan collaborator Los Alamos National Laboratory Steve Conradson collaborator Figure 2. Concentration profiles obtained for diffusion couples U-Fe single crystal annealed at different temperatures a-d and typical BSE image at 625C and 650C diffusion couples showing formation of intermetallic crystals e-f. The Center for Advanced Energy Studies is second to noneand we are lucky to have access to it. Lin Shao Associate Professor Texas AM University Nuclear Science User Facilities 88 Toward an Understanding of the Effect of Dose Rate on the Irradiation Response of F-M Alloys Janelle P.Wharry Boise State University janellewharryboisestate.edu With their high strength resistance to thermal stresses dimensional stability and low activation ferritic- martensitic F-M alloys are leading candidates for cladding and structural components in fusion and advanced fission reactors. Throughout their service lifetimes F-M alloys are exposed to neutron irradiation at doses as high as several hundred displacements per atom dpa.To simulate the neutron damage produced at these high doses ions are often used in test reactors to irradiate F-M samples.The ions higher damage rates and minimal residual radioactivity allow for quicker more cost-effective experiments however increasing the irradiation dose rate also increases the rate of point defect recombination.This means researchers may not be able to directly compare the microstructure development between the proton- and neutron-irradiated specimens. Even so understanding the differences produced by the irradiating particle types and dose rates as well as their implications is key to using ion-irradi- ation techniques to assess the long- term viability of F-M alloys as materials in advanced reactor systems. Limited studies have been carried out on neutron-irradiated F-M materials so the effects of particle types and dose rates in F-M alloys are yet to be resolved. HCM12A HT9 As Received Proton 3 dpa 500C Neutron 3 dpa 500C As Received Proton 3 dpa 500C Neutron 3 dpa 500C Grain diameter 10-6 m 0.66 0.66 0.66 0.36 0.30 0.30 0.30 0.10 Dislocation line density 1014 m-2 12.84 12.1 4.21 13.6 3.55 13.9 14.1 4.0 13.8 4.3 Carbide precipitate diameter 10-6 m 0.05 0.05 0.05 0.03 0.06 0.06 0.06 0.03 Carbide number density 1020 m-3 0.92 0.92 0.92 0.12 0.71 0.71 0.71 0.41 Void diameter 10-9 m 4.20 1.03 Void number density 1021 m-3 0.19 Dislocation loop diameter 10-9 m 10.8 4.20 12.5 4.11 11.9 6.12 10.0 3.62 Dislocation loop density 1021 m-3 0.9 0.9 2.2 0.9 Si-Mn-Ni-P cluster diameter 10-9 m 5.51 1.91 1.56 0.65 4.31 2.09 1.71 0.96 Si-Mn-Ni-P cluster no. density 1021 m-3 23 788 19 582 Cu-rich cluster diameter 10-9 m 3.96 1.42 1.97 0.47 Cu-rich cluster number density 1021 m-3 23 447 Cr-rich cluster diameter 10-9 m 0.54 0.13 0.47 0.10 Cr-rich cluster number density 1021 m-3 1312 1775 Table 1. Microstructure measurements of HCM12A and HT9 samples as-received and proton- or neutron-irradiated to 500C and 3 dpa. 2014 ANNUAL REPORT 89 Figure 1. Irradiation- induced dislocation loop diameter and number density produced by proton and neutron irradiations to 3 dpa at 500C in HCM12A and HT9. Project Description The goal of this project is to better understand the effects of varying dose rates on F-M alloys by comparing the damage inflicted on two identically heated commercial alloys HCM12A and HT9 by neutron and proton irradiation. Both alloys were irradiated in the ATR with neutrons at a dose rate of 10-7 dpasec and with 2.0 MeV protons at 10-5 dpasec. Both irradiations were carried out to 3 dpa at a temperature of 500C. Because it aims to 1 understand the response of advanced reactor candidate structural materials to irradiation and 2 assess the ability of proton irradia- tions to emulate in-reactor neutron irradiation damage this project has direct relevance to the Advanced ReactorTechnologies program being conducted by the Department of Energys Office of Nuclear Energy DOE-NE. The primary DOE-NE mission is to advance nuclear power as a viable resource for meeting the nations energy environmental and national security needs. Generation IV advanced reactor designs such as high- temperature reactors and fast-neutron spectrum reactors fulfill this mission by combining high-efficiency power generation with the environmental and national security benefits of consuming the extended-life radioactive isotopes found in spent nuclear fuel. However along with the promise of Generation IV designs comes the challenge of finding suitable structural This project enabled two graduate students to develop proficiency in microscopy techniquesand has laid the groundwork for their thesis projects. JanelleWharryAssistant ProfessorMaterials Science EngineeringBoise State University Nuclear Science User Facilities 90 materials that will withstand the harsh operating conditions in these new reactors. Ensuring the integrity of these materials under high temperatures corrosive environments cyclic loading and high irradiation damage is para- mount to the safety performance and long-term success of the Generation IV nuclear fleet. Accomplishments This project compared the microstruc- ture evolution of the commercial F-M alloys HCM12A and HT9 exposed to proton and neutron irradiations at 500C to 3 dpa. Results of this project suggest that proton irra- diation produces a dislocation loop morphology comparable to that of neutron irradiation but that irradi- ation-induced nanoscale clustering behavior varies considerably between the two particle types. Prior to this project neutron irradia- tions of HCM12A and HT9 had been completed in ATR at a dose rate of approximately 10-7 dpasec. Speci- mens from the same heats were also irradiated with 2.0 MeV protons at the Michigan Ion Beam Laboratory at a dose rate of approximately 10-5 dpasec. Researchers performed a full microstructure characterization on both the neutron- and proton- irradiated materials. Since the response of F-M alloys to irradiation has been found to be sensitive to variations in alloy heats examining these materials offered a tremendous opportunity to compare the effects of both proton and neutron irradiations on identical alloy heats. A combination of transmission electron microscopy TEM and local electrode atom probe LEAP analysis enabled thorough characterizations of the irradiated microstructures and phase evolutions.All material characterization work was conducted in the Microscopy and Characterization Suite MaCS at CAES utilizing the FEI Quanta focused ion beam FIB CAMECA 4000X HR LEAP and FEITecnai STwinTEM. TEM results showed that grain size dislocation line density and carbide precipitates had not been changed by either proton or neutron irradiation Table 1. Irradiation-induced disloca- tion loops were also observed byTEM in both HCM12A and HT9. In both alloys dislocation loop morphologies and number densities were consistent across both irradiation types although proton irradiation generated a slightly higher number density of loops in HT9 than did neutron irradiation Figure 1. Voids were found only in the neutron- irradiated HT9.These voids were small with diameters 2014 ANNUAL REPORT 91 Figure 3. LEAP tip reconstructions of top neutron- and bottom proton-irradiated HT9 to 3 dpa at 500C showing left to right Si Mn Ni P Cu Si-Mn-Ni-P-Cu cluster identification and Cr. Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor PIE facilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators Boise State University Janelle Wharry principal investigator Jatuporn Burns collaboratorYaqiao Wu collaborator Corey Dolph M.S. graduate student Matthew Swenson Ph.D. graduate student University of Idaho JoannaTaylor collaborator Figures 2 and 3.These clusters are comprised of approximately 3.57 at. Si 35 at. Mn 2.57 at. Ni and 0.251.0 at. P. Clusters in the proton- irradiated specimens contain a higher at. Si Mn Ni and P than do those in the neutron-irradiated specimens. Cu-rich clusters are found in both the proton- and neutron-irradiated HCM12A. Much like the Si-Mn-Ni-P clusters they are finer and more popu- lous in the neutron-irradiated specimen than in the proton-irradiated specimen. They are often but not always found coincident with the Si-Mn-Ni-P clusters Figure 2. Cu-rich clusters are not found in HT9 Figure 3 due to the low bulk concentration of Cu in HT9. Cr-rich clusters are found in both HCM12A and HT9 specimens but only in the alloys neutron-irradiated samples.These clusters are found at a very high number density on the order of 1024 m-3 . Their sizes are on the order of 0.5 nm.As such these clusters can only contain a few atoms each. LEAP tip reconstructions from irradiated HT9 Figure 3 show the differences in Cr clustering between neutron and proton irradiations.The mechanism for this Cr clustering is not yet understood. Work on this project was completed by Corey Dolph and Matthew Swenson graduate students at Boise State University.The project team wishes to acknowledge the assistance ofYaqiao Wu Jatuporn Burns and JoannaTaylor all of whom were instrumental in assisting with scheduling instrument training and LEAP analysis. Future Activities This project was completed in 2014. Future work will include additional irradiation experiments and computa- tional studies to further understand the mechanisms of nanoscale clustering in these alloys and their implications on mechanical behavior irradiation dose rate and particle type effects and to compare Cr clustering with radiation- induced segregation of Cr. Publications and Presentations 1. M.J. Swenson J.P.Wharry 2015 The strengthening mechanism transition in nanofeatured ferritic- martensitic alloys. The Minerals Metals Materials Society Annual Meet- ingOrlandoFL.March 2015. 2. M.J. Swenson J.P.Wharry. The comparison of microstructure and nanocluster evolution in proton and neutron irradi- ated Fe-9Cr oxide dispersion strengthened ODS steel to 3 dpa at 500C Submitted to Jour- nal of Nuclear Materials. Nuclear Science User Facilities 92 Electron Backscatter Diffraction and Atom Probe Tomography to Study Grain Boundary Chemistry Variation in Off Stoichiometric Uranium Dioxide Thin Films Michele Manuel University of Florida mmanuelmse.ufl.edu Project Description The objective of this research was to investigate the structure of a uranium oxide thin film produced by pulsed dc magnetron sputtering.This tech- nique allows films of specific micro- structures to be grown for analysis. Uranium dioxide is a pervasive mate- rial in nuclear energy and the ability to produce defined microstructures for testing makes this an exciting tech- nique especially given the difficulty in sample preparation of nuclear fuel. This specific project was to conduct atom probe tomography on the interface between the yttria-stabilized zirconia substrate and UO2 thin film to look for possible inter-diffusion that had occurred during processing. This research supports efforts to make the microstructural investiga- tion and testing of nuclear fuels more insightful and cost-effective. Figure 1. 3D APT reconstruction of the elements in the UO2YSZ thin film. Complex ions represented as UOx ZrOx YOx where x 1 2 or 3. Figure 2. Atom probe 1-D concentration profile taken from the cylinder. 2014 ANNUAL REPORT 93 Accomplishments Samples were fabricated on the focused ion beam at the Center for Advanced Energy Studies CAES in Idaho Falls Idaho and then analyzed with atom probe tomorgraphy.The goal of the project to visualize the interface between film and substrate was successful and incorporated into a publication in the journal Applied Surface Science.This research was conducted primarily by BillyValder- rama and facilitated in large part by staff at CAES including Jatuporn Burns Dr.YaqaioWu and JoannaTaylor. Access to the CAES facility has provided unprecedented insight into the behavior of nuclear fuels. Michele ManuelAssociate ProfessorDepartment of Materials Science and EngineeringUniversity of Florida Research to be completed This research is complete as it was a single study of one material for a publication. Publications 1. J. Lin I. Dahan B.Valderrama M. V. Manuel 2014 Structure and Properties of Uranium Oxide Thin Films Deposited by Pulsed DC Magnetron Sputtering Applied Surface ScienceVol. 301 pp. 475480. Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators Idaho National Laboratory Jian Gan collaborator University of Florida Michele Manuel principal investigator Billy Valderrama collaborator Hunter Henderson collaborator The ability to produce specific fuel microstructures by Pulsed DC Magnetron Sputtering is a powerful sample preparation technique and the research conducted at CAES pushes this method forward to full validation. Nuclear Science User Facilities 94 Study of the Microstructures of Krypton and Xenon In Irradiated Uranium Dioxide by Advanced Microscopy Techniques Lingfeng He Idaho National Laboratory lingfeng.heinl.gov Initiated in 2012 this project is a collaborative effort of INLArgonne National Laboratory and the University ofWisconsin.The purpose of this rapid turnaround experiment is to clarify the microstructure and stoi- chiometry of uranium dioxide UO2 under irradiations of krypton Kr and xenon Xe ions. Project Description Kr and Xe are the main fission gases in UO2 nuclear fuel.The radiation defects i.e. dislocation loops and precipitates i.e. gas bubbles induced during fission affect the fuels struc- ture and thermal transport properties. It is therefore important to understand and model the behaviors of Kr and Xe in order to develop optimized fuel microstructures that allow improved performance of the UO2 fuel. Accomplishments To simulate the fragment damage to UO2 caused by fission depleted UO2 samples were irradiated with Kr and Xe ions.To date the research team has used a focused ion beam FIB to prepare cross-section lamina of the irradiated samples and transmission electron microscopy TEM to study the extended defects including bubbles dislocation loops and dislocation lines as well as stoichiometry change. The results indicate bubble formation under irradiation at room-temperature Figure 1 with no solid precipitates forming in the bubbles Figure 2. This indicates that uranium U vacancies may be mobile at tempera- tures below room temperature. It could also imply that KrXe bubbles may directly nucleate at the vacancy clusters that are produced in cascades at room temperature. Such a process might not require that KrXe and U vacancies be diffusive. In addition electron energy loss spectroscopy EELS shows that under a vacuum the stoichiometry of UO2 is relatively stable during irradiation at room temperatures.At the same time the branching ratio of the M-edge in EELS is 0.695 which is within the range of 0.6950.720 for U4 .This implies that the stoichiometry in UO2 is unaf- fected by Xe irradiation Figure 3. We also found that the formation of dislocation-denuded zones is tempera- ture sensitive.At 800C enough U interstitials migrate toward the grain boundaries so that the concentra Figure 1. TEM images of bubbles in a UO2 crystal implanted with Xe at a dose of 5 1014 ionscm2 . 2014 ANNUAL REPORT 95 This experiment will improve our understanding of microstructure evolution in uranium dioxide under irradiation. Figure 2. a High-resolution TEM images of bubbles in a UO2 single crystal irradiated with 300-keV Xe at room temperature at a dose of 5 1014 ionscm2 . b An under-focus image of a. Some bubbles are marked with arrows. Nuclear Science User Facilities 96 Our understanding of nuclear fuels has improved greatly thanks to the state- of-the-art techniques used in this project. Lingfeng HeAssistant ScientistUniversity of Wisconsin-Madison Currently Nuclear Fuels Engineer at Idaho National Laboratory tion of interstitials near the grain boundaries becomes too low to form loops resulting in the formation of dislocation-denuded zones near the grain boundaries. However at 600C the dislocation-denuded zones were not found even at high-angle grain boundaries Figure 4. Future Activities Circular features having diameters of 100200 nm were found in UO2 irra- diated with Xe at high doses. However the nature of these features is still not clear and furtherTEM study is planned in 2015. Publications and Presentations 1. L.F. He M. Gupta M.A. Kirk J. Pakarinen J. GanT.R.Allen 2014 In SituTEM Observation of Dis- location Evolution in Polycrystalline UO2 JOMVol. 66 pp. 25532561. 2. L.F. He J. Pakarinen M.A. Kirk J. GanA.T. Nelson X.