초록 ▼

• 경수로용 수치 원자로 검증
- 전노심 수송 계산 코드 nTRACER 및 몬테칼로 코드 McCARD를 이용하여 BEAVRS, VERA와 같은 상용 노심 기반 벤치마크 해석 및 AP1000 및 APR1400과 같은 신형노심에 대한 해석을 진행하고 그 계산결과를 비교함. 증배계수의 경우 대부분의 경우 ∼200 pcm 이내 핵연료집합체 단위 출력 분포는 ∼1% 이내에서 결과를 예측하여 우수성을 입증함.
- B&W, KRITZ 및 VENUS2와 같은 다양한 경수로 기반 실험노심에 대한 해석을 진행하여 그 결과를 실측치 및 코드

• 경수로용 수치 원자로 검증
- 전노심 수송 계산 코드 nTRACER 및 몬테칼로 코드 McCARD를 이용하여 BEAVRS, VERA와 같은 상용 노심 기반 벤치마크 해석 및 AP1000 및 APR1400과 같은 신형노심에 대한 해석을 진행하고 그 계산결과를 비교함. 증배계수의 경우 대부분의 경우 ∼200 pcm 이내 핵연료집합체 단위 출력 분포는 ∼1% 이내에서 결과를 예측하여 우수성을 입증함.
- B&W, KRITZ 및 VENUS2와 같은 다양한 경수로 기반 실험노심에 대한 해석을 진행하여 그 결과를 실측치 및 코드간 비교를 진행하였고 해당단계에서 필요한 추가적인 기능을 구현하여 코드의 확장성을 도모함.
- C5G7-TD 및 SPERT III E-Core와 같은 과도 계산을 수행하여 결과를 생산, 타 코드와 비교하여 그 우수성을 입증함. 추후 원자로 설계에 필수적인 과도 계산의 패러다임의 변화를 주도함.
• nTRACER 정확도 및 기능 향상
- nTRACER에 사용하는 독자적인 핵자료집 처리의 정확성 향상을 위하여 담금질 기법을 이용하여 서브그룹 인자의 최적화 및 SPH를 이용한 공명 다군 반응 단면적의 각도 의존성 처리를 최초로 개발함.
- 계산 속도 향상을 위하여 선추적 모듈의 개선을 통하여 비등방성 산란을 고려하는 계산의 속도가 약 90%정도 향상됨. 또한 선추적 계산의 알고리즘 변화 및 기존의 SP3 SENM 대신 선추적방법론을 도입하여 수렴성 향상을 통해 코드의 적용성 확장을 도모함.
• McCARD 기능 확장
- 일정한 온도에서의 계산이 아닌 계산 중 온도 의존법 처리 및 부수로코드 MATRA와의 연계를 통해 실제 동력로 해석의 기반을 마련함.

(출처 : 보고서 요약서 5p)

Abstract ▼

● Systematic Validation of Static Reactor Simulations
- With the direct whole core calculation code nTRACER and the Monte Carlo code McCARD, benchmarks based on operating commercial pressurized water reactors (PWR) were analyzed. It was confirmed that accurate core characteristics could be predic

● Systematic Validation of Static Reactor Simulations
- With the direct whole core calculation code nTRACER and the Monte Carlo code McCARD, benchmarks based on operating commercial pressurized water reactors (PWR) were analyzed. It was confirmed that accurate core characteristics could be predicted with ∼20 ppm critical boron concentration and ∼2 % assembly power against measured data or other codes such as MPACT, KENO-VI, and SHIFT of the CASL team. In addition, AP1000 designed by Westinghouse electric (WH) with innovative design was analyzed to confirm the analysis capability of nTRACER in handling complicated geometries of the advanced reactors.
- As a systematic benchmark for static neutronic characteristics assessment, a realistic depletion benchmark was designed spanning from fuel pins to assemblies. At each step, depletion methodologies were compared and sensitivity studies were performed. Also realistic calculation conditions were induced.
- Several critical benchmarks based on PWR characteristics such as B&W 1484, B&W 1810, KRITZ and VENUS were analyzed using nTRACER and McCARD. In general, those codes provide reasonable results but a little discrepancies would be resolved by further research.
- Since the BEAVRS, VERA and AP1000 have been using WH type assemblies, new benchmark based on Combustion engineering (CE) type fuel assembly was proposed and designed using the APR1400 reactor. With selected reactor designs, preliminary calculations were performed to compare the calculated results with the detector measurements. Unfortunately, due to security problems, detailed measurements could not be provided by the utility company at this moment of report preparation.

