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DOE OSTI2016

Reactor Physics Analysis of Transitioning to a Thorium Fuel Cycle with Molten Salt Reactors

Commercial power reactors operating in the United States are light-water reactors (LWRs) fueled with low-enriched uranium (LEU) in a once-through fuel cycle; the important fissile isotopes are {sup 235}U, {sup 239}Pu, and {sup 241}Pu. In a thorium fuel cycle, the fertile isotope {sup 232}Th is converted into {sup 233}U, a fissile isotope that is not naturally occurring. In transitioning from the once-through fuel cycle to the thorium fuel cycle, the availability of man-made {sup 233}U presents a challenge: without it, the initial fissile material fueling the first reactors built in this transition is not the fissile isotope that the reactor is breeding. The transition requires an alternate fissile material until enough surplus {sup 233}U is available to start up additional reactors. In liquid-fueled molten salt reactors (MSRs), the fuel is dissolved into molten salt, which is continuously circulated through the core, simultaneously undergoing irradiation, chemical treatments and separations, and feeds. MSRs are an attractive technology in which the thorium fuel cycle can be deployed, as the {sup 233}U bred from {sup 232}Th is continuously being recycled back through the core. The bred fissile material is immediately available to fuel the core without processing or fuel fabrication. After start-up, only a fertile {sup 232}Th feed is necessary to sustain criticality in the MSR. But, in single-fluid MSRs, the initial fissile and fertile materials are mixed into the same molten fuel salt, complicating separations of initial and bred fissile materials. In solid-fuel systems, these materials are often separate. This analysis herein builds on previous work modeling thorium-fueled MSR start-up. It is based on unit cell representations of the graphite-moderated, 2250 MWt Oak Ridge National Laboratory (ORNL) Molten Salt Breeder Reactor (MSBR). This paper discusses the start-up of the first MSR in the transition to a thorium fuel cycle and identifies challenges and limitations of fuel cycle deployment with this reactor technology. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Criticality Benchmarking of the Oregon State TRIGA Reactor Using the MCNP Burn Option

Neutronic analyses of the Oregon State TRIGA Reactor (OSTR) have historically been performed using a thoroughly validated MCNP model. This model was originally created in 1997 in support of boron neutron capture therapy (BNCT) research, when the core still utilized highly-enriched uranium (HEU) fuel. The OSTR core was converted to low-enriched uranium (LEU) in the fall of 2008. To more accurately represent the converted state of the core, the MCNP model was modified to incorporate fuel material information from the fuel manufacturer (AREVA). In previous versions of the model, the fuel elements were assumed to have a uniform isotopic composition. Chemical analysis from AREVA provided far more resolved material information, allowing for increased fidelity in the representation of the fuel. The model was also updated to ENDF/B-VII.1 cross section libraries as the previous model utilized ENDF/B-VI. The primary impetus for the update of the MCNP model was the desire to have a more accurate assessment of criticality conditions of the OSTR. The OSTR utilizes three core configurations which are defined by the contents of the B1 grid position. If fuel resides in B1, this is designated as the 'normal' core. Otherwise, there are two aluminum in-core irradiation tubes (ICITs) that can be placed in B1. There is an unlined tube, designated simply as the ICIT, and there is a cadmium-lined tube, designated as the CLICIT. Table 1 contains a summary of the results of a series of MCNP criticality calculations which used the original 'fresh fuel' isotopics from the fall of 2008 in conjunction with critical rod height data from December 2013. Calculations were performed at 15 W, which is where the daily core excess measurements are taken, and at 1 MW, which is the current full-power level at the OSTR. The k-effective values indicate that the fresh fuel material cards are highly inaccurate and over-predict the criticality by at least $1.20 in every core configuration (based on a β{sub eff} value of 0.0075), thus it was desired to perform a depletion calculation to obtain more accurate fuel information. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Assessment of Near-Breeding Heterogeneous Seed/Blanket Cores in Pressure-Tube Heavy Water Reactors with Thorium-Based Fuels

