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At least 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

The impact of chromium coatings on Zircaloy cladding deformation behavior under reactivity-initiated accident-like mechanical loading conditions

A reactivity-initiated accident (RIA) occurs when a control rod ejection or control blade drop causes an increase in the fission rate. The injection of energy results in an increase in fuel temperature which in turn causes rapid thermal expansion of the fuel pellet. This thermal expansion may result in pellet-cladding mechanical interaction (PCMI) in which the fuel imparts a mechanical strain to the cladding. PCMI may cause the cladding to fail, and thus, the mechanical response of cladding due to PCMI must be investigated when characterizing new cladding materials. Chromium-coated Zircaloy-4 is a near-term accident-tolerant fuel cladding that exhibits improved high-temperature oxidation resistance. Modified burst testing was utilized to experimentally simulate the effects of PCMI on both uncoated and chromium-coated Zircaloy cladding samples at hot zero power conditions. Samples were coated using either cold spraying or physical vapor deposition to understand the differences in behavior that the coating application method may cause. Digital image correlation was used to analyze images of the deforming specimens to extract the in-situ strain behavior of the cladding. The uncoated specimens burst at hoop strains ranging from 8.8 % to 17.2 %. The cold-spray chromium-coated Zircaloy specimens burst at hoop strains of 7.0 % to 11.0 %. The physical vapor deposition coated tubes burst at hoop strains of 9.1 % to 11.5 %. These results indicate that the chromium coating causes a loss in the ductility of the cladding. Furthermore, the higher burst hoop strains of the physical vapor deposition-coated samples relative to the cold-spray samples indicate that the cold-spraying technique causes more of a loss in ductility than physical vapor deposition. All samples burst at higher hoop strains than those expected to occur in an RIA for fresh cladding.

Accident-tolerant fuel↗
DOE OSTI2024

Measuring thermal diffusivity and gap conductance in uranium nitride and Zircaloy relevant for microreactor applications

Heat transfer across nuclear fuels and structural interfaces is an important factor for evaluating the performance of nuclear power systems. Specifically, heat generated as nuclear fuel fissions must be transported through the cladding material and through the reactor to reach the steam turbine for power generation. As new microreactor designs emerge, maximizing the efficiency of this heat transfer process becomes crucial to make them commercially viable. This article examines thermal diffusivity and gap conductance in uranium nitride (UN) fuel and Zircaloy-4 (Zry4) cladding using light flash analysis (LFA). Thermal diffusivity measurements were made on monolithic UN pellets and Zry4 exposed to carbon at peak operating temperatures of microreactors and show that carbon ingress has a minimal effect on thermal diffusivity when compared with identical materials not exposed to carbon. Evaluation of gap conductance at the UN-Zry4 interface was done using one-dimensional two-layer thermal transport models as a function of applied pressure. Here the results show that increasing pressure on the UN-Zry4 interface leads to gains in gap conductance per unit area in fuel-cladding assemblies at microreactor operating temperatures. While many other variables are expected to influence UN-Zry4 interfacial gap conductance (e.g. contact surface roughness, porosity, localized heating, environmental gas pressure), the work offers a demonstration of using a conventional LFA apparatus to determine this parameter at elevated temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

PNNL FY 2022 Sibling Pin Testing Results

This report presents the results of testing two rods after heat treatment at 400C for 8 hours and compares to the results from FY 2021 of similar rods that were tested as baseline. The heat treatment resulted in a decrease in yield stress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1971

Molybdenum-UO2 cermet irradiation at 1145 K.

Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

Mcdonald, G.↗
DOE OSTI2022

Milestone 1.2.10: Steady-state H 2 “roll over” point data for aluminum alloys 1100 and 6061

Extended (> 50 years) dry storage is being evaluated by the U.S. Department of Energy (DOE) for the disposition of ~ 18 metric tons of aluminum-clad spent nuclear fuel (ASNF). Transition of the current ASNF inventory into dry storage—using the standard DOE canister—necessitates a rigorous, predictive understanding of the long-term physical and chemical factors that may influence the integrity of the proposed storage canister, including radiolytic molecular hydrogen (H 2 ) generation. Current model predictions employ initial radiolytic yields of H 2 , the values of which change as the cladding’s H 2 -precursor inventory is depleted and H 2 itself becomes progressively more involved in radiolytic and surface dissociation processes. Consequently, the absorbed radiation dose that this steady-state H 2 yield corresponds to is essential for the evaluation and improvement of model predictions. Here, we report our findings on the long-term generation of H 2 from the gamma irradiation (≤ 36 MGy) of corroded AA1100 and AA6061 coupons in helium environments at ambient temperature and ~ 50% RH. Our findings show that AA1100 systems reached steady-state by ~ 36 MGy, while higher doses were necessary for AA6061 systems. This discrepancy was attributed to the AA6061 coupons developing a thicker corrosion layer that led to the trapping of H 2 and its precursors, and potentially additional chemistries, ultimately delaying the depletion of H 2 precursors and the system’s “roll over” point. Further, current model predictions—based on previous AA1100 data—do not show steady-state attainment until above 120 MGy, which is not the case for the AA1100 data collected here. Consequently, the new alloy dependent data presented here are important for the continued improvement of predictive computer models for evaluating the feasibility of extended storage of ASNF in helium backfilled canisters.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