● Enhancement of Transient Reactor Simulation Capability
- The C5G7-TD phase1 benchmark was recently proposed for assessing the characteristics of transient simulation for the cases with with explicit pin-cell geometries. This on-going benchmark was successfully analyzed by nTRACER and McCARD. By comparing the solutions with the other codes’ results, the superior accuracy of the direct whole core calculation method was demonstrated successfully. And the steady states of the SPERT III E-Core were analyzed for future analyses using nTRACER.
- To improve transient calculation capabilities of nTRACER, backward difference formula (BDF) and its exponential transformation were implemented and compared with Crank-Nicolson method and its exponential transformation. With slow transient problem such as C5G7-TD, there are no big differences among methods. Also additional features such as the improvement of the assumption related to isotropic angular flux and the adjoint flux calculation module were implemented.
- To treat thermal/hydraulic phenomena properly, the subchannel analysis code COBRA-TF was combined with nTRACER. However, due to the inefficiency of COBRA-TF, a new subchannel analysis code ESCOT based on the dirft-flux model was developed. Also an coupled calculation control algorithm based on residual monitoring was developed for efficient neutronics-thermal/hydraulics coupled calculations.

● Methodology and Code Enhancement of Deterministic Core Simulators
- To improve the accuracy of the independent nuclear data library for nTRACER, various methods were examined and implemented. The simulated annealing method that improves subgroup levels and weights and the parametrized spectral superhomogenization (SPH) factor method that resolves the isotropic assumption employed in the group condensation process were developed. Also the macro level grip scheme was applied to improve the execution performance of the subgroup resonance treatment.
- In order to resolve the inefficiency of the PN ray tracing calculation in nTRACER, the angular flux save scheme (AFSS) which stores region averaged angular fluxes in addition to the region averaged scalar fluxes was implemented. This new scheme shows about 90% calculation speed improvements in the case of P2 calculation. Furthermore, the linear source method based on CASMO5 was applied and compared with the original method in nTRACER.
- To treat thermal/hydraulic feedback precisely, a gamma transport calculation modules were developed in nTRACER. In this process, an independent gamma cross section library was generated by using NJOY. This additional feature gives significant impacts on the low power pin-cell such as fuels mixed with gadolinium.
- The Jacobi sweep ray tracing was implemented to make the parallel calculation faster than the original calculation scheme (Gauss-Seidel). Additionally, as an effort to mitigate the inherent instability of the direct whole core calculation method, one dimensional MOC method was examined as the axial solver instead of SP3 SENM nodal method.
- A multi-group Monte Carlo solver with anisotropic scattering (PN)　was applied in nTRACER to enhance its calculation capability. To consider the anisotropic scattering, discretized angle sampling and non-uniform step function sampling were implemented.

● Monte Carlo Capability for Power Reactor Application
- In order to determine the temperature dependent cross sections on-the-fly in the McCARD Monte Carlo calculations, various on-the-fly temperature feedback modules including the Gauss Hermite quadrature method and the windowed multipole method were implemented. Also the subchannel analyses code MATRA developed by Korea Atomic Energy Research Institute (KAERI) was coupled with McCARD.
- To accelerate inactive cycles in Monte Carlo calculations, the coarse mesh finite difference (CMFD) method was applied. This scheme gives excellent performances similarly to the other deterministic acceleration scheme with CMFD method. Also to reduce the variance of Monte Carlo calculations, weight window based on adjoint flux algorithm was proposed and implemented.