-M. BaiA. El-AzabT.R.Allen 2014 Micro- structure evolution in Xe-irradi- ated UO2 at room temperature Nuclear Instrument and Methods in Phys- ics Research BVol. 330 pp. 5560. 3. L.F. He B.ValderramaA.-R. Has- san J.Yu M. Gupta J. Pakarinen H.B. Henderson J. Gan M.A. KirkA.T. Nelson M.V. Manuel A. El-Azab andT.R.Allen 2015 Bubble formation and Kr dis- tribution in Kr-irradiated UO2 Journal of Nuclear MaterialsVol. 456 pp. 125132. Figure 3. a Energy dispersive x-ray EDX and b and c EELS spectra of a UO2 single crystal irradiated with 300 keV Xe at room temperature up to a dose of 1 1016 ionscm2 . d The second derivative EELS spectrum of part c. 2014 ANNUAL REPORT 97 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor PIE facilities Argonne National Laboratory IntermediateVoltage Electron Microscopy IVEM-Tandem facility Collaborators University of Wisconsin - Madison Lingfeng He principal investigator Mahima Gupta collaborator Idaho National Laboratory Jian Gan collaborator Argonne National Laboratory Marquis Kirk collaborator Figure 4. a A scanning electron microscope SEM image of a TEM lamina prepared by FIB showing three grains A B and C. b High-resolution TEM image showing the grain boundary between grains A and B the orientations of grains A and B are close to 125 and 011 respectively. c Selected-area diffraction pattern of grains A and B. d Bright-field TEM images showing the dislocation loops in grains A and B that have been irradiated with 1 MeV Kr at 600C at a dose of 5 1014 ionscm2 . Nuclear Science User Facilities 98 X-ray Absorption Near Edge Spectroscopy and Extended X-ray Absorption Fine Structure Study of Technetium-99 Relevant to the Nuclear Fuel Cycle Silvia Jurisson University of Missouri jurissonsmissouri.edu The synchrotron radiation techniques found in x-ray absorption near-edge spectros- copy XANES and extended x-ray absorption fine-structure spectroscopy EXAFS are among the very few that provide oxidation-state and nearest-neighbor information about compounds and materials that do not crystallize suitably or are not avail- able in sufficient quantities for x-ray diffraction analysis XRD. Technetium-99 Tc-99 is an impor- tant radionuclide in the nuclear fuel cycle. It is a major fission by-product of both uranium U and plutonium Pu and serves a vital role in envi- ronmental remediation at nuclear sites such as the Department of Energys reactor near HanfordWashington. Project Description We proposed to perform XANES and EXAFS experiments at theAdvanced Photon Source APS atArgonne National Laboratory ANL to determine the end products of immobilization reactions between pertechnetate TcO4 and sulfide in the presence of various olefinic acids.The results will allow us to determine the oxidation states of technetium Tc as well as those of its nearest neighbors in theTc metal center i.e. its coordination environment.The resulting data will provide mechanistic Figure 1. Tc K-edge EXAFS spectra of the standard complexes and experiment samples. 2014 ANNUAL REPORT 99 information on the factors influencing the immobilization reactions that are important to both environmental reme- diation and nuclear fuel reprocessing. We believe that understanding the products of theseTc reactions will enable the development of better immobilization methods and poten- tially superior separation methods. Since examining theTc compounds should be a fairly straightforward process our goal is to analyze this data quickly. In order to expedite the development of a base of informa- tion for further exploration we have proposed that the work and analysis be performed in 2013 and 2014. Accomplishments Under the guidance of JeffTerry at the Illinois Institute ofTechnology graduate students Kim Reinig and Rachel Seibert ran the EXAFS and XANES spectra at the APS. They used X-ray absorption spectros- copy XAS to investigate the reaction ofTcO4 with the unsaturated acids maleic acid and fumaric acid in the presence of sulfide. Nuclear magnetic resonance NMR of these reactions has shown that sulfide reacts across the double bond forming mercaptosuc- cinic acid MSA or dimercaptosuccinic acid DMSA.The resulting mercapto- acids then chelate theTc and hinder or even prevent its immobilization to technetium heptasulfide Tc2S7.They also prevent or hinderTcs adsorption into another minerals surface. Tc2S7 is the product formed through the reaction ofTcO4 with sulfide in the absence of unsaturated acids.The Tc standard complexes TcOMSA21- and TcODMSA21- were used as comparators for the reactions of maleic acid and fumaric acid withTcO4 in the presence of sodium sulfide Na2S. The following seven solid samples were analyzed by Reinig and Seibert at the APS. Graduate students Kim Reinig left and Rachel Seibert right in the Advanced Photon Source at Argonne National Laboratory. Nuclear Science User Facilities 100 Figure 2. Fourier Transformation of the EXAFS spectra for the crotonic acid reaction. 1. NH4TcO4 2. NaTcv OMSA2 3. NaTcv ODMSA2 4. Tc2S7 5. TcO4- SH- maleic acid at pH 7 6. TcO4- SH- fumaric acid at pH 9 7. TcO4 - SH- crotonic acid at pH 7 Reinig and Seibert also investigated the effects unsaturated carboxylic acids had on the reaction ofTcO4 with sulfide which followed on the projects previous work 1.The reaction that formedTc2S7 is shown below.A phosphate buffer was used to control the pH. 2 99 TcO4- 7 SH- H2O Tc2S7 9 OH- In the presence of unsaturated carboxylic acids the sulfide appears to be building across the double bond to form the mono- or dithiol- carboxylic acids but only whenTcO4 is present.The product formed during the above reaction is a pentavalent technetium complex TcVoxo species containing the newly formed mono- or dithiol-carboxylic acid. NMR spectra of the end products of the purified reaction show the pres- ence of these species compared to the standard complexes. EXAFS and XANES helped confirm the identities of the end products. EXAFS and XANES were also used to determine whether aTc-S bond similar to that observed inTcvOMSA2 orTcvODMSA2 occurred. Compari- sons of each samples EXAFS spectra to the control standards showed profiles similar to the TcOMSA21- and TcODMSA21- samples Figure 2. 2014 ANNUAL REPORT 101 The EXAFSXANES experiments at the APS allowed us to confirm the coordination environments about theTcV centers on reaction of pertechnetate with sulfide in the presence of the unsaturated acids. Without these results the product identities would have been more speculative. Dr.Silvia JurissonProfessor of Chemistry and Radiology University of Missouri Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Argonne National Laboratory Advanced Photon Source Collaborators University of Missouri Silvia Jurisson principal investigator Kimberly Reinig graduate student collaborator Illinois Institute of Technology JeffTerry co-principal investigator Rachel Seibert graduate student collaborator The crotonic acid sample contains a TcO4 contaminant but also shows evidence of aTc-S bond Figure 3. Future Activities The collected data continues to be analyzed and a manuscript describing the chemistry ofTcO4 with sulfide in the presence of unsaturated acids is being drafted.The EXAFS and XANES data along with the 500-MHz NMR spectra will show that the presence of unsaturated acids can hinder the immobilization ofTc-99. This work will provide important data on the immobilization reactions of the Tc-based compounds that are important to both environmental remediation and nuclear fuel reprocessing. References 1. Y. Liu J.Terry S. Jurisson 2009 Potential Interferences on the Pertechnetate-Sulfide Immobiliza- tion Reaction. Radiochimica Acta Vol. 97 pp. 3341. Publications and Presentations 1. We are working on the manuscript and plan to submit it in 2015. Nuclear Science User Facilities 102 Post-Irradiation Examination of ATR-Irradiated Ultra-Fine Grained Steel K. Linga Murty North Carolina State University murtyncsu.edu Currently 436 nuclear reactors currently provide an average of about 20 of the energy in the 30 countries that have nuclear power programs 1. However the structural materials that will be used for the new advanced reactor systems will be subjected to far higher neutron fluences 150 dpa than those in todays oper- ating reactors 50 dpa 2. Therefore it is imperative that we develop new more-radiation-tolerant materials and investigate the effects irradiation has on them. Ultra-fine grained UFG materials have shown promising mechanical properties and extensive research has been conducted on the different processing techniques that could improve their properties across various applications 3. Previous studies have shown that the high densities of grain boundaries in UFG materials act as sinks for irradi- ation-induced defects reducing the influence of the irradiation defects 4. Project Description In this study researchers investigated the effects of neutron irradiation on UFG ferritic steel that had been prepared using the equal-channel- angular-pressing ECAP technique 3 4.The experimental material is an ECAPed low-carbon mild steel consisting by weight of 0.1 carbon C 0.5 manganese Mn 0.27 silicon Si and the balance 99.13 iron Fe. Both UFG samples along with their conventional grain CG steel counterpartsprepared by annealing UFG samples at 800C for 1 hourwere irradiated to 1.37 displacements per atom dpa in ATR at INL.The irradiation capsules were designed to keep the irradiation temperature of the samples below 100C.The mean grain sizes for the UFG and CG steels are 0.35 m 0.18 m and 4.4 m 1.8 m respectively Figure 1. Accomplishments After irradiation microstructural and mechanical properties of the CG and UFG steels were investigated using transmission electron microscopy TEM electron back scattered diffrac- tion EBSD atom probe tomography APT x-ray diffraction XRD and micro hardness and tensile testing. TEM micrographs showed no grain growth post irradiation in the UFG steel. ESBD showed similar results for CG steel. XRD was used to determine the dislocation density for both steels before and after irradiation by fitting the XRD patterns to a pseudo-Voigt pV function using the Modified Rietveld technique 5. CG steel exhibited an increase in the dislocation density post irradiation. On the other hand with values well within error bars UFG steel showed no significant change.APT analysis revealed a high number of nano Mn-Si-enriched precipitates in both CG and UFG steels Figure 2. However Figure 1. Pre-irradiation EBSD patterns and grain size distributions for a UFG and b CG steels. 2014 ANNUAL REPORT 103 while the radii of the clusters in both steels are similar 0.97 m 0.23 m and 0.9 m 0.16 m for UFG and CG steels respectively the density of clusters in the UFG steels 1.2 1024 m-3 is about twice that of CG steel 6.71023 m-3 Figure 3. Researchers believe this is due to the shorter distance defects have to diffuse in UFG before they reach the grain boundary and the resulting lower probability of defect recombination in the matrix. No clusters were observed before irradiation indicating that their formation was radiation-induced. Vickers micro hardness tensile test and dislocation density results are shown inTable 1. The average micro hardness values for CG steel increased by 62 after irradiation compared to only 8.6 for the UFG steel Figure 4.Tensile test results revealed that CG yield strength increased by 132 after irradiation and its ductility decreased by 82 while the yield strength of UFG steel increased by 30 and the ductility reduced by 56 Figure 5.Although irradiation hardening was minute in the UFG steel compared to its CG counterpart the irradiation-induced embrittlement is clearly apparent in the UFG steel after irradiation albeit the percentage decrease in the ductility of UFG steel is quite less than that of its CG counterparts. According to Odette and Lucas 6 the primary mechanism for embrittle- ment in ferritic steels is the hardening produced by nanometer-sized features that develop as a consequence of radiation exposure. Since our results showed that there is no significant change in dislocation density in UFG steel after irradiation the observed hardening in UFG steel is likely a product of the high density of irradiation-induced Mn-Si-enriched clusters found in UFG steel Table 2. To gain a better understanding of the effect of irradiation on microstructural changes in the steels mechanical properties refer to Alsabbagh et. als 2014 journal article 7 for a discus- sion of how the increase in yield stress Dosedpa m-2 Yield Strength MPa Tensile Strength MPa Vickers Hardness MPa Ductility CG 0 1.06 0.131014 29617 39018 128413 633 1.37 4.26 0.561014 68720 75420 208151 111 UFG 0 9.50 1.241014 7759 98011 308854 182 1.37 8.98 1.391014 100949 106043 335368 81 Table 1. Mechanical properties for both CG and UFG steels before and after irradiation. Figure 4. Micro hardness before and after irradiation for both UFG and CG low carbon steel Oro-Ash MPa Dis MPa Calculated MPa Measured Mpa CG Steel 137 44 245 35 382 57 391 27 UFG Steel 230 90 -20 73 210 115 234 50 Table 2. Estimated strength increments for different strengthening mechanisms for both CG and UFG steels. Nuclear Science User Facilities 104 is related to different strengthening mechanisms for example strength- ening due to grain size solid solution clusters and dislocations. The demonstrated agreement between the experimental results and the actual physical strength mechanisms indicates that the Orowan-Ashby andTaylor strengthening models can be useful for explaining how nano- cluster strengthening and disloca- tion forest hardening mechanisms contribute to the overall strength of the ECAP UFG and the CG steels.The results show that while irradiation- induced dislocation density is an important factor in the increase in total yield stress in neutron-irradiated CG steel irradiation hardening in the UFG steel is mainly due to the irradiation-induced clusters. Nanocrystalline and UFG metals with relatively large volumes of interfaces are expected to be more radiation resistant than conventional metals. Figure 2. Representation of a three- dimensional 3D reconstruction of CG low-carbon steel a before and b after neutron irradiation by 3D atom- probe microscopy. Figure 3. Si-Mn-enriched cluster distribution post-neutron irradiation for a CG and b UFG steel. The dimension of the interior colored boxes is 20 x 20 x 100 nm. Figure 5. Engineering stress strain curves for a UFG and b CG before and after irradiation to 1.37 dpa. In summary as the area of grain boundaries which act as sinks for radiation-induced defects is signifi- cantly increased by grain refinement UFG steel showed better irradiation tolerance compared to its CG coun- terparts. However irradiation-induced solute clustering in UFG alloys needs to be carefully considered. Future Activities Investigations are underway at the Center for Advanced Energy Studies CAES and INL to determine the effects of irradiating UFG and CG steels to higher doses which will shed light on the effects of dose on strengthening and embrittlement. 2014 ANNUAL REPORT 105 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor PIE facilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators North Carolina State University K.L. Murty principal investigatorAhmad Alsabbagh graduate student Idaho National Laboratory Doug Porter principal investigator Brandon Miller collaborator This collaborative research enabled me to work with advanced microstructural charac- terization techniques through the ATR-NSUF program.I was also able to learn from many working professionals at CAES and INL that resulted in my receiving the 2014 ANS Mark Mills Award for a technical article I wrote based on the research I did to earn my doctoral degree in nuclear science and technology. Ahmad AlsabbaghPh.D. Graduate Student until June 2014Nuclear Engineering North Carolina State University Publications and Presentations 1. A.AlsabbaghA. Sarkar B. Miller J. Burns L. Squires D. Porter J. Cole and K. Murty. 2014 Micro- structure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel Materials Science Engineering AVol. 615 pp. 128138. See additional publications from other years in the Media Library on the NSUF website. References 1. K. L. Murtry 2012 Nuclear Ma- terials ScienceEnablingTechnol- ogy for Sustained Operation fo Nuclear Power Generation and Development of Next Generation Power Plants Nuclear Energy Science Power GenerationTechnologyVol. 1 No. 1. 2. S.J. Zinkle J.T. Busby 2009 Structural materials for fission fusion energy MaterialsTodayVol. 12 No. 11 pp. 1219. 