• Nuclear Engineering and Technology
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• 제49권1호
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• pp.43-53
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• 2017

Increasing the operation cycle length can be an important goal in the fuel reload design of a nuclear reactor core. In this research paper, a new optimization approach, electromagnetism mechanism (EM), is applied to the fuel arrangement design of the Bushehr WWER-1000 core. For this purpose, a neutronic solver has been developed for calculating the required parameters during the reload cycle of the reactor. In this package, two modules have been linked, including PARCS v2.7 and WIMS-5B codes, integrated in a solver for using in the fuel arrangement optimization operation. The first results of the prepared package, along with the cycle for the original pattern of Bushehr WWER-1000, are compared and verified according to the Final Safety Analysis Report and then the results of exploited EM linked with Purdue Advanced Reactor Core Simulator (PARCS) and Winfrith Improved Multigroup Scheme (WIMS) codes are reported for the loading pattern optimization. Totally, the numerical results of our loading pattern optimization indicate the power of the EM for this problem and also show the effective improvement of desired parameters for the gained semi-optimized core pattern in comparison to the designer scheme.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.601-610
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• 2024

This paper gives an extensive analysis on the characterization of xenon isotopic ratios for various nuclear reactors and nuclear explosions through neutronic depletion codes. The results of the characterization can be used for discriminating the sources of the xenon isotopes' release among the nuclear explosions and nuclear reactors. The considered sources of the xenon radionuclides do not only include PWR, CANDU, and nuclear explosions using uranium and plutonium bombs, but also IRT-200 and 5MWe Yongbyon (MAGNOX reactor) research reactors operated in North Korea. A new data base (DB) on xenon isotopic activity ratios was produced using the results of the characterization, which can be used in discrimination of the sources of xenon isotopes. The results of the study show that 5MWe Yongbyon reactor has quite different characteristics in ¹³⁵Xe/¹³³Xe ratio from the PWRs and the nuclear reactors have different characteristics in ¹³⁵Xe/¹³³Xe ratios from the nuclear explosions.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.900-906
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• 2024

The Passively-Cooled Molten Salt Fast Reactor (PMFR) is one of the advanced design concepts of the Molten Salt Fast Reactor (MSFR) which utilizes a natural circulation for the primary loop and aims to attain a long-life operation without any means of fuel reprocessing. For an extended operation period, it is necessary to have enough fissile material, i.e., high excess reactivity, at the onset of operation. Since the PMFR is based on a fast neutron spectrum, direct implementation of a burnable absorber concept for the control of excess reactivity would be ineffective. Therefore, a localized moderator concept that encircles the active core has been envisioned for the PMFR which enables the effective utilization of a burnable absorber to achieve low reactivity swing and long-life operation. The modified PMFR design that incorporates a moderator and burnable absorber is presented, where depletion calculation is performed to estimate the reactor lifetime and reactivity swing to assess the feasibility of the proposed design. All the presented neutronic analysis has been conducted based on the Monte Carlo Serpent2 code with ENDF/B-VII.1 library.

• Advances in Energy Research
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• 제5권2호
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Nuclear Engineering and Technology
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• 제19권4호
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• pp.317-324
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• 1987

본 논문은 가압 경수형 원자로의 제어봉 이탈사고와 같이 공간 의존적 과도특성 해석에 필히 요구되는 가상적 사고 분석을 위한 핵적 동특성 코드의 개발을 위한 것이다. 본 논문에서는 1.5군 중성자 화산 방정식에 의거한 수정형 Borresen 모형을 핵적 동특성 모델로 잡고 이를 공간의존적 과도특성해석에 응용할 수 있도록 수식화 하여 고리 1호기 초기 노심의 가상적인 제어봉 이탈 사고해석에 응용했다. 본 사고 해석에 채택한 수정형 Borresen 모형에 대한 계산 정밀도의 검증을 위해 출력 분포 및 제어봉가등 계산결과를 고리 1호기 초기 노심의 노물리 실험자료와 비교했고 공간의존적 사고해석에 있어서 중시되는 핵적 동특성 방정식의 계산 효율성을 검토했다. 그리고 이 결과를 토대로 수정형 Borresen 모형이 제어봉 이탈사고, 증기관 파탄사고 등과 같은 공간의존적 사고해석에 유용하게 이용될 수 있다는 것을 보였다.

