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DOE OSTI2012

Identifying and bounding uncertainties in nuclear reactor thermal power calculations

Determination of the thermal power generated in the reactor core of a nuclear power plant is a critical element in the safe and economic operation of the plant. Direct measurement of the reactor core thermal power is made using neutron flux instrumentation; however, this instrumentation requires frequent calibration due to changes in the measured flux caused by fuel burn-up, flux pattern changes, and instrumentation drift. To calibrate the nuclear instruments, steam plant calorimetry, a process of performing a heat balance around the nuclear steam supply system, is used. There are four basic elements involved in the calculation of thermal power based on steam plant calorimetry: The mass flow of the feedwater from the power conversion system, the specific enthalpy of that feedwater, the specific enthalpy of the steam delivered to the power conversion system, and other cycle gains and losses. Of these elements, the accuracy of the feedwater mass flow and the feedwater enthalpy, as determined from its temperature and pressure, are typically the largest contributors to the calorimetric calculation uncertainty. Historically, plants have been required to include a margin of 2% in the calculation of the reactor thermal power for the licensed maximum plant output to account for instrumentation uncertainty. The margin is intended to ensure a cushion between operating power and the power for which safety analyses are performed. Use of approved chordal ultrasonic transit-time technology to make the feedwater flow and temperature measurements (in place of traditional differential-pressure- based instruments and resistance temperature detectors [RTDs]) allows for nuclear plant thermal power calculations accurate to 0.3%-0.4% of plant rated power. This improvement in measurement accuracy has allowed many plant operators in the U.S. and around the world to increase plant power output through Measurement Uncertainty Recapture (MUR) up-rates of up to 1.7% of rated power, while also decreasing the probability of significant over-power events. This paper will examine the basic elements involved in calculation of thermal power using ultrasonic transit-time technology and will discuss the criteria for bounding uncertainties associated with each element in order to achieve reactor thermal power calculations to within 0.3% to 0.4%. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Research on acceleration method of reactor physics based on FPGA platforms

The physical designs of the new concept reactors which have complex structure, various materials and neutronic energy spectrum, have greatly improved the requirements to the calculation methods and the corresponding computing hardware. Along with the widely used parallel algorithm, heterogeneous platforms architecture has been introduced into numerical computations in reactor physics. Because of the natural parallel characteristics, the CPU-FPGA architecture is often used to accelerate numerical computation. This paper studies the application and features of this kind of heterogeneous platforms used in numerical calculation of reactor physics through practical examples. After the designed neutron diffusion module based on CPU-FPGA architecture achieves a 11.2 speed up factor, it is proved to be feasible to apply this kind of heterogeneous platform into reactor physics. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Discrete ordinate quadrature selection for reactor-based Eigenvalue problems

In this paper we analyze the effect of various quadrature sets on the eigenvalues of several reactor-based problems, including a two-dimensional (2D) fuel pin, a 2D lattice of fuel pins, and a three-dimensional (3D) reactor core problem. While many quadrature sets have been applied to neutral particle discrete ordinate transport calculations, the Level Symmetric (LS) and the Gauss-Chebyshev product (GC) sets are the most widely used in production-level reactor simulations. Other quadrature sets, such as Quadruple Range (QR) sets, have been shown to be more accurate in shielding applications. In this paper, we compare the LS, GC, QR, and the recently developed linear-discontinuous finite element (LDFE) sets, as well as give a brief overview of other proposed quadrature sets. We show that, for a given number of angles, the QR sets are more accurate than the LS and GC in all types of reactor problems analyzed (2D and 3D). We also show that the LDFE sets are more accurate than the LS and GC sets for these problems. We conclude that, for problems where tens to hundreds of quadrature points (directions) per octant are appropriate, QR sets should regularly be used because they have similar integration properties as the LS and GC sets, have no noticeable impact on the speed of convergence of the solution when compared with other quadrature sets, and yield more accurate results. We note that, for very high-order scattering problems, the QR sets exactly integrate fewer angular flux moments over the unit sphere than the GC sets. The effects of those inexact integrations have yet to be analyzed. We also note that the LDFE sets only exactly integrate the zeroth and first angular flux moments. Pin power comparisons and analyses are not included in this paper and are left for future work. (authors)