3. T.C. Lowe R.Z.Valiev 2004 The use of severe plastic deformation techniques in grain refinement JOMVol. 56 No. 10 pp 6468. 4. A.Alsabbagh R.Z.Valiev and K.L. Murty 2013 Influence of grain size on radiation effects in a low carbon steel Journal of Nuclear MaterialsVol. 443 pp. 302. 5. L. Lutterotti and P. Scardi 1990 Simultaneous structure and size strain refinement by the Rietveld method Journal of Applied Crystallog- raphyVol. 23 pp. 246. 6. G.R. Odette and G.E. Lucas 2001 Embrittlement of nuclear reactor pressure vessels JOMVol. 53 No. 7 pp. 1822. 7. A.AlsabbaghA. Sarkar B. Miller J. Burns L. Squires D. Porter J.I. Cole K.L. Murty 2014 Micro- structure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel Materials Science and Engineering AVol. 615 pp. 128. Nuclear Science User Facilities 106 Microstructural and Mechanical Characterization of Self-Ion Irradiated 14LMT Nanostructured Ferritic Steel Indrajit Charit University of Idaho icharituidaho.edu Advanced reactors need high- performance materials to serve under harsh conditions such as higher temperature higher radia- tion doses and extremely corrosive environments. Nanostructured ferritic steels NFS are such a class of mate- rials.They are produced by mechanical alloying MA of the elemental or pre-alloyed metallic powder typically incorporating nanosized yttria Y2O3 powder followed by a traditional consolidation process such as hot extrusion or hot isostatic pressing HIP. NFS have excellent potential for advanced fuel cladding and structural materials applications in fast reactors. Project Description Since neutron irradiation is out of the scope of this RapidTurnaround Experiment RTE the aim of the project is to investigate a new NFS known as 14LMT Fe-14Cr-1Ti- 0.3Mo-0.5La2O3 wt. that was recently developed by this research group and other collaborators. Samples were irradiated at doses up to 100 displacements per atom dpa and relevant microstructural charac- terization and mechanical properties evaluations were performed. Titanium is generally added toY2O3 to form complex yttria-titanium-oxygen Y-Ti-O-based particles in order to make the dispersed oxides much finer and stable at elevated temperatures. The 14LMT alloy uses lanthana La2O3 instead of the traditionalY2O3. Spark plasma sintering SPS was used to consolidate the mechanically alloyed powder. NFS performance is largely determined by the ultra-high number density of nanosized oxide particles dispersed throughout the microstructure.These nanofeatures are stable at high tempera- tures and are expected to impart excel- lent high-temperature strength and irradiation stability to NFS.This work could lead to the development of high- performance fuel cladding materials for advanced fast reactors. Accomplishments The stability of nanoclusters in NFS under irradiation is critical. Collision cascades can eject solute atoms from them and change their physical char- acteristics. In order to understand their stability the lanthana-bearing 14LMT alloy was exposed to self-ion Fe2 irradiation at both room temperature 30C and elevated temperature 500C as a function of ion dose at 10 50 and 100 dpa. Subsequently the irradiated material was characterized by transmission electron microscopy TEM for microstructural character- istics atom probe tomography APT for nanocluster sizecompositional analyses and nanoindentation to measure hardness. Overall morphology and number density of the nanofeatures remained largely unchanged after irradiation. The average radius of the nanofeatures in the sample irradiated at 500C100 dpa was slightly reduced.A notable level of irradiation hardening and enhanced dislocation activity occurred after ion irradiation except at 30C and 50 dpa. Other microstructural features such as grain boundaries and a high density of dislocations also provided defect sinks to assist in defect removal. A comprehensive paper based on this work is under review with the Journal of Nuclear Materials. Future Activities While this project has been completed a new RTE project Microstructural and Nanomechanical Characterization of a Lanthana-Bearing Nanostructured Ferritic Steel Irradi- ated with High Dose Iron Ions has recently been approved.This will allow the project team to continue its work on understanding ion-irradiation response of 14LMT to higher dose levels up to 400 dpa. This work could lead to the development of high-performance fuel cladding materials for advanced fast reactors. 2014 ANNUAL REPORT 107 Publications and Presentations 1. S. Pasebani I. Charit J. Burns S. Alsagabi D.P. Butt J.I. Cole L.M. Price and L. Shao Microstructur- al stability of a self-ion irradiated lanthana-bearing nanostructured ferritic steels Journal of Nuclear Materials under review. 2. S. Pasebani I. Charit 2014 Ef- fect of alloying elements on the microstructure and mechanical properties of nanostructured ferritic steels produced by spark plasma sintering Journal of Alloys Compounds 599 pp. 206211. Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators University of Idaho Indrajit Charit principal investigator Somayeh Pasebani collaborator Boise State University Darryl Butt collaborator Boise State University Center for Advanced Energy Studies Yaqiao Wu collaborator Jatu Burns collaborator Texas AM University Lloyd Price collaborator Idaho National Laboratory James Cole principal investigator The ATR NSUF MaCS microstructural characteriza- tion facility provides easy access to state-of-the-art instruments that are not easily available at our campus. This has been extremely beneficial to our research productivity. Indrajit CharitAssociate ProfessorUniversity of Idaho Figure 1. A TEM image of the oxide particles observed in the 14LMT alloy specimen unirradiated at 500C. Figure 2. A TEM image of the 14LMT alloy specimen irradiated at 500C100 dpa. 3. S. Pasebani 2014 Processing of Oxide Dispersion Strengthened Alloys via Mechan- ical Alloying and Spark Plasma Sintering Doctoral Dissertation University of Idaho Moscow Idaho. See additional publications from other years in the Media Library on the NSUF website. Nuclear Science User Facilities 108 Microstructural and Mechanical Characterization of Self-Ion Irradiated Grade 92 Steel Indrajit Charit University of Idaho icharituidaho.edu Ferritic-martensitic F-M steels have the potential to be employed in advanced reactors. One of these Grade 92 9Cr-2W steel is being considered for structural applications in advanced nuclear energy systems. It has good mechan- ical and thermophysical properties and is considered to have potential for both in-core and out-of-core applica- tions. Still the irradiation performance of this alloy is not fully understood because of limited available data. This experiment could lead to a better understanding of Grade 92s irradia- tion performance in fuel cladding and structural materials for advanced reactors. If proven successful high- performance materials such as Grade 92 steel would help improve the safety and reliability of future reactors and give them a longer service life. Project Description In this project Grade 92 samples were irradiated with high energy Fe2 ions at theTexas AM University Ion Beam Laboratory and the induced micro- structural evolution and mechanical properties that resulted were inves- tigated. Irradiated specimens were prepared for transmission electron microscopy TEM using a focused ion beam FIB. Microstructural char- acterization of the irradiated speci- mens was performed usingTEM and mechanical properties were evaluated using the nanoindentation technique. Accomplishments Grade 92 steel specimens were irradi- ated to 10 50 and 100 displacements per atom dpa using the IoneX 1.7 MVTandetron accelerator.This effort was coordinated by Professor Lin Shao and graduate student Lloyd Price. This experiment could lead to a better understanding of Grade 92s irradiation performance in fuel cladding and structural materials for advanced reactors. Figure 1. The calculated irradiation damage profile in terms of dpa versus length. 2014 ANNUAL REPORT 109 The experiments were conducted at temperatures of 30C and 500C.The damage profile as calculated from Stopping and Range of Atoms in Matter SRIM is shown in Figure 1. The irradiated samples were then prepared forTEM studies using a focused ion beam - scanning electron microscope FIB-SEM Quanta 3D field emission gun FEG in the CAES MaCS facility. Ms. Jatu Burns assisted graduate student Sultan Alsagabi in this effort. The FIB-SEM procedure started with protecting the irradiated surface with a platinum Pt deposition layer. Figure 2 shows the specimen after irradiation. For detailed microstructural charac- terization aTF30 FEG scanningTEM STEM was utilized operating at an accelerating voltage of 300 kV.The FIB samples were attached to a copper grid before examination inside the TEM.The dislocation density of the prepared samples was estimated while they were oriented in a two-beam condition and the electron energy loss spectrum EELS technique was applied to measure sample thickness. TEM images of the irradiated samples were obtained from the irradiated area of the FIB sample. Samples from the unirradiated area were also examined for comparison purposes. Graduate student Alsagabi completed this task with the assistance of Dr.Yaqiao Wu of the CAES MaCS staff. The irradiation-induced damage evolution in the samples at 30C as a function of damage dose dpa is shown in Figure 3.The irradiated samples did not show any distinct irradiation-induced precipitates dislo- cation loops or voids. However they do show the presence of irradiation- Figure 2. The FIB sample showing the Pt-deposited layer irradiated and unirradiated areas. Nuclear Science User Facilities 110 induced defect clusters appearing as black dots.The overall concentra- tion of these clusters increased as the dose was increased.The high density of induced clusters did not appear outside the irradiated areas. The unirradiated area did not show any form of clustering at all which confirms that these clusters were not induced by the FIB. In samples irradiated at 10 dpa30C these defect clusters were very small in size. They increased to around 34 nanometers nm in samples irradiated at 50 dpa and 100 dpa.The increased irradiation dose enhanced not only the formation of induced defect clusters but also the number and density of these defects which saturated the irradiated areas in later stages. This can be attributed to the limited mobility of clusters at 30C. Finally a Hysitron Nanoindenter model TI-950TriboIndenter was used to measure hardness as a function of depth. The indentation was made in a direction normal to the surface of the sample using a constant time segment to induce a penetration depth profile of 1000 nm. Samples irradiated at 30C showed an evident hardening phenomenon with the hardness increasing as the dose increased. However at 500C the irra- diation hardening was less pronounced. Figure 3. TEM images of irradiated Grade 92 specimens under two-beam bright field conditions at a 10 dpa30C b 50 dpa500C and c 100 dpa500C. Figure 4. Hardness of Grade 92 steel as a function of irradiation dose at 30C and 500C. 2014 ANNUAL REPORT 111 The ATR-NSUF MaCs microstructural characterization facility provides easy access to state- of-the-art instruments. There is no doubt in my mind that this access has been very positive in our research and education. Indrajit CharitAssociate ProfessorUniversity of IdahoDistributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators University of Idaho Indrajit Charit principal investigator Sultan Alsagabi collaborator Boise State University Center for Advanced Energy Studies Jatu Burns collaboratorYaqiao Wu collaborator Texas AM University Lloyd Price collaborator Lin Shao collaborator Idaho National Laboratory James Cole principal investigator Clearly the higher irradiation dose created greater hardness especially in the near-surface region less than 300 nm so the monotonic increase of hardness with increased fluence was obtained. However the material irradiated at 10 dpa500C exhibited higher hardness in the near-surface region than the material irradiated at 50 dpa500C. Figure 4 shows the hardness data plotted as a function of depth at 250 nm showing a gradual increase at 30C.At 500C the hardness gradually decreased but then increased again during recovery. During the reported time under irra- diation 417 min. lower hardening was observed in the sample irradiated at 50 dpa500C which indicates an enhanced recovery effect at 500C. The main results can be summarized as follows 1. The irradiation-induced hardening was obvious and increased as the dose was increased at 30C but was less definite at 500C as an annihi- lation mechanism becomes more pronounced.This was accompanied by a higher dislocation density of samples irradiated at 30C. 2. The samples irradiated at 30C showed irradiation-induced defect clusters appearing as dots in the irradiated area and the density of these clusters increased as dose increased. Some dislocation loops were found in the sample irradi- ated at 50 dpa500C. 3. At 30C the number and density of irradiation-induced clusters increased as dose increased and at 50 dpa and 100 dpa they saturate to 34 nm. 4. The investigated samples did not show any cavities such as bubbles or voids under the testing parameters. Future Activities The RapidTurnaround Experiment- Project RTE has been completed. No further work is planned at this time. Publications and Presentations 1. S.Alsagabi 2014 HighTemperature Deformation BehaviorThermal Stability and Irradiation Performance in Grade 92 Steel Doctoral Dissertation Uni- versity of Idaho Moscow Idaho. Nuclear Science User Facilities 112 Development of Advanced Crystallographic Analysis Technique for Localized Fission Product Transport in Irradiated Silicon-Carbide Isabella van Rooyen Idaho National Laboratory isabella.vanrooyeninl.gov The proposed study aims to perform compositional analysis to gain insight into a possible link between crystallographic informa- tion and fission products location in the silicon-carbide SiC layer of tristructural isotropic TRISO fuel.This information will help identify transport mechanisms for these fission products especially silver Ag as they migrate through the SiC layer. Recent work has shown that it is only possible to identify low concentrations of fission products inTRISO-coated particles especiallyAg using transmis- sion electron microscopy TEM with a field emission gun FEG 1.The preliminary results of this study indicate that fission products are primarily associated with some but not all grain boundaries and triple junctions. Knowledge of the grain boundary crys- tallographic parameters will be required to explain the varied fission transport behaviors associated with individual grain boundaries. Currently no efficient way exists for measuring the grain boundary parame- ters while correlating them with fission product transport behavior.What is needed is a way of measuring both the chemistry at the grain boundary and the grain boundary parameters. Since the chemical analysis resolution has only been demonstrated in the FEG TEM a technique of measuring the grain boundary parameters in theTEM is also required. Attempts at measuring the grain boundary parameters using Kikuchi bands inTEM electron diffraction patterns under theAdvanced Gas Reactor AGR-1 program have been unsuc- cessful. However an emerging tech- nique utilizingASTAR software from NanoMEGAS SPRL which examines the intensity of electron diffraction spots in precession electron diffraction PED patterns shows potential for deter- mining the necessary crystallographic parameters in theTEM Project Description This project consists of two phases. In the first researchers will develop the ASTAR technique and evaluate its resolu- tion and reliability when examining SiC from an unirradiatedTRISO fuel particle. The second phase and the main focus of this project will be to characterize the grain and grain boundary parameters in the SiC layer of irradiatedTRISO fuel particles.Table 1 outlines the samples to be analyzed and the expected outcomes. Phase 1 Research on unirriadiated SiC and sample optimization forASTAR Thicker samples increase the possibility of multiple grains being present in the focused ion beam FIB sample. Electron diffraction caused by these multiple grains degradesASTARs ability to identify the correct crystal- lographic orientation.Therefore three TEM lamellae of unirradiated SiC with different thicknesses will be prepared using FIB techniques 50 100 and 150 nm at CAES or the Electron Microscopy Laboratory EML.The quality of theASTAR data will be evaluated for each thickness from the reliability map.The data obtained from these samples will also provide the baseline microstructure and crystal- lographic information texture of the as-fabricated SiC layer which will then be compared to that of SiC from irradi- atedTRISO particles. Phase 2 Research on irriadiated SiC The main objective is to demonstrate that ASTAR can be used to determine the crystallographic information of grains in irradiated SiC determine the grain boundary parameters using the crystallographic orientations of adjacent SiC grains and correlate that information with the transport of fission products. Samples have already This study provides first-of-a-kind nano-crystallographic information on neutron-irradiated TRISO fuels by the application of precession electron diffraction measurements. 