To help extend uranium resources, thorium, which is three to four times as abundant as uranium, can be exploited as an alternative fertile nuclear fuel. The pressure tube heavy water reactors (PT-HWR) is an excellent mainstream technology for harnessing the energy potential in thorium due to its high neutron economy enabled by using heavy water as a moderator and coolant, and by refueling online. The PT-HWR is also capable of flexibility in fuel management, allowing the use of multiple fuel types with different refueling rates. One approach to using thorium-based fuels in PT-HWRs is to use a homogeneous core with a once-through thorium (OTT) fuel cycle, where a fissile material is mixed with thorium (such as (LEU,Th)O{sub 2} or (Pu,Th)O{sub 2}), burned in a high-conversion reactor, and then put into storage for several decades until it is economically attractive to reprocess and recycle. An alternative approach is to use a heterogeneous seed/blanket core, where the seed/driver fuel is made of ({sup 233}U,Th)O{sub 2}, and the blanket fuel is make of ThO{sub 2}, and attempt to achieve a very-high conversion ratio (CR) core, and perhaps net breeding with a self-sufficient equilibrium thorium (SSET) fuel cycle. The objective of this study was to assess the performance of various heterogeneous seed/blanket cores in a 700-MWe-class PT-HWR using ({sup 233}U,Th)O{sub 2} fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Effect of Recrystallization on Gas Bubble Swelling in UMo Fuels

The development of low-enriched uranium (LEU) fuels such as UMo dispersion fuels and monolithic UMo fuels is part of a global effort on nuclear nonproliferation. Although UMo fuels have the advantage of high density, excellent irradiation performance and good thermal conductivity, the volumetric swelling in irradiated metallic fuels is known to be an important design parameter because it affects not only the thermal conductivity of the fuels but also the mechanical integrity of fuel structures. Experiments also show that recrystallization dramatically affects the gas bubble swelling kinetics. A series of materials processes such as casting, mechanical rolling and thermal treatment are used in fuel fabrication. Resulting microstructure, such as grain size, porosity, and second phase, strongly depends on the material processes and process parameters. Therefore, a fundamental understanding how fabricated microstructures and irradiation conditions affect the nucleation of gas bubbles, recrystallization and gas bubble swelling kinetics is essential for optimizing material processes to obtain desired microstructure and fuel performance. Great effort has been made in the development of modeling capability for predicting gas bubble swelling kinetics and fission gas release kinetics in nuclear fuels. However, these models are based on the assumptions that the inter- and intra-granular gas bubble sizes are independent of position and only depend on time or fission density. Therefore, they are applicable to a material with uniform microstructure such as equiaxed grains with homogeneous defect distributions, homogeneous thermodynamic and kinetic properties of defects, and uniform radiation conditions. Very recently, we developed a three dimension (3D) microstructure dependent gas bubble swelling model for investigating the effect of heterogeneous microstructures and inhomogeneous thermodynamic properties of defects on gas bubble swelling in polycrystalline UMo fuels. However, the model doesn't take into account the effect of recrystallization on swelling kinetics. In this work, the spatially dependent rate theory is used to describe the evolution of interstitials and interstitial loop density. Recrystallization is described by a phase-field model with a concept of 'phase transition' when the local interstitial loop density is larger than a critical value. The intra- and inter- granular gas bubble evolution is described by the spatially dependent Booth model. A model integrating all the microstructure evolution enables one to study the effect of heterogeneous microstructures, recrystallization, and the metallic fuel thermodynamic and kinetic properties on gas bubble swelling kinetics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

A New Look at Historic CRAC 'Divergence' Experiments - Implications for a CAAS Minimum Accident of Concern