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2013

The MSFR as a flexible CR reactor: the viewpoint of safety

In this paper, the possibility has first been discussed of using the liquid-fuelled Molten Salt Fast Reactor (MSFR) as a flexible conversion ratio (CR) reactor without design modification. By tuning the reprocessing rate it is possible to determine the content of fission products in the core, which in turn can significantly affect the neutron economy without incurring in solubility problems. The MSFR can thus be operated as U-233 breeder (CR>1), iso-breeder (CR=1) and burner reactor (CR<1). In particular a 40 year doubling time can be achieved, as well as a considerable Transuranics and MA (minor actinide) burning rate equal to about 150 kg{sub HN}/GWE-yr. The safety parameters of the MSFR have then been evaluated for different fuel cycle strategies. Th use and a softer spectrum combine to give a strong Doppler coefficient, one order of magnitude higher compared to traditional fast reactors (FRs). The fuel expansion coefficient is comparable to the Doppler coefficient and is only mildly affected by core compositions, thus assisting the fuel cycle flexibility of the MSFR. βeff and generation time are comparable to the case of traditional FRs, if a static fuel is assumed. A notable reduction of βeff is caused by salt circulation, but a low value of this parameter is a limited concern in the MSFR thanks to the lack of a burnup reactivity swing and of positive feedbacks. A simple approach has also been developed to evaluate the MSFR capabilities to withstand all typical double-fault accidents, for different fuel cycle options.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Updated Radiological Inventory of G1 Reactor Thanks to a Strengthened Data Processing - 17271

The radiological inventory of G1 reactor is under refinement and consolidation. This reactor, located at the CEA Marcoule center in France, presents an issue for Chlorine-36 as graphite was used as moderator. The improvement of the 3D physical description and a better understanding of historical events enable the computation of activation levels thanks to a numerical model. These results are compared to old samples and more recent ones to quantify nuclide activity levels as well as their spatial distributions. In particular a spatial and statistical correction of sample bias leads to an update of the radiological inventory with an expected reduction of average activity levels. Another part of the work consists in identifying missing pieces of information and still uncertain parts of the reactor (sample analysis of activated concrete for instance) to reinforce the radiological inventory of G1 reactor. This will be the basis of future investigation works. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Advancing Core Safety of Gen IV and Gen IV + Reactors by Encompassing High-Temperature Resistant Materials - 18673

The paper focuses on the development of materials for Gen IV/IV+ reactor core. It discusses the advancement of materials studied and their application in different components of the reactor core and other auxiliary components which will help in the development of future generation reactors. The paper summarizes the material study showing the characteristic behavior of different materials (coated with FeB and FeB{sub 2}), towards irradiation at near high temperatures and high energy neutron flux condition. The project includes different materials which are coated on the boundaries with varied thickness of FeB and FeB{sub 2}. The samples included Carbon steel, Stainless steel 310, Stainless steel 316, Inconel 718. All above-mentioned samples were received by Prof. J. Szpunar, through an industry with coatings of FeB and FeB{sub 2} with varied thickness along with a bare sample of each material. The samples were irradiated at the University of Saskatchewan reactor, with fast neutron flux of 1.1x10{sup 11} n.cm{sup -2}.sec{sup -1}. I worked on all the samples but the paper particularly focuses on, two carbon steel samples results, received so far. The methodology for the project been followed for all the samples is as follow: 1. Received samples were first polished. 2. Vickers Hardness Test. 3. Mirror polished. 4. Etched with the appropriate etchant for the material. 5. Microscopic study of the etched surface. 6. SEM imaging. 7. Sent for irradiation. 8. The same set of material testing done after irradiation, with an EBSD run on the boundary and main surface. The two Carbon Steel samples (sample 21 and 19) analyzed, were irradiated for 14 hrs. Sample 21 was a bare sample and sample 19 had a coating of FeB, FeB{sub 2} with a ∼0.0010'' thickness. The methodology was followed for material testing before and after the irradiation. The advantage of FeB and FeB{sub 2} coatings are: 1. Higher hardness - 10 times (FeB-1775 and FeB{sub 2}-1450), 2. High-stress capacity, 3. Lower abrasion, 4. Low corrosion. Through the test results obtained before and after the irradiation, I was able to get a view of coating damage at the boundaries of sample 19. The bare sample don't show any difference after irradiation but the coated sample do show cracks in the FeB layer. Which can be seen in the SEM images of sample 19. The hardness results are promising, as hardness didn't change significantly. The hardness test was done on the substrate as the coating was significantly narrow to perform the test. Though the hardness test on coatings was performed by the possessor company which show hardness higher by a factor of 10 when compared to uncoated surfaces. The XRD results, reveal irradiation effect on the crystal lattice, making it deformed, which though is usual during irradiation. The overall analysis along with EBSD done on the two samples are favorable and promising. Though it still needs to be cleared out, whether the cracks on the coating were due to irradiation or due to etching after the samples were irradiated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