2014 ANNUAL REPORT 113 been fabricated using the Materials and Fuels Complex MFC EML-FIB from Particle 35 of Compact 6-3-2 which experienced 11.3 burnup during the AGR-1 experiment.Although these samples may not be the optimum thickness they will be sufficient to demonstrate ASTARs ability to deter- mine the orientation of individual irradiated SiC grains and to reveal any other issues that may interfere with the accurate determination of the crystal- lographic orientation.The sample set includes one taken perpendicular to the growth direction of the SiC layer and three taken parallel.All these samples have already had a considerable amount of compositional analysis performed on them. Ag has been identified in these samples and several locations have been specifi- cally identified forASTAR analysis. Since the three parallel samples span the entire thickness of the SiC layer theASTAR data will show how the microstructure develops during the growth of the SiC layer.The data will also be used to determine the misorientation across specific grain boundaries especially those that have previously been found to containAg.The misorientation informa- tion will be used to determine whether it is a high-angle low-angle random or coincident site lattice CSL-related grain boundary. By assessing the crystal- lography of grain boundaries with and without fission products it may be possible to make preliminary conclu- sions on the types of grain boundaries that are resistant to fission product transport. However it is unlikely that a statistically relevant dataset will be collected during this RTE. Figure 1. Three FIB-fabricated samples were prepared as shown in the micrograph left. The FIB lamellae at post B were lost due to handling and only lamellae from posts A top right and D bottom right were analyzed using ASTAR. Nuclear Science User Facilities 114 Figure 2. The crystallographic orientation of the area in the TEM image on the left is mapped on the right. Accomplishments The work was initiated in February 2014 resulting in the successful completion of all experimental work for the two phases both on unir- radiated and irradiated samples as indicated inTable 1.The integration of all project results and papers is still to be performed although some results have already been prepared for confer- ence or workshop presentations. 2-4 Summary of Phase 1 results Research on unirriadiated SiC and sample optimization for ASTAR 2.3 Sample fabrication techniques as well as sample characteristics can criti- cally influence the quality of the data collected. Researchers initially evalu- ated the influence of sample thickness on the quality of the orientation data generated byASTAR. It is extremely difficult to prepare samples thinner than 80 nm from these materials due to the brittle nature of neutron-irradiated SiC.Although the original study plan included three FIB-prepared samples only two samples of varying thickness Figure 1 were analyzed.The thick- nesses of five areas on these two samples were determined by electron energy loss spectroscopy EELS e.g. location and height of the plasmon peak.Three areas were found to be approximately 80 nm thick while two others were found to be approximately 120 nm thick. It is extremely difficult to prepare FIB lamellae thinner than the measured 80 nm due to the brittle nature of neutron-irradiated SiC as demonstrated by the fractured regions shown in Figure 1 center.The crystallographic information in the latter two areas was collected using theASTAR system on the TecnaiTF30-FEG STwin at CAES. The Index parameter calculated by ASTAR was taken to be the primary indicator of data quality with higher values of the Index parameter indi- cating a higher confidence in the crystallographic orientation assigned by the software.An example of an area analyzed along with the orientation map is shown in Figure 2. Some boundaries between grains are slightly diffused due to the small grain size and grain overlap but in general the individual grains are clearly visible.The distribution of the Index parameter for each pixel in the five areas analyzed is shown in Figure 3. The frequency distributions of the thicker areas lie at higher values of the Index parameter compared to the thinner regions which makes it appear that thicker samples generally produce higher quality orientation data than thinner samples. However at some point grains overlap significantly resulting in the degradation of the orientation data and more diffused zones between grains.The critical thickness of the overlap is likely grain size dependent with that thickness decreasing with decreasing grain size. However the samples appear to be of sufficient overall thicknesses to produce 2014 ANNUAL REPORT 115 Table 1. Samples and expected outcomes. Sample Type Sample ID Description Sample Status Result to be Obtained Unirradiated Thickness 1 Effect of thickness on ASTAR 50mm To be FIB from unirradiatedTRISO particle fuel kernel removed at CAES Optimum sample thicknessbase- line graingrain boundary structure Thickness 2 Effect of thickness on ASTAR 100mm to be FIB from unirradi- atedTRISO To be FIB from unirradiatedTRISO particle fuel kernel removed at CAES Optimum sample thicknessbase- line graingrain boundary structure Thickness 3 Effect of thickness on ASTAR 150 mm to be FIB from unirradi- atedTRISO To be FIB from unirradiatedTRISO particle fuel kernel removed at CAES Optimum sample thicknessbase- line graingrain boundary structure Irradiated Compact 6-3-2 35-6B Particle 35 11 burnup longitudinal orientation low Ag release Already prepared and analyzed for compositional information Graingrain boundary structure perpendicular to the SiC growth direction close to IPyCSiC layer in low Ag release particle 35-transverse 1 Particle 35 11 burnup transverse orientation nearest kernel low Ag release Already prepared for compositional information Graingrain boundary structure parallel to the SiC growth direction closest to IPyCSiC interface in low Ag release particle 35-transverse 2 Particle 35 11 burnup transverse orientation middle SiC layer low Ag release Already prepared for compositional information Graingrain boundary structure parallel to the SiC growth direction from the middle of the SiC layer in low Ag release particle 35-transverse 3 Particle 35 11 burnup transverse orientation furthest kernel low Ag release Already prepared for compositional information Graingrain boundary structure parallel to the SiC growth direction closest to SiCOPyC interface in low Ag release particle Composition analysis is expected to be complete by the start of the RTE Nuclear Science User Facilities 116 high-quality orientation data.The effects of irradiation damage on the orientation data are not determined in this study and may be an area of consideration for future research work. Irradiation damage can introduce significant numbers of defects into the SiC lattice as well as generate residual stresses that may adversely affect the PED pattern and the resulting crystallo- graphic orientation determination e.g. lowering the Index parameter. Summary of Phase 2 resultsCrystal- lographic characterization of grain boundaries and triple junctions in the SiC layer by PED utilizingASTAR ASTAR data has been collected in every area previously analyzed with STEM and energy dispersive x-ray spectroscopy EDS on FIB lamellae from coated particleAGR1-632-035 enabling examination of the crystallo- graphic relationships on approximately 929 grain boundaries. Of these only 179 boundaries and triple junctions contained fission products and trans- uranic elements.Analyses of these grain boundary characteristics are in the final stages of interpretation and preliminary work already shows that using a high- angle annular dark-field scanningTEM HAADF-STEM is expedient to identify grain boundaries containing fission products. EDS was used to analyze at least qualitatively the composition of those fission products and PED was used to evaluate the grain boundary parameters in the SiC layer of irradiated TRISO particles. The final interpretation of these results and the integration of combined crystallographic and chemical analysis data still need to be completed to fully determine critical insight into the migration of fission products and transuranic elements through the SiC diffusion barrier layer.As an example of the level of information obtained one specific area Area B Sample IE 4 and an overview of the cumulative results from all areas are discussed in the following sections Area Bsample IEA total of 446 grain boundaries were analyzed in the five areas A to E of the IE sample Figure 4. However only results specific to the analysis of the grain boundaries inArea B are shown.The characteristics of these 51 grain boundaries were determined and reported in a 2015 summary paper for theAmerican Nuclear Society ANS 4. The crystallographic parameters of the grains surrounding the grain boundary fission product were collected using ASTAR. PED patterns were collected in a square grid with a 10 nm step size and spot size of 2014 ANNUAL REPORT 117 Figure 5. a Misorientation distribution of grain boundaries in Area B sample IE from particle AGR1-632-035. b A breakdown of grain boundary types 4. errors in orientation as determined by ASTAR.A plot of the misorientation angle distribution for grain bound- aries in Area B is shown in Figure 5a while the pie-chart in Figure 5b provides the relative fractions of the various types of grain boundaries in the scanned area.As mentioned the low-angle grain boundary fraction is very low.The misorientation distribu- tion of the high-angle grain bound- aries is centered around 35 degrees with a large fraction observed at 60 degrees which corresponds to the twin orientation. The large fraction of CSL-related grain boundaries indicated in Figure 5 b consist mostly of twins and higher-order twins e.g.3 699 10 and 27a and 27b 14.The remainder of the CSL-related boundaries consist of one 13a 3 and one 35a 3. Only 17 of the grain boundaries in the analyzed area have fission products associated with them.A summary of these boundaries is provided in Table 2 where both the boundary crystallography and the associated fission products are detailed. Only one CSL-related grain boundary exhibited the presence of a fission product that was exclusively palladium Pd. The rest of the boundaries with fission products are random high- angle grain boundaries containing only Ag PdAg or Pduranium U. However the uniqueness of the CSL- related boundary containing fission products is questionable considering its high value 35. Only one in 35 lattice sites between the two grains are coincident with a 35 CSL-related grain boundary. If this boundary lacks any special qualities relative to the other high-angle boundaries in Table 2 it can be concluded that fission products are only associated with random high-angle grain boundaries. Pd is the most prevalent fission product. It was found exclusively in five grain boundaries.Ag was found alone in only two grain boundaries and with Pd in one. U was not found alone and was associated with Pd in only one grain boundary. Only about 39 of the high- angle grain boundaries exhibited any presence of fission products at all.Thus some other parameter would seem to strongly influence the precipitation of fission products even in high-angle grain boundaries. All grain boundaries analyzed in sample IECtr and OEThe character- istics of 929 grain boundaries were determined in all the areas analyzed in samples IE Ctr and OE Figure 4. Full interpretation is in progress with reports or papers being prepared. Figures 6 and 7 show some of the collective results on all grain boundaries analyzed. It was found that 45.1 of all grain boundaries consist of high-angle boundaries that are not CSL bound- aries 49.2 are CSL boundaries and only 5.7 are low-angle boundaries Figure 6. Of all the boundaries analyzed only 11.1 contain fission products.The grain boundary type distributions are presented in Figure 7. Only 6.0 of the boundaries containing fission products contain Ag and Pd 14.3 contain only Ag 17.9 contain Pd and U and the remaining 61.9 contain only Pd.The 14.3 of grain boundaries containing only Ag are all high-angle grain boundaries while the grain boundaries containing Nuclear Science User Facilities 118 Table 2. Details of boundaries with fission products. Figure 6. Summary of grain boundary types in the 929 grain boundaries analyzed. Figure 7. Summary of fission products contained in the 929 grain boundaries analyzed. Location Misorientation Axis CSL Elements Angle x y z Pd Ag U 1 42.4 -7 12 -27 - 2 29.4 7 -26 0 - 3 49.5 1 -16 12 - 4 59.5 21 16 12 - 5 27.0 -22 -17 11 - 6 35.2 8 8 -5 - 7 32.7 -25 11 10 35a 8 39.7 -23 1 16 - 9 31.2 5 2 -13 - bothAg and Pd consist of four high- angle grain boundaries and one CSL grain boundary.This finding is now being pursued further as previous microstructural work does not show the presence ofAg on CSL boundaries. Understanding these phenomena could provide further clarity on the Pd-assisted Ag transport hypothesis proposed by Neethling et al. 5 Future Activities The goals of this research to continue are Publish a journal paper on theASTAR crystallographic results Interpret and integrate the results of the larger overarching project into a follow-up journal paper Continue work with the project collaborators to complete the larger project Submit a follow-up research proposal to expand the work on how irradiation damage effects the orientation data. Staff Exchanges The graduate intern and post-doctoral researcher were not directly involved in this particular study due to the timing of their appointments to INL June and August 2014 respectively. However this work formed the basis of the training that Dr HaimingWen received on the application collection and interpreta- tion of theASTAR system and he will continue his efforts on other samples as part of theAGR-1 experiment for 2015. References 1. I. J. van RooyenY.Q.WuT. M. Lillo T. L.Trowbridge J. M. Madden and D. Goran 2013Advanced Electron MicroscopicTechniquesApplied to the Characterization of Irradiation Effects and Fission Product Identi- fication of IrradiatedTRISOCoated Particles from theAGR-1 Experi- ment Global 2013Nuclear Energy at a Crossroads Salt Lake City September 29October 3 2013. 2014 ANNUAL REPORT 119 With its ability to quickly determine grain boundary parameters at the nanometer scalewe expectASTAR will be a key tool in facilitating a full understanding of the fission product transport mechanism in the SiC layer ofTRISO nuclear fueland ultimately allow the development of strat- egies to minimize or mitigate the release of fission products Dr.Tom LilloResearch Scientist EngineerIdaho National Laboratory Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite Idaho National Laboratory Materials and Fuels Complex Electron Microscopy Laboratory Collaborators Idaho National Laboratory Isabella van Rooyen principal investigator Tomas Lillo co-principal investigator James Madden collaborator Haiming Wen post- doctoral researcher non U.S. Boise State University Yaqiao Wu co-principal investigator John Youngsman co-principal investigator Idaho State University Connie Hill graduate intern 2. I. J. van RooyenY. Q.WuT. M. Lillo J.Youngsman J. H. Neethling 2014 Approach and Micro-Anal- ysisTechniques Applied to Study Fission ProductTransport Mecha- nisms in Neutron-Iirradiated SiC Layers XXIII International Materials Re- search CongressCancunMexicoAugust 17-21 2014. 