This paper examines power history parameters and results based on historic fissile solution criticality 'divergence' experiments (referenced here as Divergence Experiments) from the 1960's-70's' Consequences Radiologiques d'un Accident de Criticite' (CRAC) campaign. The results from specific Divergence Experiments have implications for an ANSI/ANS-8.3 (1997) Criticality Accident Alarm System (CAAS) Minimum Accident of Concern (MAC). Specifically, this paper examines select CRAC Divergence Experiments in conjunction with 'qualifying reactivity additions' for a MAC defined by an absolute time constraint. A previous paper presented at the 2011 International Conference on Nuclear Criticality (ICNC 2011) illustrated the differences in 'qualifying reactivity additions' above delayed critical (and below prompt) for a variety of MAC constructs. That paper was the first to examine delayed critical kinetics in greater detail over a range of process and MAC considerations. The ICNC2011 work did not consider an absolute MAC time constraint from the excursion onset and assumed an initial excursion power. That study evaluated MAC constructs based on 'relative time differences' between reaching a specific fission rate and total fissions. The work presented here thus picks up where the ICNC2011 study left off. The primary focus of this work provides simulated power histories for specific Divergence Experiments that served as a basis for justifying a MAC noted in ANS-8.3 Appendix A. This paper documents computational modeling and dynamic multiphysics simulation with known Divergence Experiment parameters for reactivity additions at, below, and above the reported terminal reactivity. The initial step considers replicating what could be a reasonable power history for the terminal reactivity addition to the extent the same peak power and total fissions are obtained over the total time considered. Power histories for all other reactivity additions are simulated by varying the input reactivity and solution heights for the same model. A relationship is obtained between reactivity addition and the time to reach specific MAC power and fission levels within a given time constraint (e.g. 60 s). The 'qualifying reactivity addition' is determined for the total MAC fissions within that time. This value is then compared to the historic terminal value (e.g. $0.33) or any other values (e.g. 'excess reactivity of only a few cents') noted in ANS-8.3 Appendix A. In general, calculated MAC power levels have ranged from 1.8 kW to 173 kW depending on material type and form, H/U, enrichment, critical volume, and dose response function. A minimum accident consideration noted a SHEBA experiment with an excess reactivity 7 cents above delayed critical (TD ∼110 s). This excursion provided an additional 10{sup 15} fissions after an undetermined buildup time resulting in a 'few times 10{sup 15} fissions'. ICNC 2011 LEU solutions calculations show 0.2 Gy in air corresponds to 1.4 x10{sup 16} fissions. Thus, while historically, 0.2 Gy has been loosely associated with 'several times 10{sup 15} fissions', in reality, the actual total fissions could be an order of magnitude larger, and the power associated with 0.2 Gy/min (7.5 kW) larger as well. The reactivity addition required to obtain the actual 0.2 Gy total fissions over 60 s could be much larger than considered. Another reason the qualifying reactivity level could be larger than previously noted, involves temperature reactivity feedback. Compensated reactivity results in a departure from a pure exponential power rise, increasing the doubling time (TD) as the excursion approaches peak power. Uncompensated reactivity additions and constant TD have been traditionally assumed for a pure exponential power rise through the kW range to determine either total MAC fissions, power, or arrival time to these values.

