• Nuclear Engineering and Technology
• /
• 제53권7호
• /
• pp.2104-2125
• /
• 2021

This paper presents the verification and validation (V&V) of a calculation module for isotope inventory prediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presented herein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion code RAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three history indices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, this method can directly consider three-dimensional core simulation conditions using history indices of the operating conditions. Therefore, this method reduces the computation time by avoiding a recalculation of the fuel depletion. The module for isotope inventory calculates the number densities using the Lagrange interpolation method and power history correction factors, which are applied to correct the effects of the decay and fission products generated at different power levels. To assess the reliability of the developed code system for back-end cycle analysis, validation study was performed with 58 measured samples of pressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF, HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can provide reasonable results with comparable confidence intervals. As a result, this paper successfully demonstrates that the isotope inventory prediction code system can be used for spent nuclear fuel analysis.

• International Journal of Internet, Broadcasting and Communication
• /
• 제15권2호
• /
• pp.218-226
• /
• 2023

In Korea, we still use function point based cost estimations for software size and cost of a project. The current problem is that we make difficultly calculating function points with requirements and also have less accurate. That is, it is difficult for non-experts to analyze requirements and calculate function point values with them, and even experts often derive different function points. In addition, all stakeholders strongly make the validity and accuracy of the function point values of the project before /after the development is completed. There are methods for performing function point analysis using source code [1][2][3][4] and some researchers [5][6][7] attempt empirical verification of function points about the estimated cost. There is no research on automatic cost validation with source code after the final development is completed. In this paper, we propose automatically how to calculate Function Points based on natural language requirements before development and prove FP calculation based on the final source code after development. We expect validation by comparing the function scores calculated by forward engineering and reverse engineering methods.

• Nuclear Engineering and Technology
• /
• 제52권9호
• /
• pp.1907-1916
• /
• 2020

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

• Nuclear Engineering and Technology
• /
• 제51권2호
• /
• pp.356-368
• /
• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Nuclear Engineering and Technology
• /
• 제53권12호
• /
• pp.3966-3978
• /
• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• 한국시뮬레이션학회논문지
• /
• 제27권1호
• /
• pp.25-32
• /
• 2018

본 논문은 모델 기반 내장형 소프트웨어의 자동 생성 코드에 대한 효율적인 신뢰성 시험 절차와 구체화된 동적 시험 방안에 대해서 제시하고 있다. 모델 정적/동적 시험 각각을 코드 정적/동적 시험 전에 수행함으로서 코드 신뢰성 시험 수행의 이점이 있음을 기술하였다. 또한, 모델과 코드의 신뢰성 시험 상관관계를 모델의 경우 Model Advisor와 Verification and Validation tool, 코드의 경우 Polyspace와 LDRA를 이용하여 살펴보고 제시한 절차대로 수행한 신뢰성 시험의 결과를 보여주고 있다.

• 대한기계학회논문집B
• /
• 제27권4호
• /
• pp.534-541
• /
• 2003

To investigate the combustion characteristics in a low-pollutant municipal waste incinerator, the generalized multi-block simulation code that can be applied to turbulent reacting flow with gaseous hydrocarbon fuel in a 3D complex geometry has been developed with nongray radiation effects. To deal with the complex geometry, structured multi-block method and the scheme which treats interfaces implicitly are adopted. The developed code is validated through various engineering problems such as curved duct flow, driven cavity flow, gray multi-block radiation, nongray radiation. and combustion in a incinerator.

• Nuclear Engineering and Technology
• /
• 제52권5호
• /
• pp.878-888
• /
• 2020

This paper presents the validation of UNIST in-house Monte Carlo code MCS used for the high-fidelity simulation of commercial pressurized water reactors (PWRs). Its focus is on the accurate, spatially detailed neutronic analyses of startup physics tests for the initial core of the Watts Bar Nuclear 1 reactor, which is a vital step in evaluating core phenomena in an operating nuclear power reactor. The MCS solutions for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) core physics benchmark progression problems 1 to 5 were verified with KENO-VI and Serpent 2 solutions for geometries ranging from a single-pin cell to a full core. MCS was also validated by comparing with results of reactor zero-power physics tests in a full-core simulation. MCS exhibits an excellent consistency against the measured data with a bias of ±3 pcm at the initial criticality whole-core problem. Furthermore, MCS solutions for rod worth are consistent with measured data, and reasonable agreement is obtained for the isothermal temperature coefficient and soluble boron worth. This favorable comparison with measured parameters exhibited by MCS continues to broaden its validation basis. These results provide confidence in MCS's capability in high-fidelity calculations for practical PWR cores.

• Nuclear Engineering and Technology
• /
• 제43권1호
• /
• pp.45-62
• /
• 2011

The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or nonstructured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.

• Nuclear Engineering and Technology
• /
• 제21권3호
• /
• pp.205-215
• /
• 1989

본 연구의 기초로 사용된 CET와 CET 정량화코드(EVNTREISS)는 WASH-1400 이후로 기 존 발전소 안전성평가에 적용되어온 어떠한 CET 보다도 상세하며 다양성을 지니고 있다. EVNTREISS 코드의 구성이 대단히 복잡하고, 코드 Validation이 철저히 수행되지 않았으며 또한 독립적 검토가 이루어지지 않음으로 인하여 코드와 CET의 한계점과 개선할 몇가지 문제점이 충분히 내포될 수 있었다. 본 연구는 BNL에서 수행된 Zion 격납용기 사건수목을 재구성하는 과정에서, 상기된 문제점과 개선분야를 파악하고 이전에 수행된 여러 PRA에서 사용된 CET와의 비교를 통하여 EVNTREISS코드를 개선시키고자 하였다. 결과로서, Zion에서 수행된 CET가 만족스럽게 재구성되었고, 코드에 대한 여러 개선분야가 파악되었으며 그중의 몇가지 문제점이 해결되었다. 이 개선된 코드는 PC에서 간단히 작업할 수 있으며 중대사 고시 대형건식 가압경수로형 격납용기반응의 분석에 효과적으로 사용될 수 있다.