Advanced Test Reactor (ATR) fuel has been identified as a resource for high-assay low-enriched uranium (HALEU) production. A survey was performed on the published literature describing ATR fuel. The geometry of the fuel is complex; different parts of the fuel compact experience differing neutron flux and burnup. The literature is sparse, and access is controlled. Therefore, fundamental studies of fuel reprocessing must use a model fuel that represents the main chemical and structural features. Advanced chlorination, or chlorination with sulfur-chlorine bearing reagents is being investigated as way to separate the fuel from metal matrix alloys. A UAl x alloy will be fabricated with x = 3, 4, and 5. The potential chlorination of individual UAl x intermetallics will be assessed in the advanced chlorination process of Al-8001 and Al-6061 as well as a representative mixture. Initial studies will track the alloying elements of the Al, which are Si, Fe, Cu, Mn, Mg, Cr, Zn, and Ti, in addition to the U itself. Further studies will include fission product simulants. Because advanced chlorination solvents include sulfur, the chemistry of sulfur with major and minor constituents will also be investigated. The experimental work accompanied by neutronic calculations will allow the assessment of the feasibility of advanced chlorination to separate aluminum from uranium. If bench-scale testing appears promising, then small-scale tests in shielded facilities with irradiated cladding, lightly irradiated fuel, and spent nuclear fuel are recommended to track the complete inventory of fissile actinides, fission product impurities, and reagent solids and liquids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Impact of corrosion layer composition and thermal pretreatment on the radiolytic generation of molecular hydrogen from aluminum alloys

To fully evaluate the feasibility of extending the dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled cannisters, the amount of radiation-induced molecular hydrogen (H 2 ) generated and the impact of absorbed radiation dose on the ASNF corrosion layer composition must be accurately assessed. Here, in this study, we report a slowing in the rate of radiolytic H 2 generation for pre-corroded AA1100 and AA6061 aluminum alloy coupons irradiated to up to 53 MGy of absorbed cobalt-60 gamma dose. By exploring a variety of thermal pretreatment conditions for AA6061 coupons, we find that the “steady-state” yield of H 2 depends more on the aluminum alloy used than on the treatments applied prior to dry storage. Scanning electron microscopy and positron annihilation lifetime spectroscopy techniques also provided evidence for gamma radiation-induced defects in the corrosion layers of the investigated aluminum alloy coupons for high absorbed doses (~50 MGy), the consequences of which on cladding integrity and H 2 generation should be explored in future works.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2025

An approach for spent nuclear fuel containment integrity verification using gas tagging

Verification of containment integrity is required for spent nuclear fuel (SNF) managed by the commercial nuclear industry and U.S. Department of Energy (DOE), especially after extended storage. Certain SNF storage systems, such as the DOE road-ready dry storage system, hold several packaged containments within a welded over-canister. These packaged containments are called Department of Energy Standard Canisters (DOESCs). DOESC leakage identification is challenging because their containment boundary cannot be accessed for testing and their contents (i.e., SNF and fill gas) are often similar. There are concerns that this could result in costly characterization and repackaging operations of DOE road-ready dry storage systems if compromised DOESCs are suspected. Here, to address these concerns, this paper presents an approach for applying a gas tagging process using xenon to uniquely identify compromised inaccessible containments following extended storage. The containments considered for this application are seven DOESCs, each packaged within a single over-canister. Two different SNF loading configurations from the Advanced Test Reactor and Fort Saint Vrain nuclear power plant are considered. These configurations are used to represent research reactor aluminum-clad spent nuclear fuel (ASNF) and TRi-structural ISOtropic (TRISO) SNF types. Results for this application show that for ASNF and TRISO type fuels for which the selected fuels are representative, the volume of taggant required at loading is determined primarily by the lower detection limit and leak rate of taggant from a compromised DOESC, rather than the amount of fission-generated xenon in the loaded fuel. While the application presented is suited for larger leaks, smaller leaks could be detected by modifying certain design parameters. This gas tagging approach can also be applied to other DOE containments and advanced reactor SNF storage systems.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2021