42 ENGINEERING↗

DOE OSTI2017

LOCA Analysis for the NIST Research Reactor

The research reactor (aka NBSR) at the National Institute of Standards and Technology (NIST) is a heavy water cooled, moderated, and reflected tank-type research reactor that operates at atmospheric pressure and a design power of 20 MW. Core power is generated in 30 fuel elements, each with 17 fuel plates in an upper section and a separate lower section. Power is removed using forced convection and heat exchangers in the cooling loop. The licensing analysis of the loss-of-coolant accident (LOCA) in the NBSR was addressed in the current Safety Analysis Report (SAR). The SAR discusses the low probability of a significant pipe break: 'the main piping is located in protected areas, system pressures are low, and flow rates are small so that wear is not an issue.' Therefore, the probability of a large break, including a double-ended guillotine break, is extremely low. For smaller breaks where the operator has time to take action, procedures are in place to 1) mitigate a loss of water by tripping pumps and closing control valves to isolate the reactor vessel after the falling water level in the vessel is detected by instrumentation, and 2) ensure that emergency cooling water continues to flow for as long as needed. The SAR also refers to an analysis of why there is sufficient water without forced cooling to cool the fuel elements (FEs) in a LOCA if the flow channels between two fuel plates (or a fuel plate and an outside plate) remain filled with water. That analysis was based on a flooding-limited critical power but assumed that reactor trip had taken place. A more recent study showed that some scenarios may lead to the draining of water from the coolant channels. In these cases, the cooling on the inside of the fuel element is from emergency cooling water that falls only along the inner surface of one of the side plates in the fuel elements; the remaining surfaces within the fuel element being exposed to gas. The analysis therein was done for both the present high-enriched uranium fuel (HEU) and the proposed low-enriched uranium fuel (LEU). The LOCA analyses done previously assumed that there is sufficient cooling during the short-term, i.e., before the water level reaches the top of the fuel plates and begins to drain from the coolant channels or before decay heat becomes low enough such that the flooding-limited critical power shows that there is no problem. The present study is to determine if that assumption is correct, i.e., to consider the situation immediately after the break, before and after reactor trip when the power is still relatively high.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Use of spectral shift to greatly improve uranium utilisation in FHR designs

Fluoride-salt-cooled high-temperature reactors (FHRs) are thermal-spectrum nuclear reactors defined by their use of liquid-fluoride-salt coolants, together with tri-isotropic (TRISO)-coated particle fuels and graphite moderator materials (Forsberg, 2004). FHRs are also characterized by their improved thermodynamic efficiency relative to gas-cooled reactors. The liquid salt coolant is designed to operate with high core outlet temperature and nearly atmosphere pressure. These features allow the FHR to possess higher core power density compared to gas-cooled TRISO-based systems. However, industry fabrication limitations result in a low volumetric fraction of fuel kernel in TRISO particles. In order to obtain reasonable operation periods, the enrichment needs to be in the high LEU (low enriched uranium) range as was proposed in the original design (Greene, 2010). Recent studies (Ilas, 2014) presented newly repurposed SmAHTR-CTC reactor concept that relies on prior experience obtained from the 2010 initial SmAHTR design (Greene, 2010). This design consists of 19 hexagonal fuel assemblies arranged in a 3-ring hexagonal geometry with 18 fuel plates each. The active length of fuel plates in each fuel assembly is 4 meters. The proposed SmAHTR-CTC achieves substantial improvements from the neutronic performance standpoint over the original concept. The results indicate that a fuel cycle length of over two years with discharged fuel burnup of 40 gigawatt-day per metric ton of initial heavy metal (HM) is possible with an 8% fuel enrichment in a once-through fuel cycle. However, the lower uranium enrichment loading requires a more frequent refueling strategy and hence considerably reduces the cycle length. It is necessary to achieve reasonable cycle length values but at the same time to extend the residence time of the fuel, especially since the costs associated with TRISO-fuel are substantially higher (Shropshire et al., 2008) than traditional UO{sub 2} fuel ($10,000/kg for TRISO vs $200/kg for UO{sub 2}). This paper presents an approach to extend the operational time and improve fuel utilization of FHR by means of spectral shift control. This is achieved by incorporating movable graphite structures into the design, instead of the static graphite blocks presently envisaged. At beginning of life (BOL), the graphite structures are withdrawn. The reactor is under-moderated, with excess neutrons being primarily captured in {sup 238}U, breeding {sup 239}Pu. Towards the end of life (EOL), the graphite structures are inserted, thermalizing the neutron spectrum and increasing reactivity. The extra {sup 239}Pu bred during the cycle is then burned, allowing the cycle and discharge burnup to be extended. Such movable structures regime should be feasible from an engineering point of view, given that it is broadly equivalent to inserting and withdrawing control rods. The preliminary results indicate that such a design may considerably improve the fuel utilization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Integrating Sudo-Kaminaga Correlation to the Safety Analysis Code PARET-ANL