TREAT Transient Test Reactor Restart Status

Transient testing involves placing nuclear fuel or material into the core of a nuclear reactor designed to operate at high power for a short time, and subjecting the nuclear fuel or material to short bursts of intense, high power radiation. After the experiment is completed, the fuel or material is analyzed to determine the effects of the radiation. The results are then used in fuel or material design and/or qualification. Transient testing is required for essentially all nuclear fuel design and qualification efforts to learn how nuclear fuel will respond during accidents involving transient overpower and/or under cooling events. For example, nuclear fuel may fragment when exposed to higher than normal power. This fragmentation can cause unacceptable performance. Transient testing is also needed to validate performance models for nuclear fuel and materials. These models, when validated, will dramatically shorten the development and qualification life cycle for nuclear fuels, supporting rapid development of low emissions, reliable power generation. In the past, transient testing was primarily done in facilities that no longer exist. The U.S. Department of Energy evaluated how to provide the required transient testing capability and, following completion of the National Environmental Policy Act process, they selected the Transient Reactor Test (TREAT) facility to resume transient testing. TREAT is located at Idaho National Laboratory (INL). The TREAT reactor was used from 1959 to 1994 to conduct more than 2,500 nuclear fuel transient tests, and was placed in standby in 1994. TREAT is capable of transients in excess of 19 GW. The paper describes the progress towards restart. The restart efforts have revealed that the plant fuel and equipment is acceptable for future operations. The safety basis development, procedures, and training are well under way. Experiment capability preparations are in progress. Reactor startup is anticipated no later than 2018. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Additive Manufacturing of Research Reactor Control Elements and Subsequent Neutron Irradiation

Additive manufacturing techniques have made major strides in recent years and are now capable of creating complex components in near-final form. The nuclear industry has many applications with complicated and even unique components that are time consuming, expensive, and more difficult to make using traditional fabrication routes. Modern fabrication techniques possess the potential to reduce fabrication lead-time and costs while increasing quality and efficiency. However, these methods have not been adopted due to the need for strict validation for nuclear applications. Unlike structural components that largely consist of a uniform material, for nuclear reactor core components incorporation of fissile or neutron poison materials inside a cladding becomes necessary. This in turn requires a more complex and flexible manufacturing method to produce these hybrid components. In this study, demonstration of ultrasonic additive manufacturing (UAM) process to produce components resembling research reactor control elements was carried out]. Specifically, the control elements for ORNL's High Flux Isotope Reactor (HFIR) were targeted. These control elements (consisting of an inner cylinder and four outer plates) are placed on the periphery of the circular HFIR core. The control elements consist of Al6061-T6 alloy with a 13-cm-long grey and a 56-cm-long black region, corresponding to 40 wt% of Ta and 30 wt% Eu{sub 2}O{sub 3} (europia) are incorporated along their length. In this study various fabrication parameters for production of aluminum blocks with and without embedded Eu{sub 2}O{sub 3} and Ta bearing compacts were explored. Select optimized specimens we then irradiated inside HFIR for one reactor cycle. The UAM process exploits ultrasonic energy to facilitate solid-state welding of metal foils in contact under pressure. This additive process is carried out in successive layers to build up the material and when coupled with CNC milling, it yields complex 3D components. The milling and subsequent additive steps may be used to embed foreign material into the material undergoing buildup. In this manner fissile or neutron poison elements may be incorporated into the complex structure, though in an inhomogeneous fashion. Multiple Al-6061 blocks were produced by UAM using 25 mm wide by 0.15 mm thick Al-6061-H18 tapes. The mechanical properties of these blocks in the as-fabricated condition was characterized in detail and it was shown that a great degree of anisotropy in the mechanical behavior exists between the direction in the tape plane and the direction perpendicular to the tapes (buildup direction) The latter exhibited reduced modulus and significantly smaller strength. Post production heat treatment appeared to greatly enhance the strength in the buildup direction. Multiple tensile specimens machined per SS-J3 geometry were prepared for irradiation testing. The tensile specimen matrix was set up to enable post irradiation examination of mechanical properties in various directions as well as after heat treatment. A small block with embedded Ta- and Eu{sub 2}O{sub 3}-bearing compacts, three of each kind, was also produced, as demonstrated previously. These compacts were 5.7 mm in diameter and 1.6 mm in height and contained 60 vol Ta or Eu{sub 2}O{sub 3}. The poison-bearing Al-6061 block alongside the tensile bars was loaded into a perforated rabbit (irradiation capsule) for irradiation in HFTK. The perforated rabbit allows penetration of HFIR water coolant into the Interior of the capsule in direct contact with the test specimens. The rabbit was loaded into the flux trap region at the uppermost position in the core. Irradiation temperature was 68 deg. C and fast neutron (E>0.1 MeV) fluence was 10{sup 25} neutrons/m{sup 2}. Currently PIE activities are ongoing that are focused on mechanical testing of irradiated tensile bars and metallographic examination of neutron poison bearing block.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Corrosion of Candidate Materials in Molten FLiBe Salt for Fluoride-Salt-Cooled High-Temperature Reactor (FHR)