• Nuclear Engineering and Technology
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• 제24권1호
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• pp.30-39
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• 1992

본 연구에서는 붕산주입 및 회석운전동안에 노심의 붕산농도를 변화시키기 위한 보충수 유량을 예측하고 화학 및 체적제어계통을 포함한 원자로 냉각재계통내에 있는 각종 계통에서 붕산농도 거동분석을 위한 종합적 붕산주입 및 희석모델(INBAD)이 제안되었다. 이 모델은 기존의 노심코드와 새로 개발된 붕산주입 및 희석모델로 구성되어 있으며 붕산주입 및 희석모델은 단일 cell 모델 및 다중 cell모델을 이용하여 본 연구목적에 맞게 개발되었다. 또한, 본 모델에서는 보다 실제적인 붕산농도 거동분석을 위하여 가변적 가압기 가열기 출력 및 선택적인 보충수 운전형태 (직접주입 또는 간접주입)가 모사되었다. 이 모델의 유용성을 증명하기 위하여 영광 3,4호기 설계자료를 이용하여 각종 계통에서 직접주입 및 간접주입운전에 대한 붕산농도 거동분석을 수행하였고, 노심의 붕산농도에 대한 가압기 가열기의 영향을 검토하였다. 그 결과 본 모델은 붕산주입 및 희석운전시에 각종 계통에서 붕산농도 변화를 정확히 예측할 수 있음을 보여 주었다.

• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.75-83
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• 2017

In severe accident conditions of light water reactors, the loss of coolant may cause problems in integrity of zirconium fuel cladding. Under the condition of the loss of coolant, the zirconium fuel cladding can be exposed to high temperature steam and reacted with them by producing of hydrogen, which is caused by the failure in oxidation resistance of zirconium cladding materials during the loss of coolant accident scenarios. In order to avoid these problems, we develop a multi-metallic layered composite (MMLC) fuel cladding which compromises between the neutronic advantages of zirconium-based alloys and the accident-tolerance of non-zirconium-based metallic materials. Cold pilgering process is a common tube manufacturing process, which is complex material forming operation in highly non-steady state, where the materials undergo a long series of deformation resulting in both diameter and thickness reduction. During the cold pilgering process, MMLC claddings need to reduce the outside diameter and wall thickness. However, multi-layers of the tube are expected to occur different deformation processes because each layer has different mechanical properties. To improve the utilization of the pilgering process, 3-dimensional computational analyses have been made using a finite element modeling technique. We also analyze the dimensional change, strain and stress distribution at MMLC tube by considering the behavior of rolls such as stroke rate and feed rate.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1871-1885
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• 2019

The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic P_N scattering, and the multi-group cross-section generation model was performed on the k_eff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a k_eff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2578-2590
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• 2023

In this work, a 24-month two-batch fuel management strategy for the APR1400 using LEU + has been investigated, where enrichments of 5.9 and 5.2 w/o are utilized in lieu of the conventional 4-5 w/o UO₂ fuel. In addition, an Accident Tolerant Fuel (ATF) clad based on the swaging technology is applied to APR1400 fuel assemblies. In this special ATF clad design, both outer and inner SS316 layers protect the conventional zircaloy clad. Erbia (Er₂O₃) is introduced as a burnable absorber with two-fold goals to lower the critical boron concentration in the long-cycle LEU + loaded core as well as to handle the LEU + fuel in the existing front-end fuel facilities without renewing the license. Two types of fuel assemblies with different loading of gadolinia (Gd₂O₃) are considered to control both the reactivity and the core radial power distribution. The erbia burnable absorber is uniformly admixed with UO₂ in all fuel pins except for the gadolinia-bearing ones. In this study, two core designs were devised with different erbia loading, and core performance and safety parameters were evaluated for each case in comparison with a core design without any burnable absorbers. The core analysis was done using the two-step method. First, cross-sections are generated by the SERPENT 2 Monte Carlo code, and the 3-D neutronic analysis is performed with an in-house multi-physics nodal code KANT.

• Nuclear Engineering and Technology
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• 제44권5호
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• pp.551-560
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• 2012

The Wolsong-1 (W-1) Phase-B pre-simulations were carried out in preparation for tests to be conducted for the restart of the reactor after a major refurbishment project that included replacement of the pressure tube. These pre-simulations for Wolsong-1 Phase-B differ from those in the past that were performed for the Wolsong-1,2,3,4 tests in that these tests use the WIMS/DRAGON/RFSP-IST code suite for verification of the tests and gadolinium instead of the traditional PPV/MULTICELL/RFSP code system and boron as poison in the moderator system. The use of gadolinium is deemed not to have domestically accumulated experience gained from the previous Phase-B tests. Thus, it is appropriate to conduct a study in order to gain a correct understanding and interpretation of potential differences in test results stemming from using gadolinium rather than boron. Although the calibration of the reactivity device will not be noticeably different using boron and gadolinium at a constant moderator temperature, the temperature dependency of the neutronic behavior due to the presence of gadolinium in the moderator system might be pronounced. The results of the pre-simulations using gadolinium revealed that the moderator temperature reactivity coefficients indeed showed significant differences in comparison with those with boron. In order to secure the validity of the analysis results, the newly acquired WIMS/DRAGON/RFSP-IST code suite was verified against the W-2,3,4 Phase-B test results. The results of the new code suite verifications revealed some overall improvements in accuracy; justification of the use of the code can be claimed for the validation of the W-1 Phase-B test results.