The accuracy and reliability of the reactor safety analysis code are critical because the specifications of the safety systems will depend on the analytical results for initiating events that could occur to the reactor. In safety analyses for power reactors, there exist many proficient and applicable computer codes such as RELAP5, CATHARE, RETRAN, or CATHENA. For non-power research reactors, an issue arises, however, due to lacking of such specialized codes. As a result, computer codes developed for the transient analysis of power reactors need to be applied carefully after proving their applicability to a specified research reactor. The safety analysis discussed in this paper concerns the prevention of fuel damage in hypothetical accidental scenarios for the new research reactor design at National Institute of Standards and Technology (NIST). One fuel integrity criterion is analyzed by investigating the Minimum Critical Heat Flux Ratio (MCHFR) in the core during the accidental transients. Some previous studies develop a relationship that correlates the nominal value of the minimum CHFR with the probability of CHF occurring and causing fuel damage. Using a statistical approach, the nominal value of the MCHFR for the low-enriched uranium (LEU) reactor needs to remain above the recommended limiting value to better ensure the safety of the research reactor. The probability of no fuel failure increases as the MCHFR increases. More importantly, as long as the MCHFR remains above the recommended limiting value of 1.301, the reactor will have a 90% probability that CHF is not reached. The safety analysis program PARET, developed by Argonne National Laboratory (ANL), is employed to obtain the CHFR information in this study. PARET/ANL is a digital computer programming code intended primarily for the analysis of test and research reactors that use plate-type (flat) fuel elements. This program has its own set of correlations that are used to calculate the CHFR. One of them is the Mirshak correlation, which was believed to provide the best estimation of the critical heat flux under the operating conditions of the new reactor at the time. Following the critical heat flux (CHF) experiments for vertical rectangular channels in the JRR-3 (Japan Research Reactor unit 3), the CHF calculations now use the Sudo- Kaminaga correlation. This correlation is of our interest because this method has an enhanced geometric similarity and an increased range of applicability that are more representative of the actual operating conditions of our current design, and has a more mechanistic approach [9]. Since the source code of PARET/ANL is inaccessible, a MATLAB-based utility has been written to integrate the Sudo-Kaminaga correlation as a substitute for the Mirshak correlation in the code. As a result, both the Mirshak and the Sudo-Kaminaga correlations are used to analyze the standard output from the PARET program to calculate the CHFR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Use of tsunami in the validation of scale 6.1 for the Swedish spent fuel repository - 14381

In March 2011 SKB applied to the Swedish Radiation Safety Authority (SSM) for a permit to build a final repository for spent Fuel and an encapsulation plant (Clink). The application contained a criticality safety analysis for the canisters to be used in Clink and the copper canister to be used in the final repository. In 2012 SSM requested several updates to the application. One of SSM's concerns was the validation of the codes. SSM emphasized that SKB must improve the method to select critical experiments to the validation suite. To satisfy this demand, SKB used the tool Tsunami in the Scale 6.1 code package to select experiments from the handbook of criticality experiment (IHECSBE). With help of sensitivity files SKB evaluated almost 700 experiments from the LEU-COMP-THERM library. SKB then picked the experiments that most resembled our safety cases. The sensitivity coefficient, ck, from Tsunami has been used to evaluate the similarity between the experiments and the applications. It has been considered that a ck over 0.8 provides a 'good enough' similarity between the experiment and our evaluated safety cases. For the different safety cases between 50-600 experiments satisfied the 0.8 criteria giving a good base of data for statistical analysis. It is worth noticing that PWR-fuel generally receives a higher ck than BWR fuel. Initial analysis shows that the differences in results between BWR and PWR can be explained by a larger sensitivity for the U{sup 238} n,n' part of the cross section in the PWR cases. To make sure the validation covers all materials and required physical parameters a gap analysis was made using engineering judgement by criticality safety experts. From this exercise additional experiments were added to the suite. This paper will discuss the Tsunami results for different safety cases and the reasons for differences between them. It will also present the final results of the validation and discuss experiments that were excluded from the validation suite. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Sensitivities and correlations of critical experiments due to uncertainties of system parameters and nuclear data - 14416