Molten fluoride salts have attracted considerable attention because of their desirable thermophysical properties that lead to superior heat transfer. Recently, fluoride salts are being considered as coolants in the fluoride salt-cooled high temperature reactors (FHRs). The successes of the molten salt reactor experiment (MSRE) program in 1950's-70's at Oak Ridge National Laboratory (ORNL) has also driven the pursuit of molten salt nuclear reactor as one of the lead next generation nuclear reactor concepts. Unlike MSRE where the uranium fuel was dissolved in the molten fluoride salt, the FHR concept is expected to use molten salt (2LiF-BeF{sub 2}, FLiBe) only as coolant in conjunction with solid TRISO fuel particles immersed in the molten salt. One of the important challenges in FHR development is corrosion of structural materials in high-temperature molten 2LiF-BeF{sub 2} (FLiBe) salt. Currently, 316 stainless steel (nominal composition: Fe-12Ni-17Cr-2.5Mo-1Si-2Mn-0.03C) and Hastelloy N (nominal composition: Ni-7Cr-16Mo-5Fe-1Si-0.8Mn) are deemed as promising candidate alloys for the construction of some of the structural components of FHR. 316 stainless steel is ASME code certified for high temperatures (up to 760 deg. C) applications. Hastelloy N has been shown to have outstanding corrosion resistance in molten FLiBe but present code certification data limit the use of this alloy to only about 704 deg. C. Modifications of Hastelloy N with higher creep strength are presently being developed at ORNL. Other materials such graphite, SiC/SiCf and C/Cf composites, are also considered for structural applications in the FHR. The presence of these multiple materials in a molten FLiBe salt medium can further exacerbate and complicate corrosion effects. Therefore, it is necessary to understand the corrosion behavior of these materials in molten FLiBe salt. The corrosion behavior of Hastelloy N, 316 stainless steel, and CVD SiC and two types of SiC-SiC{sub f} composites have been tested in molten FLiBe salt at 700 deg. C. When tested in metal-lined crucibles both Hastelloy N and 316 stainless steel showed very good corrosion resistance. Samples tested in graphite crucibles showed greater propensity for corrosion and carburization was observed in the near surface regions of the alloys. CVD SiC showed outstanding corrosion resistance. However, the two SiC-SiC{sub f} composites exhibited corrosion attack implying that the manufacturing process may play a role in the corrosion of these composites. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Long-Term Stewardship Optimization for a Decommissioned Nuclear Reactor - 19579

