Validation of PWR-Fuel Code for Static Parameters in the LWR Core Benchmark

Kuntoro, IK Iman; Pinem, Surian; Sembiring, Tagor Malem

doi:10.17146/tdm.2018.20.3.4650

Cited by 3 publications

(3 citation statements)

References 16 publications

(17 reference statements)

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“…The need to verify and validate becomes an important issue, especially in specially developed computer programs. For this reason, many researchers perform code validation to ensure that their results are consistent with their case [1]- [3]. This study will verify a program package called SRAC code system developed by JAERI with a deterministic method to calculate core parameters [4].…”

Section: Introductionmentioning

confidence: 99%

Validation of Srac Code System for Neutronic Parameters Calculation of the PWR Mox/Uo2 Core Benchmark

Luthfi

Pinem

2021

urania

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VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK. Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choosed because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction, and Power Peaking Factor (PPF) and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly power error) value of power-weighted error (PWE) and error-weighted error (EWE) is 2.93% and 3.94%, respectively. Maximum difference between PPF value in axial direction with reference reaches a value of 4.57%. SRAC calculation results also show consistency with the calculation results of other program packages or code. Results of this study indicate that SRAC code system is still quite accurate for the calculation of neutronic parameters of PWR reactor core benchmark. Therefore, SRAC code system can be used to calculate neutronic parameters of PWR reactor core, especially when using MOX (mixed oxide) fuel.Keywords: Neutronic parameter, critical boron concentration, power peaking factor, SRAC code system.

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Section: Introductionmentioning

confidence: 99%

Validation of Srac Code System for Neutronic Parameters Calculation of the PWR Mox/Uo2 Core Benchmark

Luthfi

Pinem

2021

urania

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“…The location of the .. RSG-GAS reactor. The RSG-GAS reactor is a water-moderated pool-type reactor designed for radioisotope production and testing many types of objects in the steady-state modes of operation [3]. One of the distinguishing features of the RSG-GAS core is the large irradiation facility (4 cm x 8.1 cm x 7.71 cm) in central irradiation position.…”

Section: Introduction mentioning

confidence: 99%

“…BATAN's standard neutron diffusion code, Batan-3DIFF [3] code, has been developed in BATAN for neutronic design and safety analysis, especially design and analysis of research reactors. Those codes solve the three-dimensional multigroup neutron diffusion problems, respectively.…”

Section: Introduction mentioning

confidence: 99%

Calculation of Control Rods Reactivity Worth of RSG-GAS First Core Using Deterministic and Monte Carlo Methods

et al. 2019

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The control rod worth is a key parameter for the research reactor operation and utilization. Control rod worth computation is a challenge for the fully-deterministic and Monte Carlo calculations, including the few-group cross section generation, and the core analysis. The safe and reliable utilization of research reactor demands the possible accurate information of control rod worth because they are used to compensate the excess reactivity for safe reactor operation and its controlled shut down. The criticality positions of the control rods change with time due to buildup of fission products during the reactor operation. It is therefore important to determine the reactivity worth of control rods. The aim of this article is to obtain reliable control rod worth of the first core of RSG-GAS as a verification and validation result. For this purpose, deterministic and Monte Carlo models of the reactor core were developed and confirmed by the experimental results of excess reactivity, shutdown margin, and combined control rod reactivity worth using the combination of WIMSD-5B and Batan-3DIFF computer codes. WIMSD-5B is a neutron transport theory-based lattice cell modeling code that is used for the generation of group constants for different regions of the reactor core. These are provided as input to the diffusion theory based Batan-3DIFF code which performs the global core calculations for the reactor system. For the Monte Carlo model, to estimate the reactivity worth of control rods, the MCNP6 code is used. The result of this analysis showed that for the integral control rod worth a good agreement was found between experimental data and Monte Carlo simulation results but up to 5 % difference occurred between experimental results and diffusion result.

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