With increasing demand on accuracy of criticality calculations, correlations between benchmark experiments due to system parameter uncertainties become of interest in bias determination. The impact of different interpretations of system parameter uncertainties on the calculation of covariances and correlations of k{sub eff} is discussed in this article. Following the suggestion of the OECD/NEA EGUACSA Benchmark phase IV [1], 21 experiments from the ICSBEP international Handbook of Evaluated Criticality Safety Benchmarks [2] are used for this analysis (LEU-COMP-THERM series 07, exp. 01-04 and 39, exp. 01-17). The experiments share certain assembly and fuel features and were carried out in the same lab. Therefore, correlations in the calculation of k{sub eff} appear. Correlation coefficients between experiments due to nuclear data uncertainties are calculated with TSUNAMI [3] and show high values for most experiments. For the correlation analysis due to system parameter uncertainties mainly two scenarios are considered: In scenario A all uncertain parameters are varied mutually for all fuel rods, representing a systematic deviation of the actual experimental setup from the design value. This analysis is performed for all experiments listed above. In scenario E certain parameters like the outer cladding diameter are varied for each fuel rod individually, representing a random scattering of the varied parameter. Scenarios B, C and D represent intermediate scenarios between A and E, which are discussed shortly. Uncertainty and sensitivity calculations of k{sub eff} on the uncertain experimental parameters based on Monte Carlo sampling are presented and discussed. The calculations were performed with SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool [4]) and SCALE 6.1.2 [3]. For the 21 EGUACSA proposed experiments the outer cladding diameter is the leading effect for the uncertainty of k{sub eff} for scenario A. In contrast, for scenario E the impact of the cladding diameter on k{sub eff} cancels out due to rod wise variations leaving the fuel diameter as the leading effect. The correlation analysis based on system parameter uncertainties indicates a significant dependence of the correlation coefficients on the interpretation of the uncertainties. It shows high values for scenario A, if the neutronic features are similar (EALF, H/U). For scenario E the overall correlations are significantly smaller and variations between experiments decrease. The impact of the determined correlations on bias calculations of application cases will be discussed by the authors in the accompanying paper 'Impact of correlated data in validation procedures'. (authors)

42 ENGINEERING↗

DOE OSTI2015

Impact of correlated data in validation procedures - 14483

The increasing demand on accuracy of criticality calculations leads to the question, how to treat correlated data of benchmark experiments correctly. Correlations in the process of validation arise for example if a configuration in a series of experiments shares certain components, e.g. fuel rods. Traditional methods of code validation do not treat correlations explicitly, but give a more conservative estimation of the bias. Including correlations can lead to a more precise estimation. In the work at hand, a method is discussed based on Bayes theorem to include correlations in the experimental data when estimating an application case k{sub eff}. After a short introduction to the applied method and the demonstration of the general capability of it to predict the k{sub eff} correctly, it is shown how the calculated k{sub eff} of an application case depends on correlations in experimental data for a simple Toy Model. The impact of ignoring correlations and taking to account lower and higher correlations is shown and discussed. Further results are shown using a database of correlated experimental data from the International Handbook of Criticality Safety Experiments to estimate the bias of two application cases. The experiments were chosen following the suggestion of the OECD/NEA EGUACSA Benchmark phase IV and consists of 21 experiments from the LEU-COMP-THERM series 007 and 039. With the extracted data results the calculated k{sub eff} values of the application cases are shown and discussed for different examples of correlations. The calculations were done by applying Monte-Carlo sampling methods with the GRS tool SUnCISTT, using ORNL's SCALE 6.1.2. This work discusses a possible way of treating correlations in the recalculation of critical experiments to get a more precise bias estimation of the used code system. It will be shown, that the presented method is capable to treat similarities in experimental setups of data used to calculate the bias correctly, independent of the size of similarities. The presented method is in principle capable of assessing regimes with little experimental data to estimate a bias. (authors)