The decommissioned Hallam Nuclear Power Facility (Hallam), located in Lancaster County, Nebraska, approximately 19 miles south of Lincoln, was developed as a prototype to investigate the technical and economic feasibility of a sodium-cooled, graphite-moderated nuclear reactor. As part of its Power Demonstration Program, the US Atomic Energy Commission (US AEC), a predecessor agency of the US Department of Energy (US DOE), operated the Hallam reactor from 1962 to 1964. The radiological surveillance of the Hallam site was established when the operating agreement between AEC and the Nebraska Public Power District was terminated. The US DOE Chicago Operations Office and the Nebraska Department of Health (NDH) agreed to install a shallow groundwater monitoring system as part of the environmental surveillance program. Surveillance at the site was scheduled to end in fiscal year 2005, with no further activities planned after that. At that time, the plan was to transition the site to the Nebraska Public Power District with future use of the facility to remain restricted. US DOE has title to and responsibility for the entombed radioactive materials. A surveillance and monitoring program was initiated in 1970 by NDH and was funded by US AEC. It included the analysis of samples from deep production wells (groundwater from the regional aquifer at depths greater than 55 meters [180 feet]) at the Sheldon Station. NDH was concerned in 1990 about the possibility of shallow groundwater coming in contact with the buried radiological materials along the buried walls of the reactor. Subsequently, DOE and NDH agreed to further characterize hydrologic conditions and establish a monitoring program in the shallow, perched groundwater zones. The current monitoring program focuses on the shallow, perched groundwater. Today, decades since facility decommissioning, site inspections and groundwater monitoring continue at this Cold War-era legacy site. The US DOE Office of Legacy Management (LM) monitors groundwater as a best management practice in response to a request from the Nebraska Department of Health and Human Services (DHHS). In 2006, LM recommended discontinuing groundwater monitoring because analytical results since 1970 demonstrated there had been no impact to shallow, perched groundwater and no current or anticipated unacceptable risk to human health and the environment. The State of Nebraska did not concur with the recommendation to stop monitoring but did agree to a reduction in sampling and analysis from once a year to once every 2 years--a schedule that began in 2008. An updated assessment of the groundwater monitoring effort was issued in 2016. The entire body of water-quality data continues to demonstrate that there have been no negative impacts to the shallow, perched groundwater from radioactive materials entombed at the site 46 years ago. LM, with concurrence from Nebraska DHHS, changed the sampling frequency, effective in 2018, as follows: once every 5 years from 2021 to 2041 and once every 10 years from 2041 to 2071. In summary, LM remains committed to its primary mission to protect human health and the environment at each of its 92 legacy sites. With results from groundwater sampling events demonstrating compliance for 48 years, the Hallam site was able to set an LM benchmark to optimize long-term stewardship and define an end-state strategy for other decommissioned reactor facilities across the DOE complex. Continued collaboration with local communities, stakeholder, and regulators often leads to outcomes of sound protectiveness without compromise to safety or compliance, as demonstrated by the optimization of the long-term stewardship at the Hallam site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

In-pile measurements of fuel rod dimensional changes utilizing the test reactor loop pressure for motion - 359

Different types of materials are being considered for fuel, cladding, and structures in existing and advanced nuclear reactors. These materials can undergo significant dimensional changes during irradiation. Currently in the US, such changes are measured with the 'cook and look' method (i.e. repeatedly irradiating a specimen for a specified period of time, and then removing it from the reactor for evaluation) - which leads to costly and time consuming experiments. In addition, such techniques provide limited data, and handling may disturb the phenomena of interest. Therefore, in-pile detection of geometric changes is needed to understand real-time behavior during irradiation testing of fuels and materials in high flux US Material and Test Reactors (MTRs). The developmental results of an advanced Linear Variable Differential Transformer (LVDT)-based test rig capable of detecting real-time changes in diameter or length of fuel rods or other material samples during irradiation in US MTRs are presented. The LVDT is being developed at the Idaho National Laboratory (INL), and will provide experimenters with the capability to measure dimensional changes associated with fuel and clad swelling, pellet-clad interaction, and crud buildup. The LVDT design utilizes the MTR's loop pressure to drive the test specimen at a constant rate. This leads to a simplified construction which only requires three pressure boundary penetrations. Further, the LVDT pressure boundary is a leak tight hydraulic ram capable of actuating under low differential pressure (e.g. 1 psid), and is able to survive the high irradiation tests at Pressurized Water Reactor (PWR) conditions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

In-Line Method for Calculating Peak Neutron Absorption by Fission Products Following a Change in Reactor Power