Alloying of U-Al-SS as a Simulant for Pu-Al-SS Alloying

The Savannah River National Laboratory evaluated several options for disposition of stainless steel (SS)–clad plutonium metal, particularly Pu-10.6 at. % Al (Pu- 1.3 wt% Al) alloy fuel. One technology considered was alloying fuel with SS. The goal of the alloying would be to make a SS-Pu alloy that was a nonproliferable waste form with secondary Pu-rich microencapsulated regions distributed throughout the refractory SS. The microencapsulation of the Pu regions should therefore allow the waste form to meet the requirements for a low attractiveness waste as defined by the U.S. Department of Energy. Plutonium-bearing alloys at these levels could potentially be suitable for disposal at a waste isolation pilot plant. Four metal ingots were successfully fabricated using U and Al as a surrogate for Pu-Al. The U was distributed and microencapsulated by the alloy matrix, thereby setting the stage for subsequent tests using SS-clad fuel elements containing Pu-10.6Al.

36 MATERIALS SCIENCE↗
DOE OSTI2020

EVALUATION OF HOT CHANNEL FACTOR FOR SODIUM-COOLED FAST REACTORS WITH MULTI-PHYSICS TOOLKIT

The evaluation of hot channel factor (HCF) is of great significance to the quantification of safety margins for reactor designs. In this paper, HCFs for a sodium-cooled fast reactor (SFR) are evaluated with the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit, which is developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of DOE for multi-physics reactor performance and safety simulations. The high-fidelity neutronics and thermal hydraulics solvers PROTEUS and Nek5000 in the SHARP toolkit are coupled to perform the multi-physics simulations for HCF evaluation. The HCFs induced by cladding manufacturing tolerance, fissile content mal-distribution, wire orientation and uncertainties on the cladding, coolant, and fuel properties are evaluated for a reference core SFR design (AFR-100). The HCFs calculated with the SHARP toolkit are compared to legacy HCFs for similar reactor types. The comparison demonstrates the reduction or elimination of modeling uncertainties in the calculation of HCFs using high fidelity advanced modeling and simulation tools without the need of expensive experiments. Moreover, the reduction of the uncertainties on HCFs evaluation allows an increase in nominal parameters and safety margin, which in turn improves the economic competitiveness of the SFR.

Hot channel factor (HCF)↗
DOE OSTI2020

Evaluation of Radiolysis Data for Hydrogen Gas Generation During Gamma Irradiation of Pre-Corroded and Pristine Aluminum Samples - An Aluminum SNF Dry Storage Study Interim Report

Information and data from radiolysis testing to measure hydrogen (H₂) generated from hydrated oxides on aluminum exposed to ⁶⁰Co radiation were reviewed to evaluate hydrogen generation with radiation dose. Hydrogen generation rate is a primary input to the coupled thermal-chemical system model of the performance of aluminum-clad spent nuclear fuel in a dry storage canister (ASNF-in-canister) over its storage life. Hydrogen generation data and hydrogen generation rate (hydrogen generated per absorbed radiation dose, $\Delta$(H₂)/$\Delta$(dose)) were reported in two reports for radiolysis testing [1, 2]. The testing involved first-time data¹ for radiolytic yield of hydrogen from hydrated-oxides-on-aluminum substrates, and from pristine (non-corroded) aluminum substrates. The laboratory-grown hydrated-oxides-on-aluminum substrates contained a ~ 5 μm film consisting of the trihydroxide bayerite (Al(OH)₃) with amounts of the oxyhydroxide boehmite (AlOOH). These specimens were assumed to also contain physisorbed water (unquantified). The pristine substrates were flat coupons of aluminum that were ground to a 600-grit finish but were not immersed in water to grow a hydrated oxide. The pristine substrates were tested as companion specimens to provide information on hydrogen generation from material without chemisorbed water and were assumed to contain physisorbed water (unquantified).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Measurement of radiolytic hydrogen generation and impact of drying treatments on reactor exposed and surrogate aluminum materials

