• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1813-1824
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• 2022

This paper shows the verification work of SARAX code system in the reactor core transient calculation based on the simplified EBR-II Benchmark. The SARAX code system is an analysis package developed by Xi'an Jiaotong University and aims at the advanced reactor R&D. In this work, a neutron-photon coupled power calculation model and a spatial-dependent reactivity feedback model were introduced. To verify the models used in SARAX, the EBR-II SHRT-45R test was simplified to an ULOF transient with an input flowrate change curve by fitting from reference. With the neutron-photon coupled power calculation model, SARAX gave close results in both power fraction and peak power prediction to the reference results. The location of the hottest assembly from SARAX and reference are the same and the relative power deviation of the hottest assembly is 2.6%. As for transient analysis, compared with experimental results and other calculated results, SARAX presents coincident results both in trend and absolute value. The minimum value of core net reactivity during the transient agreed well with the reported results, which ranged from -0.3$ to -0.35$. The results verify the models in SARAX, which are correct and able to simulate the in-core transient with reliable accuracy.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4518-4526
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• 2023

NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutron-moderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4272-4279
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• 2022

In this study, a cross section stochastic sampling (S.S.) capability is implemented into both the McCARD continuous energy Monte Carlo code and MIG multiple-correlated data sampling code. The ENDF/B-VII.1 covariance data based 30 group cross section sets and the SCALE6 covariance data based 44 group cross section sets are sampled by the MIG code. Through various uncertainty quantification (UQ) benchmark calculations, the McCARD/MIG results are verified to be consistent with the McCARD stand-alone sensitivity/uncertainty (S/U) results and the XSUSA S.S. results. UQ analyses for Three Mile Island Unit 1, Peach Bottom Unit 2, and Kozloduy-6 fuel pin problems are conducted to provide the uncertainties of k_eff and microscopic and macroscopic cross sections by the McCARD/MIG code system. Moreover, the SNU S/U formulations for uncertainty propagation in a MC depletion analysis are validated through a comparison with the McCARD/MIG S.S. results for the UAM Exercise I-1b burnup benchmark. It is therefore concluded that the SNU formulation based on the S/U method has the capability to accurately estimate the uncertainty propagation in a MC depletion analysis.

• Nuclear Engineering and Technology
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• 제13권3호
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• pp.130-138
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• 1981

ETOT-3-ETOG-3 전산체계와 INDF/B-IV 평가핵자료를 이용하여 LEOPARD코드용 핵자료라이브러리를 생산하였다. 그리고 생산된 라이브러리의 신뢰성을 입증하기 위하여, 선정된 많은 실험자료에 대한 임계계산을 수행하였다. 이 결과를 토대로 경수형 $UO_2$핵연료계에 대한 수정, 평가계산을 수행하여 조정된 라이브러리가 유용함을 확인하였다.

• 정보처리학회논문지:컴퓨터 및 통신 시스템
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• 제2권2호
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• pp.59-66
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• 2013

본 논문에서는 임베디드 시스템에서 DSP를 위한 메모리 접근 변수의 저장 방법에 대한 최적화 ILP 알고리즘을 제안하였다. 본 논문은 0-1 ILP 공식을 이용하여 DSP 주소 생성 유닛의 메모리 변수 데이터 레이아웃을 최소화한다. 제약 조건을 기반으로 변수의 메모리 할당 여부를 식별하고, 변수가 지시하는 주소코드를 프로그램 포인터에 등록한다. 프로그램의 처리 순서가 프로그램 포인터에 선언되면, 해당 변수의 주소코드에 대한 자동증감 모드를 적용한다. 주소 레지스터에 대한 로드를 최소화하여 변수의 데이터 레이아웃을 최적화한다. 본 논문에서 제안한 알고리즘의 효율성을 입증하기 위하여 FICO Xpress-MP Modeling Tools을 이용하여 벤치마크에 적용하였다. 벤치마크 적용 결과, 기존의 선언적 주문 메모리 레이아웃보다 제안한 알고리즘을 적용한 최적의 메모리 레이아웃이 주소/수정 레지스터에 대한 로드 수를 감소시켰고, 주소코드의 접근을 줄임으로써, 프로그램의 실행 시간을 단축시켰다.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.343-352
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• 2014

Uncertainty and sensitivity analyses with respect to nuclear data are performed with depletion calculations for BWR and PWR fuel assemblies specified in the framework of the UAM-LWR Benchmark Phase II. For this, the GRS sampling based tool XSUSA is employed together with the TRITON depletion sequences from the SCALE 6.1 code system. Uncertainties for multiplication factors and nuclide inventories are determined, as well as the main contributors to these result uncertainties by calculating importance indicators. The corresponding neutron transport calculations are performed with the deterministic discrete-ordinates code NEWT. In addition, the Monte Carlo code KENO in multi-group mode is used to demonstrate a method with which the number of neutron histories per calculation run can be substantially reduced as compared to that in a calculation for the nominal case without uncertainties, while uncertainties and sensitivities are obtained with almost the same accuracy.