3. I. J. van Rooyen J.YoungsmanT. M. LilloY. Q.Wu D. Goran M. E. Lee W. E. Goosen J. H. NeethlingT. L. Trowbridge J.W. Madden 2014 Methods for Identification of Crys- tallographic Parameters of Irradiated SiC to Understand Fission Product Transport The 3rdWorkshop on High Temperature Gas-Cooled Reactor SiC Material Properties Jeju Island South Korea Sept. 30Oct. 1 2014. 4. T. M. Lillo I. J. van RooyenY. Q.Wu 2015Grain Boundary Character and Fission Product Precipitation in SiC 2015American Nuclear Society Annual Meeting SanAntonioTX June 711 2015. 5. J. H. Neethling J. H. OConnell and E. J. Olivier 2012Palladium assisted silver transport in polycrystalline SiC Nuclear Engineering and DesignVol. 251 pp. 230-234. Publications and Presentations 1. I. J. van RooyenY. Q.WuT. M. Lillo J.Youngsman J. H. Neethling 2014 Approach and Micro-Analysis TechniquesApplied to Study Fission ProductTransport Mechanisms in Neutron- Iirradiated SiC Layers XXIII International Materials Research Con- gressCancunMexicoAugust 17-212014. 2. I. J. van Rooyen J.YoungsmanT. M. LilloY. Q.Wu D. Goran M. E. Lee W. E. Goosen J. H. NeethlingT. L. Trowbridge J.W. Madden 2014 Methods for Identification of Crys- tallographic Parameters of Irradiated SiC to Understand Fission Product Transport The 3rdWorkshop on High Temperature Gas-Cooled Reactor SiC Material PropertiesJeju IslandSouth Korea Sept. 30Oct. 1 2014. 3. T. M. Lillo I. J. van RooyenY. Q.Wu 2015Grain Boundary Character and Fission Product Precipitation in SiC 2015 American Nuclear Society Annual MeetingSan AntonioTXJune 7112015. Nuclear Science User Facilities 120 Microstructure Evolution in Ion-Irradiated Uranium Dioxide Mahima Gupta University of Wisconsin Madison mahimawisc.edu Understanding irradiation damage evolution in uranium dioxide UO2 is crucial to protecting the worlds fleet of current and future nuclear power plants. Since all of the phenomena caused by radia- tion damage originate at point defects understanding the effects of irradiation at the atomic scale is crucial. However because the irradiation defects are aperi- odic standard approaches such as trans- mission electron microscopy TEM and x-ray diffraction XRD are ineffective necessitating the use of techniques that are sensitive to short-range order. X-ray absorption fine-structure spectroscopy XAFS measures the population- weighted local structures and chemical speciation of the examined elements making it perhaps the most incisive method for determining the local-range order in irradiated materials. TEM measurements are crucial to relating the short-range changes observed using extended X-ray absorp- tion fine-structure spectroscopy EXAFS to the complicated long-range micro- structures created through irradiation. Understanding these defects at various length scales is necessary for accurately predicting fuel degradation under reactor conditions. Project Description The goal of this experiment is to study irradiation-induced damage structure evolution in UO2 on multiple-length scales.The main objective is to combine spatially resolved synchro- tron-based X-ray absorption spectros- copy XAS withTEM observations of ion-irradiated depleted UO2. This study uses XAFS observations combined with microscopy tech- niques performed at CAES to examine lamellae having spatially varied irradiation damage. Ion irradiations that produced spatially varied micro- structures were performed at 150C on depleted UO2 samples at the University of Wisconsin Madisons Ion Beam Laboratory.The damage profiles in the helium He-implanted samples were about 10 m and were as flat as possible for the first 15 m. With the completion of these experiments the targeted research has increased our understanding of fuel degradation and added to the knowl- edge base of the nations nuclear infra- structure.The benefits of knowing how the local structure of ion-irradiated UO2 evolves extend beyond nuclear fuel operations into long-term storage. Being able to predict the stability of The first demonstration of micro-focused x-ray absorption fine-structure measurements on FIB UO2 lamellae was successful clearing the way for analysis of highly radioactive neutron-irradiated samples at synchrotron facilities. reactor-irradiated UO2 is crucial to determining what chemical reactions will occur while a fuel is being stored. Accomplishments All the milestones described in the ATR-NSUF rapid-turn-around proposal have been met.A technique to perform -EXAFS measurements has been developed at the Stanford Synchrotron Radiation Lab SSRL and a focused ion beam FIB lamella consisting of the damage layer created by injecting He ions into bulk UO2 samples was extracted at CAES. Figure 1 compares the extracted FIB lamellae Figure 1.1 to the overlay of the ion-damage profile created using Stopping Range of Ions in Matter Figure 1.2. Five such lamellae two each from the two irradiated samples and one from the pristine sample were mounted on an FIB grid.The grid was then inserted into a -EXAFS holder specially designed for the SSRL.Two of these samples are shown in Figure 1.2.They were measured at a 45-degree angle from the incident beam in fluorescence geometry. Further bulk EXAFS analysis was performed on krypton-implanted samples at an angle of 10 degrees from the incident beam which interrogated the irradiated region to a depth of 1. 2014 ANNUAL REPORT 121 Figure 1. 1 The extracted FIB lamellae 2 the overlay of the ion-damage profile created using Stopping Range of Ions in Matter. Nuclear Science User Facilities 122 In this experimentwere learning that identifying the differences between oxygen doping and irradiation isnt so simple.There may even be differences between the proton- and He-irradiated materialsand we want to make sure we get it right. Steve ConradsonBeamline ResponsibleSynchrotron SOLEIL current position Figure 2. k3 R EXAFS of Ref UO2 UO2 irradiated to 0.006 dpa with He2 ions and UO2 irradiated to 0.035 dpa with He2 ions. Plots 1-3 the modulus of the real part of the transform of both data and fit. Insets top k3 spectra overlaid with curve-fit bottom moduli of the data fit difference between data and fit and the individual contributions to the fit inverted for clarity. Plot 4 modulus of the Fourier transforms of the k3 -weighted EXAFS spectra of indicated samples. Transforms were performed over a range of 2.70 to 14.75 -1 . 1 m. He-implanted samples irradi- ated to 0.0060.035 displacements per atom dpa and proton-irradiated samples 0.010.5 dpa were also successfully studied using EXAFS at SSRL.This was a first-of-its-kind demonstration of -EXAFS measure- ments on FIB lamellae and brought together high-energy x-ray studies from synchrotron sources and the microscopic analysis techniques used in material characterization. EXAFS analysis has shown that irradia- tion disrupts local structure by giving rise to multisite oxygen distribution at approximately 1.9 angstroms from the absorbing atom which is 0.2 farther than in documented UO2x .This is similar to oxygen interstitials created in UO2x however there is no observed oxidation of the material.The multisite distribution results from uranyl-type bonds that are 1.8 in length and extremely stable due to their oblate geometry which causes them to distort the original local structure. The EXAFS data from the irradiated UO2 shows a consistent decrease in amplitude in the crystallographic shells as irradiation doses increase longer bond distances between near neighbors and the appearance and increase of a shoulder on the low R side at 1.9 of the first crystal- lographic U-O shell. Random disorder in the material would result in a significant loss of amplitude and addi- tional loss of overall structure which is not seen in this set of samples. Due to the consistent amplitude reduction as shown in Figure 2 it can be theo- rized that the defects are not random but instead cluster in such a way that the overall lattice retains its UO2 struc- ture even at higher irradiation doses. This is consistent with several studies that indicate fluorite ceramics in a reactor maintain their lattice structure even at high irradiation doses. 2014 ANNUAL REPORT 123 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor Center for Advanced Energy Studies Microscopy and Characterization Suite University of Wisconsin Madison PIE facilities Collaborators University of Wisconsin Madison Mahima Gupta principal investigator Janne Parkarinen collaborator Idaho National Laboratory Todd Allen co-principal investigator Jian Gan collaborator Los Alamos National Laboratory Steve Conradson collaborator Future Activities The most important future goal for the project is the full characteriza- tion of -EXAFS data from the FIB UO2 samples. Processing this data has its own challenges due to the lower signal strength microscopic samples emit.Analysis is underway and is to be completed by March 2015. A subsequent ATR NSUF rapid turnaround proposal to continue the research was submitted and accepted. Samples of depleted UO2 were prepared using polishing techniques at CAES and were sent to the University of Wisconsin Madison for proton implantation at its ion beam facility which initiated radiation damage in the 0.20.3 dpa range.The irradiated samples were mounted on specially designed sample holders for local structure and speciation analysis at the Advanced Photon Source.These samples will be used to extract FIB lamellae from the irradiation damage layer and their microstructure evolu- tion will be studied using theTEM at the Microscopy and Characterization Suite at CAES. The goal of this experiment is to bridge the gap between existing information on lattice structure and microstructure evolution under proton irradiation in UO2. Preparation of samples for irradiation at ATR is currently under way.These irradiations will provide a detailed understanding of the evolution of the material struc- ture in UO2 under reactor irradiation. Publications and Presentations 1. M. Gupta 2014 Damage Struc- ture Evolution in Ion Irradiated UO2 TMS conferenceSan DiegoCA February 16202014. 2. M. Gupta 2014 Atomic Struc- ture Evolution in Ion Irradi- ated UO2 presentation for Dr. Franklin Lynn Orr Undersec- retary for Science and Energy May 19th 2014. 3. M. Gupta 2014 Defects in Ion Irradiated UO2 on Multiple Length ScalesATEM and XAFS Study presentation for Dr. Patri- cia Dehmer Deputy Director for Science Programs in the Office of Science Center for Advanced Energy StudiesAugust 12 2014. 4. M. Gupta 2014 Defects in Ion Irradiated UO2 on Multiple Length ScalesATEM and XAFS Study presentation for Dr. Ernest Moniz US Energy Secretary Cen- ter for Advanced Energy Studies Idaho Falls IDAugust 19 2014. 5. M. Gupta S. Conradson J. Pakar- inen andT.Allen 2015 Identi- ficatio3n of Collective Effects in He2 irradiated UO2 via Extend- ed X-ray Absorption Fine Struc- ture Spectroscopy submitted to the Journal of Inorganic Chemistry March 2015. Nuclear Science User Facilities 124 Radiation Induced Segregation in Nickel-Chromium Alloys Janne Pakarinen University of WisconsinMadison jpakarinsckcen.be Nickel Ni-based alloys are employed in current light water reactors LWR though typically not as in-core components. They are also candidate materials for molten salt reactor applications due to their excellent corrosion resistance in fluoride salt systems.As such understanding how radiation damage influences their microstructures and performance in these systems is crucial to the development of safe and reliable molten salt reactor technologies. Project Description This project will investigate the micro- structural effects of ion irradiation in Ni-based alloys with a focus on the manifestation of voids dislocation loops and radiation-induced segrega- tion RIS.The formation of these features in binary Ni-Chromium Cr alloys is being studied as a function of three experimental variables Cr composition 5wt Cr 18wt Cr and 33wt Cr Irradiation temperature 400C and 500C Irradiating species proton vs. Ni-ion irradiation These variables are characterized using analytical transmission electron microscopy TEM techniques to show the expected radiation responses of these alloys in a nuclear reactor system as well as to provide experimental data to be used as benchmarks in the development of predictive models for the formation of these features. Figure 1. Dislocation loop density and size distribution in 500C proton- irradiated Ni-Cr. Understanding the radiation responses of Ni-based alloys is vital to developing reactor core components for next-generation molten salt technologies. 2014 ANNUAL REPORT 125 Accomplishments Prior to 2014 both Ni-5Cr and Ni-18Cr samples were irradiated with both protons and Ni-ions at 500C in the University ofWisconsin UW Tandem Accelerator Ion Beam. In early 2014TEM sample prepara- tion and preliminary analysis of the 500C proton-irradiated materials was performed in the CAES Microscopy and Characterization Suite MACS using the Quanta 3D Focused Ion Beam FIB andTecnaiTF-30TEM.AdditionalTEM characterization was also performed at the University ofWisconsin Materials Science Center MSC on theTecnai TF-30TEM and the FEITITANTEM. TheseTEM observations characterized and compared the density and size distributions of voids and dislocation loops in the samples Figures 1 and 2. Abnormal RIS behavior was also observed at 500C and was thought to be due to grain boundary migration effects Figure 3.This prompted the 400C proton irradiation in which the Ni-33Cr model alloy was also included. In addition to providing another composition with which to compare the formation of these microstruc- tural features Ni-33Cr was added to investigate whether or not irradiation could induce a low-temperature stable ordered phase which is present in this material at that composition. The second proton irradiation was also performed at the UWTandem Accelerator Ion Beam and the sample preparation and preliminary analysis were again performed at CAES-MACS. Figure 2. Void density and size distribution in 500C proton-irradiated Ni-Cr. Nuclear Science User Facilities 126 Figure 3. Typical RIS profile observed in 500C proton-irradiated materials. A large Cr-depleted region is observed that is also denuded in voids and dislocations. The FIB andTEM capa- bilities at CAES are world classand provide means for productive sample fabrication and analysis.The rapid turn- around system fit the needs of this project perfectly. Dr.Janne PakarinenVisiting ScientistUniversity ofWisconsin Madison 2014 ANNUAL REPORT 127 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Center for Advanced Energy Studies Microscopy and Characterization Suite University of Wisconsin Madison Tandem Accelerator Ion Beam Sandia National Laboratory Ion Beam Laboratory Drexel University Central Research Facilities Collaborators University of Wisconsin - Madison Janne Pakarinen principal investigator Samuel Briggs co-principal investigator Leland Barnard co-principal investigator Kumar Sridharan co-principal investigator Dane Morgan co-principal investigator Mitra Taheri co-principal investigator Christopher Barr co-principal investigator Simultaneously during 2014 Ni-ion irradiation and post-irradiation exami- nation PIE were being performed by Christopher Barr at Drexel University. These samples were irradiated with 20 MeV Ni-ions at the Sandia National Laboratory Ion Beam Laboratory. Sample preparation and additional analysis was performed at the CAES-MACS facility. Future Activities TEM characterization of the 400C proton-irradiated materials is ongoing. Additionally characterization of dislocation loop density in the 500C Ni-irradiated materials is still in progress. Completion ofTEM analysis on these two alloy systems will allow us to compare the effects of composition temperature and irradiating species on the micro- structure resulting from irradiation. Three papers are currently planned for this project in 2015.The first will compare and contrast the resulting microstructures in the binary Ni-Cr system as functions of composition irradiating temperature and irradi- ating species.The second will discuss the unusual RIS profiles observed in the 500C irradiation.The third will discuss the viability of inducing low-temperature stable ordered phases using short ion irradiations as opposed to long-term thermal annealing experiments. Publications and Presentations 1. S. Briggs J. Pakarinen L. Barnard D. D. MorganT. R.Allen and K. Sridharan 2014 Radiation- Induced Segregation and Other Radiation-Induced Effects in NiCr Alloys Materials Science Technology 2014PittsburghPAOct.12162014. 