Technical challenges associated with dry storage of Aluminum-clad Spent Nuclear Fuel (ASNF) include a need to better understand the potential extent of gaseous molecular hydrogen (H 2 ) production through radiolytic degradation of the aluminum (oxy)hydroxide films present on ASNF. The characterization of radiolytic gas generation from ASNF (oxy)hydroxide layers has been identified as a key knowledge gap which poses a technical challenge to the long-term storage of ASNF. Task 2 of the action plan for the extended (>50 years) dry storage of ASNF addresses this gap. Previous radiolysis studies performed under Task 2 established baseline estimates of H 2 generation rates from the attendant hydrated oxides. The results confirmed that net radiolytic H 2 production has a dependency on absorbed gamma dose, as well as relative humidity and cover gas composition (air, nitrogen, and argon). Further experiments revealed that the physisorbed water on the samples may significantly impact the radiolytic H 2 yield. This phenomenon complicates the determination of H 2 generation rates for hydrated oxides, which may lead to inaccurate modeling predictions of the long-term H 2 yields in sealed storage systems containing ASNF, particularly when compared to spent fuel casks which have undergone some drying process intended to remove physically- and chemically-bound water. This report describes the testing methods utilized and the hydrogen generation results obtained in an investigation of the effects of gamma irradiation on aluminum materials for a variety of (oxy)hydroxide surface compositions and drying conditions. The testing methods included small-area aluminum material testing in ampules, and large-area aluminum material testing in steel vessels. Test material preparation, irradiation, and radiolytic H 2 measurement methods are summarized. The measured H 2 concentrations, which reflect the variable initial hydrated inventory and drying treatments, are compared to one another as well as to previously published data, to identify the primary factors affecting radiolytic H 2 generation rates and potential equilibrium H 2 concentrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a Heat Pipe Reactor

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to includ e non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR techno logies under Strategy 2 of the NRC?s near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enha ncements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform heat pipe reactor (HPR) safety evaluations. The generic plant mode l is based on a publicly-available Los Alamos National Laboratory (LANL) Megapower design as modified in the Idaho National Laboratory (INL) Design A description. For plant aspects (e.g., reactor building size and leak rate) that are not described in the LANL and INL references , the analysts made assumptions needed to construct a MELCOR full-plant model. The HP R uses high assay, low-enrichment uranium (HALEU) fuel with steel cladding that uses heat pipes to transfer heat to a secondary Brayton air cycle. The core region is surrounded by a stainless-steel shroud, alumina reflector, core barrel and boron carbide neutron shield. The reactor is secured inside a below-grade cavity, with the operating floor located above the cavity. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing re activity control, the primary and secondary system heat removal, and the effectiveness of th e confinement natural circulation flow into the reactor cavity (i.e., a flow blockage).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Milestone 1.2.12: The Fate of Oxygen Radical Species in Corroded Aluminum Alloys under Irradiation

The U.S. Department of Energy tasked Idaho National Laboratory to evaluate the feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in heliumbackfilled canisters. A significant research effort has been devoted to determining the amount of radiation-induced molecular hydrogen (H 2 ) generation from corrosion layers that would be present on the fuel assembly surfaces. However, limited attention has been directed to the oxygen radical species that are concurrently generated in these H 2 producing radiation environments. This report collates and summarizes the available experimental and computational studies on the fate of these reactive oxygen radicals during the irradiation of aluminum oxyhydroxide and hydroxide polymorphs typically formed on ASNF. Based on this review, it is determined that radiation-induced oxygen-centered radicals do not react to give molecular oxygen gas (O 2 ), but rather remain trapped in the aluminum corrosion layers in their original crystal lattice positions. Overall, O 2 generation will not be a concern for extended storage of ASNF in helium-backfilled canisters. That is, no O 2 contribution to pressurization of sealed dry storage systems or creation of a flammable atmosphere is anticipated.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Roadmap for the Completion of the Design and Implementation of a Remote Canister Monitoring System

The Office of Environmental Management within the Department of Energy is currently investigating technical issues associated with extended dry storage (50+ years) of aluminum spent nuclear fuel. As part of these efforts, a remote canister-monitoring system (RCMS) has been proposed to collect the real-time data of the environmental conditions in the canisters used for the dry storage of aluminum spent nuclear fuel in the CPP-603 facility at Idaho National Laboratory (INL). The proposed RCMS would provide the opportunity to evaluate the appropriate technologies for monitoring, to collect canister environment conditions as soon as possible, to verify and validate current laboratory-based study results and analytic modeling approaches, and to potentially identify additional dry storage options for aluminum spent nuclear fuel at the INL Site. The document provides a brief update on the progress that has been made toward deploying the RCMS and will present at a high level, the general system design philosophy behind the project and the key future tasks necessary for the completion of the RCMS system design, fabrication of the RCMS prototype, testing, and eventual deployment of the RCMS, and a tentative timeline for these tasks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Literature Review Investigating Adsorbents for Mercury Removal from Salt Batch Feed Material