42 ENGINEERING↗

DOE OSTI2015

The composition of seawater and the effect of seawater immersion on reactivity - 14499

Following the Fukushima Daiichi accident a need arose to perform defense-in-depth analyses to assess the effect of seawater immersion on the reactivity of nuclear fuel. This paper derives a composition for Standard Seawater based on the Reference Composition using the International Thermodynamic Equation of Seawater - 2010 (TEOS-10). The Reference Composition for seawater is based on measurements of the Atlantic surface seawater and assumes no organic or inorganic material is entrained in the seawater. Using the Reference Composition, a methodology for determining the composition of seawater as a function of temperature, pressure and salinity is developed. These seawater compositions may be used in nuclear criticality safety analyses to determine the effect of seawater immersion on reactivity. Results of a sensitivity study are presented that illustrate the effect of seawater immersion and salinity on the reactivity of a representative low-enriched uranium critical assembly, LEU-COMP-THERM-002. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Evaluation of an unexpected reflector temperature effect for water-isolated array configurations - 14500

It has been conventional wisdom in the criticality safety discipline that for spent fuel systems like water pit storage and transportation, modeling water at its peak density will maximize reactivity. Thus, modeling water at 3.96 deg. C will produce a conservatively bounding k{sub eff} for that configuration. This assumption has been found to be incorrect for certain high-leakage water-isolated array configurations. In fact, temperatures up to the boiling point of water can in fact be more reactive, and the most reactive temperature is highly dependent upon the specific fuel configuration being analyzed. In light of the accident at Fukushima-Daiichi - where the spent fuel pool temperature increased - knowledge of this unexpected behavior is essential to ensuring that criticality safety evaluations are conservatively bounding. To illustrate the reflector temperature effect, MC21 calculations are provided for a sensitivity study based on LEU-COMP-THERM-044, which models the IPEN/MB-01 reactor at various temperatures between 3.9 deg. C and 93.3 deg. C. (authors)

42 ENGINEERING↗

DOE OSTI2024

Generation of Enrichment-Dependent Thermal Neutron Scattering Data

This work details the generation of enrichment-dependent thermal neutron scattering cross sections for several crucial uranium fuel compounds. The evaluations of the thermal scattering law (TSL) and associated cross sections for uranium dioxide (UO 2 ), uranium carbide (UC), and uranium nitride (UN) were performed using standard ab initio lattice dynamics (AILD) methods. The data for uranium metal was produced using a novel hybrid approach of molecular dynamics combined with lattice dynamics methods. 235 U enrichments of 5%, 10% (LEU+), 19.75% (HALEU), 93% (HEU), and 100% were considered, in addition to natural uranium. The enrichment-dependent masses and free atom cross sections were used in the generation of elastic and inelastic thermal neutron scattering cross sections, while the calculation of the phonon density of states (DOS) and resulting TSL considered only the natural isotopic composition of uranium. The use of an identical DOS for all enrichments is expected to have minimal impact on the final data, as the small change in uranium mass should not significantly affect lattice vibrations. The cross sections are shown to exhibit significant dependence on 235 U enrichment. The submission of this data to the National Nuclear Data Center (NNDC) for release in the ENDF/B-VIII.1 database should support the design of advanced reactor concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Development of New Reactor Core Configuration for Power Uprate - Fuel Reload & Heat Processing Analyses, Core Design, System Safety Assessments, and Fuel Performance Analyses