2. S. Briggs J. Pakarinen L. Barnard D. D. Morgan I. SzlufarskaT. R. Allen and K Sridharan 2014 Study of Radiation-Induced Seg- reation Using Nickel-Chromium Binary Alloys TMS conferenceSan DiegoCAFebruary 16202014. Nuclear Science User Facilities 128 Ion Irradiation of Nuclear Grade NBG-18 and Highly Ordered Pyrolytic Graphites K. Linga Murty North Carolina State University murtyncsu.edu Figure 1. MeV carbon- implantation and damage profiles calculated using SRIM 2012.03 software. By developing mechanistic thermo-mechanical models for graphite that can be applied to next-generation high-temperature reac- tors our studies address the industrys general lack of clarity on the damage mechanisms of graphite.To this end the project was conceived to investigate the radiation damage on highly oriented pyrolytic graphite HOPG and NBG-18 through ion irradiation at the University ofWisconsin Madisons UWTandem Accelerator Ion Beam. Project Description The general intent of the ion-irradiation experiments at UW was to characterize the early-to-late-stage damage mecha- nisms in graphite under irradiation. This follows the low-dose neutron- irradiation studies conducted in the PULSTAR reactor at North Carolina State University NCSU and the high-dose studies carried out at the Oak Ridge National Laboratory ORNL. 2014 ANNUAL REPORT 129 Figure 2. Temperature and beam-current profiles during irradiation at 900 K 25 dpa and 2.0 MeV C . Temperature K dpa Flux ionscm2 .s 300 1 1.31013 600 1 1.11013 900 1 1.31013 600 25 1.41013 300 25 1.21013 Table 1. Ion irradiation cases. This project seeks to clarify our understanding of the damage mechanisms of graphite. Ion Irradiation Condition IDIG ID ID 300 K1dpa 0.87 2.54 600 K1 dpa 0.58 4.47 900 K1 dpa 0.84 2.18 600 K25 dpa 1.38 1.79 900 K25 dpa 0.80 2.43 Table 2. Raman analysis for ion- irradiated NBG-18 samples. Ion Irradiation Condition IDIG 300 K1dpa 0.02 600 K1 dpa 0.01 900 K1 dpa 0.06 600 K25 dpa 0.58 900 K25 dpa 0.25 Table 3. Raman analysis for ion- irradiated HOPG samples. Nuclear Science User Facilities 130 Figure 3. Temperature and beam-current profiles during irradiation at 900 K 1 dpa and 2.0 MeV C . Figure 4. Temperature and beam-current profiles during irradiation at 600 K 25 dpa and 2.0 MeV C . Figure 5. Temperature and beam-current profiles during irradiation at 600 K 1 dpa and 2.0 MeV C . The team at the Univer- sity ofWisconsin has been methodicalinsightful and extremely professional throughout this project. Jacob EapenAssistant ProfessorAssociate Professor Nuclear EngineeringNorth Carolina State University 2014 ANNUAL REPORT 131 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities University of Wisconsin - Madison Tandem Accelerator Ion Beam Collaborators University of Wisconsin Madison K. Linga Murty principal investigator Jacob Eapen co-principal investigator Ram Krishna co-principal investigator Figure 6. Temperature and beam-current profiles during irradiation at 300 K 1 dpa and 2.0 MeV C . Accomplishments Ion-irradiation experiments were conducted on the HOPG and NBG-18 samples at displacements per atom dpas of 1 to 25 and tempera- tures ranging from 300 to 900 K. Damage was initiated with C ions at 2-MeV energy.The fluences at 25 dpa and 1 dpa are 2.21017 ions cm2 and 8.81015 respectively.The test durations for the 25-dpa and 1-dpa samples were 4.5 hours and 12 minutes respectively.The average displacement energy was 28 eV.Table 1 shows details of the irradiations at different temperatures and dpas. Figure 1 shows the 2-MeV carbon- implantation and damage profiles calculated using Stopping and Range of Ions in Matter SRIM 2012.03 software and assuming a displace- ment-threshold energy of 28 eV and a mass density of 2.253 gcm3 . As indicated in the plot 25 dpa peak corresponds to a fluence of 2.21017 Ccm2 13 at. of carbon in the peak.Temperature and beam-current profiles are shown in Figures 2 through 6.Tables 2 and 3 show the ratio of the peaks for each sample. Publications and Presentations 1. J. Eapen R. KrishnaT. D. Burchell K. and L. Murty 2014 Early Damage Mechanisms in Nuclear Grade Graphite Under Irradia- tion Materials Research Letters Vol. 2 pp. 4350. Nuclear Science User Facilities 132 Transmission Electron Microscopy Study of the Microstructure Evolution in Kr-Irradiated UO2 Lingfeng He Idaho National Laboratory lingfeng.heinl.gov In-situ transmission electron microscope TEM observation of uranium dioxide UO2 irradiated with krypton Kr ions at the Inter- mediateVoltage Electron Microscope IVEM-Tandem facility at Argonne National Laboratory has shown the evolution of the basic radiation defects of dislocation loops and bubble formation. However theTEM foil samples used for in situ observation were too thin and therefore unsuitable for thermal transport measurement. Project Description In this rapid turnaround work we are investigating defect production and dislocation loop and bubble evolu- tion in UO2 under ex-situ 1.8 MeV Kr irradiation.We are also providing theoretical modeling support that connects the materials microstructure to its thermal transport properties for a parallel project being conducted at INL. This experiment will improve our understanding of microstructure evolution in UO2 under irradiation. Figure 1. Intragranular bubble size and density as a function of the annealing temperature. 2014 ANNUAL REPORT 133 Accomplishments In this work we investigated bubble and defect evolution in Kr-irradiated and high-temperature-annealed UO2 . The results showed bubble formation at room-temperature irradiation. Bubbles grew gradually at the annealing temperature while their densities decreased gradually Figure 1. A bubble denuded zone in the vicinity of the grain boundaries was found Figure 2 which indicates that the Kr Figure 2. Cross-section TEM images showing bubbles near grain boundaries in Kr-irradiated polycrystalline UO2 after annealing at 1000C a and d 1300C b and e and 1600C c and f for one hour. Figures a-c are under-focus images and Figures d-f are over-focus images. diffused towards the grain boundaries during high-temperature annealing. Grain boundaries contained more vacancies than the bulk regions and they proved to be good reservoirs for inert Kr gas resulting in the rapid growth of intergranular bubbles at grain boundaries Figure 3. In addition high-angle grain bound- aries have more room for inert gas compared to low-angle grain bound- aries allowing the formation of larger Nuclear Science User Facilities 134 Figure 3. Cross-section scanning transmission electron microscope STEM images showing the evolution of intergranular bubbles during post-irradiation annealing at a and e 25C b and f 1000C c and g 1300C and d and h 1600C. Figures a-d are low-magnification images and figures e-h are high-magnification images.The dashed line in c indicates the bubbles distance from the samples surface. Figure 4. a Electronic backscatter diffraction EBSD map of Kr-irradiated polycrystalline UO2 annealed at 1300C for one hour. Arrows mark a high-angle grain boundary between Grain 1 and Grain 2 GB1 and a low-angle grain boundary between Grain 3 and Grain 4 GB2. b Cross-section STEM images comparing the bubble distribution at GB1 and GB2. 2014 ANNUAL REPORT 135 Figure 5. a Cross-section STEM images showing the dislocations in as-irradiated and annealed polycrystalline UO2 at 1300C for one hour. b Dashed lines show the damage profiles in displacements per atom dpa calculated by stopping-and-range- of-matter-in-ions SRIM software. bubbles Figure 4.The small defects induced by Kr-ion irradiation also grow at the annealing temperature. This results in the formation of three distinguished zones that appear at various distances from the surface of the sample Figure 5 a defect denuded zone at the surface a disloca- tion network zone farther from the surface and a dislocation loop zone at the deep region. Future Activities The project has revealed bubble evolu- tion as a function of annealing tempera- ture however single-bubble pressure in annealed samples of UO2 has never been measured. Examining the relationship between bubble pressure bubble size and annealing temperature will be the primary focus of our work in 2015. Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor PIE facilities Argonne National Laboratory IntermediateVoltage Electron Microscope IVEM-Tandem facility Collaborators University of Wisconsin - Madison Lingfeng He principal investigator Janne Pakarinen collaborator Idaho National Laboratory Jian Gan co-principal investigator Boise State University Darrell Butt collaborator State-of-the-art techniques have led to cutting-edge research on nuclear materials. Lingfeng HeAssistant ScientistUniversity of Wisconsin-Madison Currently Nuclear Fuels EngineerIdaho National Laboratory Publications and Presentations 1. B.Valderrama L.F. He H.B. Hen- derson J. Pakarinen B. Jaques J. Gan D.P. ButtT.R.Allen and M.V. Manuel 2014Effect of Grain Boundaries on Krypton Segrega- tion Behavior in Irradiated Ura- nium Dioxide JOMVol. 66 pp. 25622568. 2. L.F. He J. Pakarinen J.GanA.T. Nel- son B. Jaques D. ButtA. El-Azab andT.R.AllenBubble Evolution in Kr-irradiated UO2 under Post-irra- diationAnnealing Journal of Nuclear Materials to be submitted. Nuclear Science User Facilities 136 Irradiation and Post-Irradiation Examination of Alloys X-750 and XM-19 Electric Power Research Institute Pilot Program Phase III Cooperative Research and Development Agreement No. 12-CR-06 John H. Jackson Idaho National Laboratory john.jacksoninl.gov As a means of establishing a basis for the development and execution of joint ATR NSUF industry programs the Electric Power Research Institute EPRI and ATR NSUF have developed a pilot program involving shared costs and responsi- bilities. In addition to providing data this EPRI pilot project is designed to Develop administrative protocols for the projects research such as cooperative agreements and funding. Develop portions of the research capa- bilities and staffing required to address future research and development needs. Develop a protocol for validating data with industry particularly stress cor- rosion crack SCC growth rate data. This project is important for three reasons first it is the initial industry pilot project for ATR NSUF and estab- lishes protocols for these types of proj- ects second it is a full cradle-to-grave characterization of reactor internal material including baseline character- ization irradiation and post-irradiation examination PIE third it is the first project to utilize two newly installed tools the controlled water chemistry Loop 2A in the ATR center flux trap and the irradiation-assisted stress corrosion crack IASCC test systems. Project Description Discussions between ATR NSUF and EPRI resulted in a decision to focus on investigation of the fracture toughness and IASCC growth rates of irradiated high-strength alloys used for boiling water reactor BWR repair hardware. Very little of this data exist for the nickel Ni-based alloy X-750 or for XM-19 a nitrogen N-strengthened austenitic stainless steel at the exposure levels of interest up to 1x1021 ncm2 . The focus of this EPRI pilot project is on irradiation and characterization of these alloys in both un-irradiated baseline and irradiated states and is being conducted in three phases. In Phase I researchers completed the fabrication of specimens from mate- rials provided by EPRI and established the baseline fracture toughness and crack growth rates CGR of un-irra- diated material. In Phase II they completed the design and fabrication of specimen holders and performed a safety analysis on a test train in order to meet EPRI objectives for irradiation of tensile and compact tension speci- mens utilizing Loop 2A in the center flux trap of ATR Figure 1. In the current and final Phase III researchers are performing irradiation and PIE of the EPRI specimens previously tested in their unirradiated state. This is the first project to utilize two newly-installed tools Loop 2A in the ATR center flux trap and the irradiation stress corrosion cracking IASCC test systems. Figure 1. Repair hardware of interest for this study. 2014 ANNUAL REPORT 137 Figure 2.Typical baseline SCC test results for alloy X-750. CGR is the dark blue line. It is very exciting to see this program bearing fruit. The completion of the IASCC testing rigs and their use in determining crack growth rates of the irradiated materials marks a significant milestone and provides INL with a key capability for the future. Bob CarterTechnical ExecutiveEPRI Figure 3. Typical fracture toughness test results for alloy X-750 at elevated 289C temperature. Nuclear Science User Facilities 138 Figure 4. Typical irradiation test train containing CT tensile and TEM specimens. Accomplishments During this project several baseline SCC tests were completed on alloys X-750 and XM-19. Cold-worked 9.3 XM-19 was utilized to study potentially similar material property effects between neutron embrittlement phenomena and embrittlement induced by cold-working the material. In addition to SCC tests fracture toughness tests at temperature not in environment and tensile tests at temperature not in environment were conducted. Results of the SCC tests were compared to data produced at the General Electric Global Research Company as a means of benchmarking INLs capability to perform these highly specialized experiments. In all cases the measured crack growth rates from the INL tests compared favorably with those produced by the benchmark laboratory. Figure 2 shows typical SCC test results for alloy X-750 and Figure 3 shows a typical fracture toughness test for alloy X-750. Results of the XM-19 tests are not included here for the sake of brevity but may be found in INL Report INL EXT-11-24173 Revision 1 or in the EPRI BWRVessel Integrity Program Report 1025135 on the EPRI website. Phase II of this project involved exper- iment physics and safety analysis as well as test train design for irradiating the specimens.A report was issued in February 2012 showing the results of these analyses and descriptions of the test train designs Figure 4. The first irradiation was completed during 2013 on the EPRI-2 capsule with a target fluence of 2.0 x 1020 ncm2 E1MeV in Loop 2A of the ATR center flux trap.Actual measured fluence for this capsule was 1.93 x 1020 ncm2 . The capsule was shipped from ATR to the Hot Fuels Examination Facility HFEF following irradiation and cool down and was disassembled in preparation for PIE. In early 2014 the first of two irradia- tion cycles for the EPRI-3 capsule at a target fluence of 1.0 x 1021 ncm2 was completed also using Loop 2A in the ATR center flux trap. Complica- tions arising from a flow restriction have delayed completion of the second irradiation cycle but it is expected to be completed in 2015 along with the lowest fluence 5.0 x 1019 ncm2 irradiation of the EPRI-1 capsule. The very first test on the newly constructed IASCC test rig system was also conducted during 2014 on an X-750 specimen that was irradiated in the EPRI-2 capsule. This represented an exciting milestone for the project and for INL.Very stable crack propa- gation was measured and excellent chemistry control was exhibited in the new systems. The measured crack growth rates in this irradiated X-750 equal to nominally 3.1 6.2 x 10-8 mms in hydrogen water chemistry HWC and 1.1 22 x 10-7 mms in normal water chemistry NWC were comparable to CGRs that were previously measured in the same alloy in an un-irradiated state. Figure 5 shows a typical CGR plot for this very first IASCC test demonstrating the system chemistry control in switching between HWC and NWC. 2014 ANNUAL REPORT 139 Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor Hot Fuel Examination Facility Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators Idaho National Laboratory John H. Jackson principal investigator SebastienTeysseyre co-principal investigator Electric Power Research Institute Robert Carter co-principal investigator Peter Chou co-principal investigator Future Activities Fracture toughness tests are planned for the first part of 2015 and will be combined with the results of the initial IASCC tests in order to determine if there is any benefit to testing the EPRI-1 specimens at a lower fluence.The IASCC results suggest that even at the medium EPRI-2 fluence the material properties have not been affected. EPRI-1 