• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1097-1104
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• 2023

A preliminary criticality analysis for novel pyrochemical apparatuses for the reprocessing of mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor was performed. High-temperature processing apparatuses, "metallization" electrolyzer, refinery remelting apparatus, refining electrolyzer, and "soft" chlorination apparatus are considered in this work. Computational models of apparatuses for two neutron radiation transport codes (MCU-FR and MCNP) were developed and calculations for criticality were completed using the Monte Carlo method. The criticality analysis was performed for different loads of fissile material into the apparatuses including overloading conditions. Various emergency situations were considered, in particular, those associated with water ingress into the chamber of the refinery remelting apparatus. It was revealed that for all the considered computational models nuclear safety rules are satisfied.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.19-29
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• 2021

This paper presents the validation of the MCS code for critical safety analysis with burnup credit for the spent fuel casks. The validation process in this work considers five critical benchmark problem sets, which consist of total 80 critical experiments having MOX fuels from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The similarity analysis with the use of sensitivity and uncertainty tool TSUNAMI in SCALE was used to determine the applicable benchmark experiments corresponding to each spent fuel cask model and then the Upper Safety Limits (USLs) except for the isotopic validation were evaluated following the guidance from NUREG/CR-6698. The validation process in this work was also performed with the MCNP6 for comparison with the results using MCS calculations. The results of this work showed the consistence between MCS and MCNP6 for the MOX fueled criticality benchmarks, thus proving the reliability of the MCS calculations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.315-330
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• 2018

A subcritical facility must ensure nuclear criticality safety under all circumstances. For this purpose, it is essential to have a procedure to validate that calculated values do not exceed upper subcritical limit (USL), determined by quantifying the bias and uncertainty. However, there are several validation methodologies of nuclear criticality safety and these can yield different USL. Therefore, it is necessary to analyze the validity of the methodologies to establish one methodology that can provide the most appropriate USL. In this study, two documents, a guide for validation of nuclear criticality safety calculational methodology (NUREG/CR-6698) and a criticality benchmark guide for light water reactor fuel in transport and storage package (NUREG/CR-6361), are compared and analyzed. In particular, the methodology in NUREG/CR-6361 is applied to the USLSTATS code. However, the analysis results show that the methodology in NUREG/CR-6698 is more appropriate, for several reasons. This is applied to decision of USL to design casks using SCALE code version 6.1.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.624-634
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• 2016

Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that $k_{eff}$ values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.709-716
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• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.207-214
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• 2001

The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1708-1714
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• 2019

The purpose of the critical evaluation of the spent fuel pool (SFP) is to verify that the maximum effective multiplication factor ($K_{eff}$) is less than the critical safety limit at 100% stored condition of the spent fuel with the maximum reactivity. At nuclear power plants, the storage standard of spent fuel, ie, the loading curve, is established to prevent criticality from being generated in SFP. Here, the loading curve refers to a graph showing the minimum discharged burnup versus the initial enrichment of spent fuel. Recently, US NRC proposed the new critical safety assessment guideline (DSS-ISG-2010-01, Revision 0) of PWR SFPs and most of utilities in US is following it. Of course, the licensed criterion of the maximum effective multiplication factor of SFP remains unchanged and it should be less than 0.95 from the 95% probability and the 95% confidence level. However, the new guideline is including the new evaluation methodologies like the application of the axial burnup profile, the validation of depletion and criticality code, and trend analysis. Among the new evaluation methodologies, the most important factor that affects $K_{eff}$ is the axial burnup profile of spent fuel. US NRC recommends to consider the axial burnup profiles presented in NUREG-6801 in criticality analysis. In this paper, criticality effect was evaluated considering three profiles, respectively: i) Axial burnup profiles presented in NUREG-6801. ii) Representative PWR axial burnup profile. iii) Uniform axial burnup profile. As the result, the case applying the axial burnup profiles presented in NUREG-6801 showed the highest $K_{eff}$ among three cases. Therefore, we need to introduce a new methodology because it can be issued if the axial burnup profiles presented in NUREG/CR-6801 are applied to the domestic nuclear power plants without any other consideration.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.336-336
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• 2002

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.1-5
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• 2017

The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor ($K_{eff}$) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of $D_2O$, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.