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1994년도 추계학술발표회 초록집
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• pp.141-145
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• 1994

A computer code for solving the 3-D time-dependent multigroup neutron diffusion equation by a coupled reactor kinetics method recently developed has been developed and for evaluating its applicability in CANDU transient analysis applied to a 3-D kinetics benchmark problem which reveals non-uniform loss of coolant accident followed by an asymmetric insertion of shutdown devices. The performance of the method and code has been compared with the CANDU design code, CERBERUS, employing a finite difference improved quasistatic method.

• Journal of Radiation Protection and Research
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• 제29권1호
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• pp.41-48
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• 2004

A safety assessment code, called SAGE (Safety Assessment Groundwater Evaluation), has been developed to describe post-closure radionuclide releases and potential radiological doses for low- and intermediate-level radioactive waste (LILW) disposal in an engineered vault facility in Korea. The conceptual model implemented in the code is focused on the release of radionuclide from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. The radionuclide transport equations are solved by spatially discretizing the disposal system into a series of compartments. Mass transfer between compartments is by diffusion/dispersion and advection. In all compartments, radionuclides ate decayed either as a single-member chain or as multi-member chains. The biosphere is represented as a set of steady-state, radionuclide-specific pathway dose conversion factors that are multiplied by the appropriate release rate from the far field for each pathway. The code has the capability to treat input parameters either deterministically or probabilistically. Parameter input is achieved through a user-friendly Graphical User Interface. An application is presented, which is compared against safety assessment results from the other computer codes, to benchmark the reliability of system-level conceptual modeling of the code.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.45-60
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• 2006

Presently, the VHTGR (Very High Temperature Gas-cooled Reactor) is considered the most attractive candidate for a GEN-IV reactor to produce hydrogen, which will be a key resource for future energy production. A new concept for a reactor cavity cooling system (RCCS), a critical safety feature in the VHTGR, is proposed in the present study. The proposed RCCS consists of passive water pool and active air cooling systems. These are employed to overcome the poor cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. In order to estimate the licensibility of the proposed design, its performance and integrity were tested experimentally with a reduced-scale mock-up facility, as well as with a separate-effect test facility (SET) for the 1/4 water pool of the RCCS-SNU to examine the heat transfer and pressure drop and code capability. This paper presents the test results for SET and validation of MARS-GCR, a system code for the safety analysis of a HTGR. In addition, CFX5.7, a computational fluid dynamics code, was also used for the code-to-code benchmark of MARS-GCR. From the present experimental and numerical studies, the efficacy of MARS-GCR in application to determining the optimal design of complicated systems such as a RCCS and evaluation of their feasibility has been validated.

• 한국멀티미디어학회논문지
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• 제9권9호
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• pp.1247-1256
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• 2006

가상 기계란 하드웨어로 이루어진 물리적 시스템과는 달리 소프트웨어로 제작되어 논리적인 시스템 구성을 갖는 개념적인 컴퓨터이다. 그러나 가상 기계는 실제 프로세서로 처리하는 것보다 실행 속도가 매우 느리기 때문에 실행되는 코드의 최적화가 매우 중요하다. 본 논문은 가상 기계 코드 최적화기의 실험대상으로 EVM(Embedded Virtual Machine)의 중간 코드인 SIL(Standard Intermediate Language)을 이용하였다. 현존하는 최적화 방법론에 관한 연구를 통하여 가상 기계 코드 특성을 고려한 최적화 방법론을 제시하고, 최적화된 코드를 생성하기 위한 코드 최적화기를 설계하고 구현하였다. 가상 기계 코드 최적화기는 주어진 패턴을 찾아서 패턴에 해당하는 부분을 최적화 코드로 바꾸어, 전체 코드의 크기를 줄이고 실행 속도의 개선효과를 가진다. 또한, 구현된 최적화기의 실험 결과를 도출하였다.