will be irradiated to a target fluence of 5.0 x 1019 ncm2 in theATR center flux trap as originally planned 5 day cycle and then will be set aside pending fracture toughness test results from EPRI-2. EPRI-3 with a target fluence of 1.0 x 1021 ncm2 will continue an additional cycle of irradia- tion and should be ready for shipment to HFEF in late 2015. Specimens from these test capsules are scheduled for testing in 2015 and 2016. Testing on the EPRI-2 specimens through the first part of 2015 will include fracture toughness and confir- matory IASCC tests for both X-750 and XM-19. If the EPRI-3 irradiation is completed during that time frame those specimens may be tested as well. Publications and Presentations 1. C. R.Tyler P. E. Murray and J.W. Nielsen 2014 CRADA EPRI Phase II Design Report INLLTD 12 24400 Rev. 1 February 2014. See additional publications from other years in the Media Library on the NSUF website. Figure 5. IASCC test results for X-750 irradiated to 1.93 x 1020 ncm2 . CGR is the dark blue line. Nuclear Science User Facilities 140 Irradiation and Post-Irradiation Examination to Investigate Hydrogen- Assisted Anomalous Growth in Zirconium Alloys Paul Murray Idaho National Laboratory paul.murrayinl.gov Zirconium Zr alloy specimens are being irradiated inATR to study the mechanisms of irradiation- induced growth and its dependence on hydrogen H content and neutron fluence.The H produced by corrosion and dissolved in Zr alloys during service in light water reactors LWR can form hydrides and may promote irradiation growth. Differential strain resulting from H-assisted irradiation growth is postulated to be partly responsible for fuel channel bowing observed in boiling water reactors BWR. Project Description The objective of this project is to irradiate 200 specimens of various Zr alloys with various H concentra- tions in ATR up to four different neutron fluence levels.The change in length of the irradiated specimens will be measured to determine the irradiation-induced growth strain and transmission electron microscopy TEM will be performed to study the mechanism of irradiation growth. The knowledge gained from this experiment will be used to select Zr alloys that exhibit small growth strain which will improve the reliability of commercial reactors by minimizing fuel channel distortion caused by irradiation growth. Accomplishments The irradiation is taking place in an inert environment at a temperature of 285C and at four neutron damage levels expressed in terms of displace- ments per atom dpa. Four sets of 50 identical specimens were placed in four separate irradiation capsules identified as A B C and D. Capsule A completed irradiation to 6.7 dpa in January 2013 Capsule B completed irradiation to 12.3 dpa in January 2014 and Capsules C and D are currently being irradiated with a goal of reaching approximately 20 and 30 dpa respectively Figure 1. CapsulesA and B were subsequently transferred to INLs Hot Fuel Examina- tion Facility HFEF for post-irradiation length measurements.Those were completed for CapsuleA in September 2013 and for Capsule B inAugust 2014. An instrument was designed and built at INL to measure the length of irradiated specimens in HFEF using remote manipulators. It was tested and validated using specimens of known length.The geometric Figure 1. Measured growth strain of selected Zr alloys as a function of dpa normalized using the maximum growth strain. This experiment will provide the data needed to select Zr alloys that exhibit small growth strain which will improve the reliability of commercial reactors by minimizing fuel channel distortion caused by irradiation growth. 2014 ANNUAL REPORT 141 profiles of the irradiated specimens from Capsules A and B were then measured and the results analyzed to determine the average growth strain of each specimen. In addition to strain measurements other objectives included performing as-run reactor physics analysis to determine dpa measuring neutron fluence using flux wires installed in the experiment performing heat transfer calculations to determine irradiation temperature and obtaining temperature indica- tions from monitors melt wires and silicon-carbide installed in the experiment.These objectives have also been completed for Capsules A and B. Future Activities The growth strain data at 6.7 dpa CapsuleA and 12.3 dpa Capsule B are being evaluated in 2015 to ascertain the effects of neutron fluence alloy composition and H content on growth strain.The results will be given to the Electric Power Research Institute EPRI and shared with private-sector fuel suppliers AREVA Global Nuclear Fuels andWestinghouse that provided some of the Zr alloys used in this experiment. TEM will be used to analyze selected irradiated specimens to study the mech- anism of irradiation growth. Procedures for preparing specimens forTEM analysis are currently being developed at the irradiated materials characterization facilities at INL and include electro- polishing and ion-polishing. Moreover this project will continue to develop the research capability and staffing required to meet industry needs for research and development of nuclear materials. Publications and Presentations 1. P. E. Murray J. R. Parry J. H. Jackson 2014 Interim Report on CapsuleA of the ZirconiumAlloy Irradiation Growth Experi- ment in theAdvancedTest Reactor. INLLTD 14 31110 Rev. 1 June 2014. 2. P. E. Murray J. R. Parry J. Navarro 2014 Interim Report on Capsule B of the ZirconiumAlloy Irradiation Growth Experi- ment in theAdvancedTest Reactor. INLLTD 14 33859 December 2014. See additional publications from other years in the Media Library on the NSUF website. This project demonstrates that the nuclear industry can benefit from INLs capability in nuclear materials testing. Dr.Paul MurrayINL Distributed Partnership at a Glance ATR NSUF and Partners Facilities and Capabilities Idaho National Laboratory AdvancedTest Reactor Hot Fuels Examination Facility Center for Advanced Energy Studies Microscopy and Characterization Suite Collaborators Idaho National Laboratory Paul Murray principal investigator Electric Power Research Institute SureshYagnik collaborator Nuclear Science User Facilities 142 2014 ANNUAL REPORT 143 AES............................................................................................................................. auger electron spectroscopy AFCI........................................................................................................................Advanced Fuel Cycle Initiative AGR...................................................................................................................................... Advanced Gas Reactor ANL..........................................................................................................................Argonne National Laboratory ANS............................................................................................................................... American Nuclear Society APS.................................................................................................................................. Advanced Photon Source APT.................................................................................................................................. atom probe tomography ASCII............................................................................... American Standard Code for Information Interchange ATLAS................................................................................................. ArgonneTandem Linac Accelerator System ATR.......................................................................................................................................AdvancedTest Reactor ATRC.......................................................................................................................AdvancedTest Reactor Critical BF...........................................................................................................................................................bright field BWR...................................................................................................................................... boiling water reactor CAES.............................................................................................................. Center for Advanced Energy Studies CASS..........................................................................................................................cast austenitic stainless steels CG.............................................................................................................................................. conventional grain CGR............................................................................................................................................. crack growth rate CSL......................................................................................................................................coincidence site lattice CVD...............................................................................................................................chemical vapor deposition DMSA............................................................................................................................... dimercaptosuccinic acid DOE...................................................................................................................................... Department of Energy DOE-NE.......................................................................................Department of Energy Office of Nuclear Energy dpa..................................................................................................................................... displacements per atom DSC.....................................................................................................................differential scanning calorimeter EBR-II.................................................................................................................Experimental Breeder Reactor II EBSD......................................................................................................................electron backscatter diffraction ECAP...................................................................................................................... equal channel angular pressing NSUF LIST OF ACRONYMS Nuclear Science User Facilities 144 EDS...............................................................................................................energy dispersive X-ray spectroscopy EELS....................................................................................................................electron energy loss spectroscopy EML......................................................................................................................Electron Microscopy Laboratory ENDF........................................................................................................................... Evaluated Nuclear Data File EPRI.................................................................................................................... Electric Power Research Institute EUV...........................................................................................................................................extreme ultraviolet EUVR..................................................................................................................extreme ultraviolet reflectometry EXAFS....................................................................................................extended X-ray absorption fine structure FCC...........................................................................................................................................face-centered cubic FEG............................................................................................................................................ field-emission gun FFT....................................................................................................................................... fast fourier transform FHR............................................................................................ Floride Salt-Cooled High-Temperature Reactor FIB............................................................................................................................................... focused ion beam F-M........................................................................................................................................... ferritic-martensitic FP................................................................................................................................................... fission products HFEF........................................................................................................................Hot Fuel Examination Facility HFIR.............................................................................................................................. High Flux Isotope Reactor HIP.........................................................................................................................................hot isostatic pressing HOPG............................................................................................................... highly oriented pyrolytic graphite HRTEM................................................................................... high resolution transmission electron microscopy IASCC..............................................................................................irradiation-assisted stress corrosion cracking ICPMS..........................................................................................inductively coupled plasma mass spectrometry IFEL....................................................................................................... Irradiated Fuels Examination Laboratory IIT......................................................................................................................... Illinois Institute ofTechnology IMET................................................................................. Irradiated Materials Examination andTesting Facility IMPACT.............................................................Interaction of Materials with Particles and ComponentsTesting INL............................................................................................................................... Idaho National Laboratory ISU.......................................................................................................................................Idaho State University IVEM....................................................................................................intermediate voltage electron microscope LAMDA................................................................................ Low Activation Materials Development and Analysis LEAP.............................................................................................................................local electrode atom probe LEISS ..............................................................................................................low energy scattering spectroscopy 2014 ANNUAL REPORT 145 LWR............................................................................................................................................light water reactor MA........................................................................................................................................... mechanical alloying MaCS.........................................................................................................Microscopy and Characterization Suite MANTRA........Measurement of Actinide NeutronicTransmutation Rates with Accelerator Mass Spectrometry MCNP...................................................................................................................................monte carlo n particle MCOE................................................................................................Materials Center of Excellence Laboratories MFC...........................................................................................................................Materials and Fuels Complex MIT............................................................................................................ Massachusetts Institute ofTechnology MITR.............................................................................................Massachusetts Institute ofTechnology Reactor MRCAT...........................................................................................Materials Research Collaborative AccessTeam MSA......................................................................................................................................mercaptosuccinic acid MSTL.............................................................................................. Materials Science andTechnology Laboratory NAA............................................................................................................................ Neutron Activation Analysis NASA.......................................................................................... National Aeronautics and Space Administration NCNR.............................................................................................................. NIST Center for Neutron Research NCSU................................................................................................................... North Carolina State University NDE................................................................................................................................nondestructive evaluation NGA................................................................................................................................... Nuclear Energy Agency NE.................................................................................................................................................... Nuclear Energy NEET..........................................................................................................Nuclear Energy EnablingTechnologies NEUP............................................................................................................ Nuclear Energy University Programs NFS........................................................................................................................... nanostructured ferritic steels NGNP.................................................................................................................... Next Generation Nuclear Plant NIST............................................................................................National Institute of Standards andTechnology NMR..........................................................................................................................Nuclear Magnetic Resonance NRC.....................................................................................................................Nuclear Regulatory Commission NSUF ....................................................................................................................... Nuclear Science User Facility ODS.........................................................................................................................oxide dispersion strengthened OECD......................................................................Organisaton for Economic Co-operation and Developments ORNL....................................................................................................................Oak Ridge National Laboratory OSU...................................................................................................................................Oregon State University PED.........................................................................................................................precession electron diffraction Nuclear Science User Facilities 146 PI...........................................................................................................................................Principal Investigator PIE............................................................................................................................post-irradiation examination PNL.......................................................................................................................... Pacific Northwest Laboratory PNNL.........................................................................................................Pacific Northwest National Laboratory PSU...........................................................................................................................Pennsylvania State University PWR................................................................................................................................pressurized water reactor REDC........................................................................................ Radiochemical Engineering Development Center RERTR....................................................................................... Reduced Enrichment Research andTest Reactors RF................................................................................................................................................... radio frequency RIS.......................................................................................................................... radiation-induced segregation RPL............................................................................................................Radiochemistry Processing Laboratory RTE.......................................................................................................................... rapid turnaround experiment SCC...................................................................................................................................stress corrosion cracking SEM......................................................................................................................... scanning electron microscopy SPS....................................................................................................................................... spark plasma sintering SRB.....................................................................................................................................Scientific Review Board SRIM...........................................................................................................Stopping and Range of Ions in Matter SS.......................................................................................................................................................stainless steels SSRL..................................................................................................Stanford Synchrotron Radiation Laboratory STEM.................................................................................................scanning transmission electron microscopy TAMU...................................................................................................................................Texas AM University TEM...................................................................................................................transmission electron microscopy TMS.......................................................................................................The Minerals Metals Materials Society TRIGA............................................................................................... Training Research Isotope General Atomics TRISO.................................................................................................................................. tristructural isotropic TRU............................................................................................................................................. transuranic waste UCB................................................................................................................... University of California Berkeley UCF...........................................................................................................................University of Central Florida UFG.............................................................................................................................................ultra-fine grained UM..................................................................................................................................... University of Michigan UNLV......................................................................................................................University of Nevada LasVegas USU....................................................................................................................................... Utah State University 2014 ANNUAL REPORT 147 UW....................................................................................................................................University ofWisconsin VHTR..................................................................................................................... very-high temperature reactor XAFS...............................................................................................X-ray absorption fine-structure spectroscopy XANES...................................................................................................X-ray absorption near-edge spectroscopy XAS......................................................................................................................... X-ray absorption spectroscopy XPS....................................................................................................................X-ray photoelectron spectroscopy XRD.............................................................................................................................................. X-ray diffraction m.........................................................................................................................................................micrometre Nuclear Science User Facilities 148 NSUF INDEX ATR NSUF Partners Facilities Capabilities Argonne National Laboratory Advanced Photon Source389899101 Argonne National Laboratory IntermediateVoltage Electron Microscopy IVEM-Tandem facility97132135 Center for Advanced Energy Studies Microscopy and Characterization Suite3435 50606771798185879193105 107111119123127139141 Drexel University Central Research Facilities127 Idaho National Laboratory AdvancedTest Reactor325054576065 85879197105123135139141 Idaho National Laboratory Hot Fuel Examination Facility Analytical Laboratoryor Electron Microscopy Laboratory Irradiation Assisted Stress Corrosion Cracking Facility3460119139141 Idaho National Laboratory PIE facilities717791105135 Illinois Institute ofTechnology MRCAT at Argonne National Laboratorys Advanced Photon Source 38 40 Massachusetts Institute ofTechnology Reactor 18 26 33 60 72 77 Massachusetts Institute ofTechnology PIE facilities 77 North Carolina State University Nuclear Services Laboratories3435 North Carolina State University PULSTAR Reactor2632333839128 Oak Ridge National Laboratory High Flux Isotope Reactor33 Oak Ridge National Laboratory Hot CellsRadiological Laboratories LAMDA Facility35 Pacific Northwest National Laboratory Radiochemistry Processing Laboratory or Materials Science Technology Laboratory 3435 Purdue University IMPACT Facility3436 Sandia National Laboratory Ion Beam Laboratory127 University of California Berkeley Nuclear Materials Laboratory3436 University of Michigan Michigan Ion Beam Laboratory 38 39 University of Michigan Irradiated Materials Complex3436 University of Nevada LasVegas Harry Reid Center Radiochemistry Laboratories 3437 University of Wisconsin Tandem Accelerator Ion Beam3839120123125 127128131 University of Wisconsin Characterization Laboratory for Irradiated Materials3437 University of Wisconsin PIE facilities8587123 Westinghouse Materials Center of Excellence Laboratories737 Collaborators AllenTodd Idaho National Laboratory87123 AlsabbaghAhmad North Carolina State University105 Alsagabi Sultan University of Idaho111 Ban Heng Utah State University50 Barnard Leland University ofWisconsin Madison127 Barr Christopher University ofWisconsin Madison71127 Berg Jeff Idaho National Laboratory54 Betanco Felipe University of Central Florida65 Briggs Samuel University ofWisconsin Madison127 Burns Jatuporn Boise State University Center for Advanced Energy Studies91107111 Butt Darryll Boise State University107135 Carter Robert Electric Power Research Institute139 Charit Indrajit University of Idaho107111 ChenYiren Argonne National Laboratory81 Chien H.T. Argonne National Laboratory77 Chou Peter Electric Power Research Institute139 2014 ANNUAL REPORT 149 Cole James Idaho National Laboratory71107111 Conradson Steve Los Alamos National Laboratory87123 Daw Joshua Idaho National Laboratory77 Dolph Corey Boise State University91 DongYan University of Michigan67 Eapen Jacob University ofWisconsin Madison131 Eriksson Nicholas University of Central Florida65 Gan Jian Idaho National Laboratory79879397123 Gan Jian Idaho National Laboratory135 Garcia-Diaz Brenda L. Savannah River National Laboratory60 Garner Frank Consultant67 Geller Esin University of Central Florida65 GerezakTyler University ofWisconsin Madison85 Giglio Jeff Idaho National Laboratory54 Gupta Mahima University ofWisconsin Madison8797123 Harris Kurt Utah State University50 HartmannThomas University of NevadaLasVegas50 He Lingfeng Idaho National Laboratory University ofWisconsin Madison6097135 Henderson Hunter University of Florida7993 Hill Connie Idaho State University119 Hoffman Elizabeth N. Savannah River National Laboratory60 Hua Zilong Utah State University50 Imel George Idaho State University54 Jackson John H. Idaho National Laboratory139 Jurisson Silvia University of Missouri101 Keiser Jr. Dennis D. Idaho National Laboratory65 Kirk Marquis Argonne National Laboratory 97 Kohse Gordon Massachusetts Institute ofTechnology77 Krishna Ram University ofWisconsin Madison131 Leng Bin University ofWisconsin Madison85 Nuclear Science User Facilities 150 Li Zhangbo University of Florida81 LilloTomas Idaho National Laboratory 85 119 LoWei-yang University of Florida81 Madden James Idaho National Laboratory85119 MaddockTom Idaho National Laboratory54 Manuel Michele University of Florida7993 Marquis Emmanuelle University of Michigan67 Miller Brandon Idaho National Laboratory105 Morgan Dane University ofWisconsin Madison127 Murray Paul Idaho National Laboratory141 Murty K. L. North Carolina State University105 Murty K. Linga - University ofWisconsin Madison131 Newell Ryan University of Central Florida65 Okuniewski Maria Idaho National Laboratory65 Pakarinen Janne University ofWisconsin Madison 8187123127135 Palmer Joseph Idaho National Laboratory77 Palmiotti Giuseppe Pino Idaho National Laboratory54 ParkYoungjoo University of Central Florida65 Pasebani Somayeh University of Idaho107 Porter Doug Idaho National Laboratory105 Post Guillen Donna Idaho National Laboratory50 Price Lloyd Texas AM University107111 Ramuhalli Pradeep Pacific Northwest National Laboratory77 Reinhardt Brian Pennsylvania State University77 Reinig Kimberly University of Missouri101 Rempe Joy - Idaho National Laboratory77 Seibert Rachel - Illinois Institute of Technology101 Sencer Bulent Idaho National Laboratory67 Shao Lin Texas AM University111 SohnYongho University of Central Florida65 Sridharan Kumar University ofWisconsin Madison85 Sridharan Kumar - University ofWisconsin Madison127 Sterbentz Jim Idaho National Laboratory54 SuprockAndy Pennsylvania State University77 Swenson Matthew Boise State University 91 Szlufarska Izabela University ofWisconsin Madison85 Taheri Mitra Drexel University University ofWisconsin Madison71127 Tallman Darin J. Drexel University60 Taylor Joanna University of Idaho91 Terry Jeff Illinois Institute ofTechnology101 Teysseyre Sebastien Idaho National Laboratory139 Tittmann Bernhard Pennsylvania State University77 Valderrama Billy University of Florida7993 van Rooyen Isabella Idaho National Laboratory85119 Wampler Heather Utah State University50 Wen Haiming Idaho National Laboratory119 Wernsman Ben Bettis Atomic Power Laboratory77 Wharry Janelle Boise State University91 WuYaqiao Center for Advanced Energy Studies85 WuYaqiao Boise State University91 WuYaqiao Boise State UniversityCenter for Advanced Energy Studies8591107111119 Yagnik Suresh Electric Power Research Institute141 YangYong University of Florida81 Youinou Gilles Idaho National Laboratory54 Youngsman John Boise State University119 ZabriskieAdam Utah State University50 2014 ANNUAL REPORT 151 15-50145 Nuclear Science User Facilities 995 University Boulevard Idaho Falls ID 83401-3553 www.nsuf.inl.gov