● Demonstration of High Fidelity Simulation with Small Modular Reactor Analyses
- With nTRACER as a group constant generation code, the S-70 small modular reactor was designed by using the nTRACER/RENUS conventional two step calculation procedure. Furthermore, nTRACER standalone calculations were performed to generate direct whole core solutions for comparison between two methods. The results show that there are large discrepancies of about ∼360 pcm and the difference in axial power distribution of 2 % which might results from the heterogeneity of the S-70 core where neutron leakage and T/H behavior affect core characteristics significantly. Therefore, it was shown that a direct whole core calculation code such as nTRACER has critical advantages for analyzing small modular reactors.

(출처 : SUMMARY 13p)

목차 Contents

• 표지 ... 1
• 제 출 문 ... 3
• 보고서 요약서 ... 5
• 요 약 문 ... 7
• SUMMARY ... 13
• CONTENTS ... 17
• 목차 ... 19
• 그림목차 ... 21
• 표목차 ... 33
• 제1장 연구개발 과제의 개요 ... 39
• 제1절 연구개발의 목적 및 필요성 ... 39
• 제2절 연구개발 목표 및 내용 ... 41
• 1. 제1차년도 연구개발 목표 및 내용 ... 42
• 2. 제2차년도 연구개발 목표 및 내용 ... 42
• 3. 제3차년도 연구개발 목표 및 내용 ... 43
• 제2장 국내외 기술개발 현황 ... 44
• 제1절 고신뢰도 노심 모의를 위한 국내외 연구 현황 ... 44
• 제2절 수냉각동력로용 원자로 모의기 개발이력과 현황 ... 45
• 1. 실용적 수치원자로의 특장점 ... 46
• 2. 실용적 수치원자로의 문제점 및 개선 방안 ... 47
• 제3장 연구개발 수행내용 및 결과 ... 49
• 제1절 정상상태 노심 계산능 검증 ... 49
• 1. 정밀 노심 벤치마크 해석 ... 50
• 2. 실제적 연소계산 벤치마크 개발 ... 98
• 3. 임계실험 해석 ... 139
• 4. OPR1000 노심 벤치마크 확보 및 가동 노심 추적 계산 ... 187
• 제2절 과도계산능 개선 및 검증 ... 203
• 1. 과도 수송 계산능 검증 벤치마크 구성 및 해석 ... 203
• 2. 과도수송계산 실행성능 향상 ... 225
• 3. COBRA-TF 열유동 코드와 연계, 효율화 구현, 특장 평가 ... 233
• 제3절 노심모의 핵심 계산모듈 기능 확대 ... 267
• 1. 에너지군 구조 소격화, 연소사슬 단순화, 공명간섭 처리능 개선 ... 267
• 2. 선추적 계산 최적화, 선형 선원법 개선 ... 311
• 3. 감마선 수송 처리능 구현 ... 326
• 4. 다주기 연소노심 중간 시발 계산능 구현 ... 366
• 5. 계산 알고리즘 및 방법론 최적화 ... 373
• 6. 다군 몬테칼로 계산 비등방성 산란 처리 기법 구현 ... 395
• 제4절 동력로 몬테칼로 계산능 개발 지원 ... 405
• 1. 몬테칼로 계산 중 반응단면적 온도 의존성 처리법 평가 ... 405
• 2. CMFD기반 대량 입자 사용법 구현 ... 439
• 3. 노심전반 분산감소기법 개발 ... 459
• 4. MATRA 열유동 계산 연계 몬테칼로 반복계산기법 효율화 ... 462
• 제5절 소형 모듈화 원전 적용성 평가 ... 477
• 1. 대상 SMR 선정 및 2단계 노심 해석 ... 477
• 2. SMR 대상 직접 전노심 해석 및 기존방식 대비 특장 분석 ... 482
• 제4장 목표달성도 및 관련분야에의 기여도 ... 485
• 제5장 연구개발결과의 활용계획 ... 493
• 제6장 연구개발과정에서 수집한 해외과학기술정보 ... 495
• 제7장 참고문헌 ... 497
• 끝페이지 ... 502