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Validation of Code Packages for Light Water Reactor Physics Analysis
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Validation of Code Packages for Light Water Reactor Physics Analysis

A special issue of Journal of Nuclear Engineering (ISSN 2673-4362).

Deadline for manuscript submissions: closed (31 May 2025) | Viewed by 4489

Special Issue Editor

Dr. Kang Seog Kim

E-Mail Website
Guest Editor
Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37830, USA
Interests: computational reactor physics; nuclear data; transport and diffusion theory; resonance self-shielding methods; modeling and simulation; in-core fuel management; multiphysics simulation
Special Issues, Collections and Topics in MDPI journals

Special Issue Information

Dear Colleagues,

Light water reactor (LWR) analysis can be performed using a typical 2-step code package, which includes a 2D lattice physics code and a 3D nodal diffusion code with thermal–hydraulic feedback capabilities. The 2D lattice physics code processes few-group assembly-homogenized cross sections, incorporating the functionalization of various reactor states. The 3D nodal code conducts 3D whole-core calculations for LWR multi-physics analysis, including neutronics, thermal-hydraulics and transmutation. In recent years, direct 3D whole-core multi-physics analysis has become more practical, providing more reliable simulation results and thus enhancing LWR technology.

Nuclear vendors validate their code packges by estimating the uncertainties in key nuclear parameters via comparisons between the simulated and measured data, which can be used in in-core fuel management. In recent years, various LWR-based small modular reactors (SMRs) have been developed. Evaluating the uncertainties of the design code packages for these SMRs is challenging due to the lack of available measured data, which may require different approacges to uncertainty analysis to be employed, for example, by integrating individual components of uncertainty, such as those arising from nuclear data and manufacturing.

This Special Issue of JNE will focus on the validation of 2-step and direct LWR analysis code packages by comparing simulation results with plant-measured data, and by performing uncertainty analysis based on nuclear data and manufacturing uncertainties. The scope of this Special Issue includes validations of the analysis code packages used for the key nuclear parameters of operating LWRs, LWR-based SMR, and advanced LWRs.

Dr. Kang Seog Kim
Guest Editor

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Keywords

  • light water reactor
  • validation
  • uncertainty
  • reactor physics

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Published Papers (4 papers)