When the power level in an operating reactor is reduced the concentration of xenon-135 and samarium-149 temporarily increases as the nuclides (and their immediate precursors) equilibrate to the new power level. During this period, the increased inventory of these fission products causes an insertion of negative reactivity that can last for tens of hours. Understanding and predicting this transient reactivity effect is important in reactor design and operation. Explicit depletion calculations are often used to predict the reactivity behavior following a power change. However, these calculations require many time-steps to fully resolve the reactivity trend, and are correspondingly expensive. In 2010 an in-line method for estimating the peak xenon mass in a core following shutdown was described. However, there is no guarantee that the peak negative reactivity condition for a core will occur at peak xenon mass. Therefore, in this summary, we extend this original work using a new method that can predict the peak neutron absorption rate in the core due to {sup 135}Xe and {sup 149}Sm following any reduction in power. The proposed method can also be used to calculate the peak positive reactivity transient due to an increase in reactor power, but this scenario is of less interest in most reactor design applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Helium Brayton Cycle Design for Advanced High Temperature Reactor System

Gen-IV reactors have many advantages over current generation reactors. Higher outlet temperatures in these next generation reactors allow for other power cycles to be considered, such as the helium Brayton cycle. Additionally, the higher temperatures allow the energy to be utilized in other industrial applications. The advanced high temperature reactor concept utilizes the helium Brayton cycle for electricity production. The system consists of four components shown in Fig. 1. This includes a turbine, recuperator, precooler, and compressor. The helium coolant exits the secondary heat exchanger (SHX) at a high temperature and enters the turbine. Helium spins the turbine causing the generator to spin producing electricity. The pressure and temperature drop as the coolant travels through the turbine. The helium enters a recuperator where the cold side stream exiting the compressor is heated up. Helium then travels through a precooler where water is used to cool down the helium before entering the compressor. Helium enters the compressor where the temperature and pressure of the coolant increase. Compressed helium travels through the recuperator heating up to the desired temperature before entering the SHX again. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

An Investigation on the Sustainability of a Sodium-cooled Breed-and-Burn Fast Reactor (B and BR)

Fast reactor can effectively handle the used nuclear fuel (UNF) by transmuting long-living transuranic (TRUs) into relatively short-living fission products. It can also be used to produce fissile from fertile nuclides as fuels for another reactor. The breed-and-burn fast reactor (B and BR) is a unique concept of fast reactor, which can breed the fissile fuels and use the bred fuels in situ. Thanks to this characteristic, the fuel utilization in a B and BR can be extremely high and even the recycled fuel of a B and BR can be used as a fuel to spawn the next generation B and BR. The UNF from the first B and BR should be reprocessed or reconditioned so that it can be reused in the second B and BR. A fuel reprocessing technology named melt refining was used in EBR-II from 1964 to 1969 and it was also applied in our previous study which showed that second B and BR can be started using the recycled UNF of first B and BR. However, about 95% of rare earths are removed during the melt refining and the associated processes are not very simple. Moreover, the actinide loss can also be significant. To enhance the proliferation-resistance and economy of the reprocessing, a super-simplified melt and treatment (SSMT) process without removing rare earth nuclides is suggested and the feasibility study to start the second B and BR using SSMT is carried out in this paper. The neutronics simulations were done using the Monte Carlo code McCARD in conjunction with ENDF/B-VII cross-section library. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

LOCA and LOFA Safety Margin Analysis for U-10Mo Monolithic LEU Fuel Conversion of the MURR{sup R} Reactor

The University of Missouri Research Reactor (MURR{sup R}) is a multi-disciplinary research and education facility providing a broad range of analytical and irradiation services to the research community and the commercial sector. The source of neutrons for the experimental facilities of MURR{sup R} is a compact 10 MW core fueled with highly-enriched uranium (HEU) fuel. The continued presence of HEU fuel in civilian installations such as research and test reactors poses a threat to national and international security. Therefore, most research and test reactors, both domestic and international, have started a program of conversion to the use of low-enriched uranium (LEU) fuel. MURR{sup R} is one of five U.S. high-performance research and test reactors (USHPRRs) that are actively collaborating with the U.S. National Nuclear Security Agency (NNSA) Office of Material Management and Minimization (M3) to develop a suitable LEU fuel replacement for the currently required HEU fuel. A new type of high-density LEU fuel based on an alloy of uranium and molybdenum is currently undergoing qualification testing, and is expected to allow the conversion of all USHPRRs. As a part of the LEU conversion of MURR{sup R}, an analysis of core behavior under a set of accident conditions for a core fueled with an LEU monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) has been conducted. This paper presents the newest results for the loss-of-coolant accidents (LOCAs) and loss-of-flow accidents (LOFAs). Other classes of accidents have also been investigated but were not included here. Namely, reactivity insertion accidents (RIAs), fuel failure during operations accidents (FFAs), and fuel handling accidents (FHAs)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