With the passage of the Infrastructure Investment and Jobs Act in 2021 and the Inflation Reduction Act (IRA) in 2022, the United States stands at a critical juncture for the future of nuclear power. These landmark policies provide significant support for clean energy initiatives, positioning nuclear power as a key component of the nation’s strategy to reduce carbon emissions and achieve energy security. This growing emphasis on nuclear energy is driven by the need for reliable, low-carbon power sources as the country transitions away from fossil fuels. Federal policy, along with increasing state-level support, is encouraging investment in nuclear technology advancements to meet these demands. Building new nuclear power plants (NPPs), however, presents significant challenges due to high costs and long construction timelines. As a result, increasing the power output of existing NPPs through power uprates has emerged as a more feasible and cost-effective strategy. One key area of advancement is the development of accident-tolerant fuel (ATF), such as chromium-coated zirconium alloy cladding, which offers enhanced material performance, enabling power uprates in light water reactors (LWRs). Given the growing demand for nuclear energy fueled by federal policies and state initiatives, it is essential to evaluate the feasibility and benefits of significant power uprates in existing pressurized water reactors (PWRs) using advanced fuel technologies. The introduction of ATF concepts opens new opportunities for safely and economically achieving these power increases. Assessing whether these innovations can support substantial power uprates while maintaining operational safety is crucial to maximizing the potential of the nation’s existing nuclear infrastructure. This project aims to explore how power uprates can be achieved by boosting reactor thermal power output and optimizing reactor core design, while ensuring the safety and economic viability of NPPs. Specifically, it will focus on demonstrating the technical and economic feasibility of power uprates in a PWR using low 5-10% enrichment uranium (LEU+) high burnup (HBU) fuel combined with ATF concepts. In fiscal year 2024 (FY24), the research and development focus on building foundational models and conducting multi-physics performance and safety analyses to support the power uprate. The findings of the study would be shared through LWRS Seasonal Meetings, conferences and workshops with utility companies and researchers. These also serve as a basis for further study of fuel reloading optimization with ATF claddings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Preliminary MITR Design-Demonstration Element Thermal-Hydraulics and Structural Analyses

The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). Massachusetts Institute of Technology Reactor (MITR) is one of five research reactors selected for this program. The objective of this report is to provide preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the MITR Design Demonstration Element (DDE) under conservative approximations for the plate power distribution. This report provides details on the modeling approach and the simulation results obtained, including pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment for the irradiation cycles in the Belgian Reactor (BR)-2.

42 ENGINEERING↗

DOE OSTI2022

Margin to Onset of Nucleate Boiling studies for MITR and NBSR Design Demonstration Experiments

The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). The Massachusetts Institute of Technology Reactor (MITR) and the National Bureau of Standards Reactor (NBSR) are two of five research reactors selected for this program. Previous studies have addressed the thermal-hydraulics performance of the fuel Design Demonstration Elements (DDE) for these reactors under conservative assumptions in the Belgian Reactor (BR)-2. This report extends these previous studies by analyzing the margins to Onset of Nucleate Boiling (ONB) for these two DDEs under the conservative conditions used in the previous thermal-hydraulics analyses

42 ENGINEERING↗

DOE OSTI2023

Preliminary NBSR Design-Demonstration Element Thermal-Hydraulics and Structural Analyses

1.1 OBJECTIVE The United State High Performance Research Reactor (USHPRR) program aims to eliminate more than 200kg of High Enriched Uranium (HEU) from commerce annually by converting five U.S. high-performance research reactors and one associated critical assembly to Low Enriched Uranium (LEU) fuel using a high-density alloy of uranium-10 wt% molybdenum (U-10Mo). The National Bureau of Standards Reactor (NBSR) is one of five research reactors selected for this program. The objective of this report is to provide preliminary thermal-hydraulic and mechanical analyses of the hydrodynamic effects in the NBSR Design Demonstration Element (DDE) under conservative approximations for the plate power distribution. This report provides details on the modeling approach and the simulation results obtained, including pressure, flow velocity, temperature, and oxide layer over the design demonstration experiment for the irradiation cycles in the Belgian Reactor (BR)-2.

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