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Research

16 pages, 4802 KiB  
Article
Validation of the New TLANESY Thermal–Hydraulic Code with Data from the QUENCH-01 Experiment
by Nahum Contreras-Pérez, Heriberto Sánchez-Mora, Sergio Quezada-García, Armando Miguel Gómez Torres and Ricardo Isaac Cázares Ramírez
J. Nucl. Eng. 2025, 6(3), 32; https://doi.org/10.3390/jne6030032 - 12 Aug 2025
Viewed by 297
Abstract
Hydrogen generation and the correct simulation of severe accidents have been of utmost importance since the Fukushima Dai-ichi accident. QUENCH experiments are quite useful for validating mathematical models implemented in system codes for early-phase severe accidents, where hydrogen generation, fuel rod temperature, and [...] Read more.
Hydrogen generation and the correct simulation of severe accidents have been of utmost importance since the Fukushima Dai-ichi accident. QUENCH experiments are quite useful for validating mathematical models implemented in system codes for early-phase severe accidents, where hydrogen generation, fuel rod temperature, and their deterioration during these conditions are of vital importance. This paper presents a new system code, TLANESY, designed for the simulation of thermal–hydraulic systems with two-phase flow (mainly water) and with application in the analysis of severe accidents during the early phase. The computational implementation consists of fast-running numerical methods and their validation with experimental data from the QUENCH-01 experiment. The results showed an error with respect to the total hydrogen generation of approximately 0.6%. A stand-alone sensitivity analysis was also performed with some parameters related to the cladding, where it was shown that variation in the thermal conductivity by 15% can alter the total hydrogen generation by up to 5%, indicating that impurities in this material can have a significant impact on this Figure of Merit. Full article
(This article belongs to the Special Issue Validation of Code Packages for Light Water Reactor Physics Analysis)
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14 pages, 2310 KiB  
Article
A High-Fidelity Model of the Peach Bottom 2 Turbine-Trip Benchmark Using VERA
by Nicholas Herring, Robert Salko and Mehdi Asgari
J. Nucl. Eng. 2025, 6(3), 28; https://doi.org/10.3390/jne6030028 - 4 Aug 2025
Viewed by 309
Abstract
This work presents a high-fidelity simulation of the Peach Bottom turbine trip (PBTT) benchmark using the Virtual Environment for Reactor Applications (VERA), a multiphysics reactor modeling tool developed by the U.S. Department of Energy’s Consortium for Advanced Simulation of Light Water Reactors energy [...] Read more.
This work presents a high-fidelity simulation of the Peach Bottom turbine trip (PBTT) benchmark using the Virtual Environment for Reactor Applications (VERA), a multiphysics reactor modeling tool developed by the U.S. Department of Energy’s Consortium for Advanced Simulation of Light Water Reactors energy innovation hub. The PBTT benchmark, based on a 1977 transient event at the end of cycle 2 in a General Electric Type-4 boiling water reactor (BWR), is a critical test case for validating core physics models with thermal feedback during rapid reactivity events. VERA was employed to perform end-to-end, pin-resolved simulations from conditions at the beginning of cycle 1 through the turbine-trip transient, incorporating detailed neutron transport, fuel depletion, and subchannel thermal hydraulics. The simulation reproduced key benchmark observables with high accuracy: the peak power excursion occurred at 0.75 s, matching the scram time and closely aligning with the benchmark average of 0.742 s; the simulated maximum power spike was approximately 7600 MW, which is within 3% of the benchmark average of 7400 MW; and void-collapse dynamics were consistent with benchmark expectations. Reactivity predictions during cycles 1 and 2 remained within 1500 pcm and 400 pcm of criticality, respectively. These results confirm VERA’s ability to model complex coupled neutronic and thermal hydraulic behavior in a BWR turbine-trip transient, which will support its use in future studies of modeling dryout, fuel performance, and uncertainty quantification for transients of this type. Full article
(This article belongs to the Special Issue Validation of Code Packages for Light Water Reactor Physics Analysis)
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14 pages, 4475 KiB  
Article
Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor
by Kang Seog Kim, Andrew Ward, Ugur Mertyurek, Mehdi Asgari and William Wieselquist
J. Nucl. Eng. 2024, 5(3), 260-273; https://doi.org/10.3390/jne5030018 - 1 Aug 2024
Cited by 1 | Viewed by 1598
Abstract
The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by [...] Read more.
The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data. Full article
(This article belongs to the Special Issue Validation of Code Packages for Light Water Reactor Physics Analysis)
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14 pages, 5726 KiB  
Article
Validation of the SCALE/Polaris−PARCS Code Procedure with the ENDF/B-VII.1 AMPX 56-Group Library: Pressurized Water Reactor
by Kang Seog Kim, Byoung-Kyu Jeon, Andrew Ward, Ugur Mertyurek, Matthew Jessee and William Wieselquist
J. Nucl. Eng. 2024, 5(3), 246-259; https://doi.org/10.3390/jne5030017 - 23 Jul 2024
Viewed by 1445
Abstract
This study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of [...] Read more.
This study was conducted to validate the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library for pressurized water reactor (PWR) analysis, by comparing simulated results with measured data for critical experiments and operating PWRs. Uncertainties of the SCALE/Polaris–PARCS code procedure for PWR analysis were evaluated in the validation for the PWR key nuclear parameters such as critical boron concentrations, reactivity, control bank work, temperature coefficients, and pin and assembly power peaking factors. Full article
(This article belongs to the Special Issue Validation of Code Packages for Light Water Reactor Physics Analysis)
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