New Czech R and D Program on Fluoride Salt-cooled Nuclear Reactor Systems

A technology of nuclear reactor systems with liquid molten salt fuel has been investigated in the Czech Republic since 1999. The original effort came from the national Partitioning and Transmutation concept based on the subcritical Accelerator Driven System for incineration of transuranium elements with liquid fluoride fuel and pyrochemical partitioning fuel cycle technology. After 2005, the original R and D intentions were gradually converted to classical Molten Salt Reactor (MSR) technology and to thorium - uranium fuel cycle. The aim of this choice was to contribute to the development of an advanced nuclear technology, which can minimize environmental impact of nuclear power, save the natural resources and which has some potential to be deployed also in a non-superpower country. The basic technological development of selected areas of MSR and Th - U fuel cycle technology was realized in the frame of the SPHINX project solved in 2005 - 2012. In addition to the domestic activities of the MSR technology development, in 2012 the Czech Ministry of Industry and Trade and the US - Department of Energy concluded an agreement about the collaborative R and D on Molten Salt Reactor and Fluoride-salt-cooled High-temperature Reactor (FHR) technologies. These facts and results created the principal background for the present national program on the MSR/FHR technology development which was approved by the Ministry of Industry and Trade in 2016 and is granted by the Technological Agency of the Czech Republic.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Growth of the international criticality safety and reactor physics Benchmark experiment evaluation projects since ICNC 2011 - 14102

Since the International Conference on Nuclear Criticality Safety (ICNC) 2011, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts. Twenty countries have continued to participate in the ICSBEP since 2011; participating countries in the IRPhEP have grown from 16 in 2011 to 20 in 2015. Since ICNC 2011, the contents of the International Handbook of Evaluated Criticality Safety Benchmark Experiments have increased from 533 evaluations (approximately 58,000 pages), containing benchmark specifications for 4,552 critical or subcritical configurations to 561 evaluations (approximately 67,000 pages), containing benchmark specifications for 4,839 critical or subcritical configurations in the 2014 Edition of the ICSBEP Handbook. While the current number of criticality-alarm-placement/shielding benchmark configurations has remained at 24, the number of fundamental physics measurements relevant to criticality safety applications has increased from 200 to 207. Since ICNC 2011, the contents of the International Handbook of Evaluated Reactor Physics Benchmark Experiments have increased from 53 experimental series that were performed at 31 different reactor facilities to 143 experimental series that were performed at 50 different reactor facilities in the 2015 Edition of the IRPhEP Handbook. The status of the ICSBEP and the IRPhEP is discussed and benchmarks that have been more recently added to the handbooks are highlighted. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Flattening the Radial Temperature Profile across the Transformational Challenge Reactor Core

The Transformational Challenge Reactor (TCR) program is demonstrating an agile development approach to advanced nuclear reactor design, which has traditionally utilized a linear design process. In leveraging artificial intelligence, additive manufacturing, advanced materials, and cutting-edge modeling and simulation, the TCR program aims to minimize the high cost and lengthy deployment timelines now standard in the nuclear industry. Within a relatively short period of time, a robust and mature advanced gas-cooled reactor was iteratively designed under the TCR program, using these cutting-edge technologies. The TCR is a 3 MWt gas-cooled microreactor fueled with uranium nitride (UN) tristructural isotropic (TRISO) fuel particles. Though manufactured via traditional means, these UN TRISO particles are loaded into additively manufactured silicon carbide (SiC) cans [4]. Once loaded with TRISO particles, the SiC cans are densified using a chemical vapor infiltration process. The additively manufactured SiC enables significantly more freedom in the design of the fuel form than could ever be achieved using traditionally manufactured SiC. The helium coolant, pressurized to 5 MPa, enters the core at 300°C and nominally exits it at 500°C. Typically, the most thermally limiting components in any reactor design are the fuel assemblies in the core center. To provide a wide thermal margin in these central fuel assemblies, the flow may be biased toward the center of the core to more effectively cool these fuel assemblies with more power deposition and flatten the core’s radial temperature distribution. An analytical fluid model of the TCR core was developed to explore methods for biasing the flow away from the cooler outer fuel assemblies and towards the hotter inner ones. Higher-fidelity models developed in STAR-CCM+ 2020.3.1, a computational fluid dynamics code, were then utilized to verify the analytical model’s findings.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