The Savannah River Site (SRS) Liquid Waste System (LWS) contains mercury that originated from decades of canyon processing where it was used as a catalyst for dissolving aluminum cladding of reactor fuel. Since the start of radioactive operations, the Salt Waste Processing Facility (SWPF) has experienced processing difficulties due to the presence of mercury-laden solids. In response to these challenges, Savannah River National Laboratory (SRNL) has been requested by Savannah River Mission Completion (SRMC)/Salt Waste Processing Facility (SWPF) personnel to perform a literature review to assess the feasibility of chemical strikes to salt waste-containing tanks at the SRS as an option to reduce mercury introduction into SWPF. The literature review should evaluate known technologies and chemistries for decreasing mercury concentration and the anticipated efficacy of each technology on various mercury species known to exist in SRS waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Evaluation of the Potential for Precipitation of Solids during Storage of Non-Aluminum SNF Solutions

Non-aluminum clad spent nuclear fuels (NASNF) stored in the L-Area basin will be dissolved in H-Canyon using the 6.3D electrolytic dissolver. The solutions will be stored in either the hot or warm canyon until the preparation of a sludge batch for the Defense Waste Processing Facility. Spent nuclear fuel solutions could be stored for 1-2 years before transfer to the H-Area Tank Farm depending on the interval between sludge batches. The solution level in the storage tanks will be maintained; therefore, precipitation of solids due to evaporation is not an issue. However, the precipitation of solids from completely dissolved SNF due to solution instabilities has been observed during intermediate storage of solutions generating hydrated oxides.The presence of fissile material in these solids is generally associated with zirconium molybdate, which is known to act as a host lattice for Pu and can carry the actinides upon precipitation. The formation of zirconium molybdate solids which carry fissile material is a potential concern for the storage of NASNF solutions. To address this concern, the Savannah River National Laboratory performed a literature review to identify knowledge gaps which may require experimental work to determine if the formation of solids is a concern during storage of these solutions. Based on the literature review, the precipitation of zirconium molybdate solids from the Campaign 1 NASNF solutions during intermediatestorage is expected. This conclusion is supported by the identification of zirconium molybdate solids found on the H-Canyon 6.1D Dissolver MK-12 insert spacer. The formation of the zirconium molybdate solids is attributed to hydrolysis and radiolytic processes in the nitric acid solution. As the molybdate solids form, U and Pu can substitute for Zr in the crystal lattice resulting in co-precipitation. Generally, the Pu substitutes directly into the crystal lattice during precipitation while the U associated with the molybdate solids more likely absorbs from the solution. The U in the NASNF solutions is present as uranyl nitrate, a 2+ cation which will not substitute as easily into the molybdate crystal lattice for the Zr 4+ ion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Requirements Description of GAMSOR and GAMSRC

This report presents the modeling and simulation capabilities of Argonne National Laboratory’s GAMSOR and GAMSRC code that is used in present reactor analysis activities. These capabilities will then be used to establish the set of verification tasks necessary to verify GAMSOR for use on commercial projects. A similar approach was taken for the PERSENT, REBUS and DIF3D software packages. The GAMSOR program is a modified version of DIF3D used to obtain the coupled neutron gamma flux distributions for a given reactor system. GAMSOR was first developed in the 1980s and its work has progressed in parallel with DIF3D throughout its various development phases. While GAMSOR is compatible with DIF3D-FD and DIF3D-Nodal, DIF3D-VARIANT is most commonly used today and will be the focus of the present verification effort. As covered in [1], GAMSOR was built to provide more accurate heating distributions by explicitly solving for both the gamma and neutron heating distributions. Without this consideration, gamma heat is normally assumed to be deposited locally at the site of gamma production. This is acceptable for modeling some phenomena because the total energy production is preserved, but it can lead to inaccurate approximations of fluence-dependent behaviors such as fuel performance, swelling, cladding embrittlement, radiation induced creep, and power peaking. The outputs from GAMSOR are contained in multiple intermediate and final output files: FIXSRC, VARSRC, GTFLUX, GHFLUX, PWDINT, NPDINT, and GPDINT. DIF3D normally produces a PWDINT file containing power density results, with the assumption that gamma rays are absorbed where they are emitted (i.e., energy deposition is local with no modeled transport). With GAMSOR, neutron power density (NPDINT) and gamma power density (GPDINT) are both produced, with their summed power densities replacing the results in PWDINT for total heating (i.e., recoverable) power density.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