• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.483-488
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• 1998

A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development.

• Nuclear Engineering and Technology
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• 제15권2호
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• pp.98-109
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• 1983

2차원 다군 확산 이론에 의한 원자로 동특성 방정식의 해를 구하기 위해서 two-step alternating direction explicit method를 도입하였다. Alternating direction implicit method의 특별한 경우로써 이 방법의 정확도 및 안전성을 해석하였다. 이 방법의 타당성을 시험하기 위해서 TWIGL 전산조직에 사용한 implicit difference method와 비교하여 두 방법의 결과가 일치함을 알았다. 이 방법을 이용하여 가압경수형 원자로(PWR)의 제어봉 삽입시의 중성자 신속의 시간변화와, CANDU-PHW 원자로의 가상된 냉각재상실 사고시의 중성자 신속의 시간변화를 계산하여 이들 원자로의 제어능력을 확인하였다.

• Nuclear Engineering and Technology
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• 제16권4호
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• pp.202-216
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• 1984

Nodal method가 소격격자 해석방법의 하나로 정립됨으로써, 계산격자가 비교적 크더라도 각 격자의 평균출력분포를 정확히 계산할 수 있게 하는 균질화변수틀 찾는 방법이 중요하게 되었다. 본 연구에서는 simplified equivalence theory와 approximate node equivalence theory의 두가지 근사방법을 가압경수형 원자로 문제에 적응하여 시험하여 보았다. 균질화계산과 노심분석계산 방법으로서 analytic nodal method에 기초를 둔 ANM 코드를 개발하였다. 여러 균질화 방법외 정확성을 KTDD 코드에 의한 reference solution과 비교하여 본 결과, 균질화 계산은 핵연료영역에서는 영역별 핵연료집합체 계산으로, baffle과 reflector의 공존 격자영역은 이들을 포함하는 color set 계산으로 수행할 수 있음을 알았다. Approximate node equivalence theory에 입각해서 approximate homogenized cross-section들과 approximate discontinuity factor들의 균질화 변수를 사용하면 출력분포와 임계도가 각각 0.8%, 0,1% 오차 범위내에서 예측되었다.

• Nuclear Engineering and Technology
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• 제21권1호
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• pp.48-55
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• 1989

본 논고에서는 국내의 PWR 및 CANDU 사용후 핵연료로부터 발생하는 붕괴열의 장기적인 거동을 보다 손쉽게 분석하기 위하여 붕귀열을 추정할 수 있는 간단한 근사식을 도출하였다. 근사식의 장기적인 붕괴열 추정에서 ORIGEN 2코드 결과와의 차이를 줄이고 중요한 변수 조건하에서도 붕괴열을 추정할 수 있도록 하기 위하여 민감도 분석을 수행하였다. 그 결과로서 얻어진 근사식은 사용후 핵연료의 이력자료중 중요변수인 연도를 포함함으로써 3~500년정도의 냉각시간 범위내에서는 임의의 연소도를 가진 사용후 핵연료의 붕괴열이라도 추정할 수 있게 되었다. 그리고 대표적으로 30, 37 및 40 GWD/MTU등의 연소도를 갖는 사용후 핵연료의 붕괴열 추정에 있어서는 1년부터 $10^{5}$ 년까지의 냉각시간에 따라 ORIGEN2 ,코드의 결과와 $\pm$10%이내의 차이를 보이고 있어 사용후 핵연료 관리를 위한 관련시설의 열적설계 및 평가 등과 같은 공학적 목적에 유용하게 사용될 수 있을 것이다.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 춘계학술대회
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• pp.610-616
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• 2003

Recently, the common concern of nuclear power industry in the development of technology mitigating and preventing the aging of steam generator tubes prevails, because the trends of steam generator flaws at Uljin unit #1,2 and KSNP(Korea Standard Nuclear Power Plant) impose a burden on the operation of nuclear power plant. While the regulatory agency is demanding the establishment of the advanced general performance maintenance system, the steam generator management program adapting advanced technology is being developed which may comply with EPRI PWR SG Guidelines based on NEI 97-06 ‘ General Guidelines including all the maintenance aspects consist of the tube integrity assessment criteria, repair limit, allowable leakage level, water chemistry will be composed in order to obtain the approval of regulatory agency and be applied to Nuclear power plant early 2005. This presentation is to introduce maintenance state including SG tube degradation and main contents of advanced SG management program being developed, and futhermore update present and future plan, and estimate the alternation after the completion.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.148-153
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• 1996

In this paper, an effective method for resolving difficulty resulting from the heterogeneity of the PWR baffle/reflector region is developed on the basis of the AFEN method. The essential difference of the new method from the conventional approach based on the equivalence theory is that the heterogeneous baffle/reflector is directly, without homogenization, considered as a node in nodal calculation Numerical results show that AFEN method with the new method can accurately predict both the multiplication factor and the power distribution of thermal reactors with baffle explicitly modeled.

• Nuclear Engineering and Technology
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• 제19권4호
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• pp.292-309
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• 1987

미국 원자력 안전규제위원회(U.S. NRC)의 "원자로 위험도 참고문서(NUREG-1150)"에 입력자료로 제공하기 위해 실시한 Zion발전소 안전성 재평가 작입의 일환으로, 가상적 중대사고에 대한 대형건식 가압경수로 격납용기 반응해석을 수행하였다. 본 연구에서 사용한 방법론들은 Sandia 국립연구소에서 "중대사고 위험도 감소계획"의 일환으로 특히 Surry 발전소에 대한 연구를 위해 개발한 것이며, 이 방법론을 Zion발전소에 외삽법으로 적용하였다. 먹저, 원자력발전소의 위험도를 정량적으로 평가하는 주요절차를 개설하였다. 그리고, Zion발전소의 중대사고에 대한 격납용기 반응해석을 위해 사용한 방법론들을 상세히 기술하였다. 즉, 격납용기 반응해석을 위해 사용한 방법론들을 상세히 기술하였다. 즉, 격납용기 사건수목 해석 전산코드의 주요 특징과 격납용기 사건수목의 정량적 평가절차를 요약하여 높액다. 격납용기 반응해석에 있어서 중요한 발전소 고유의 특성과 본 연구의 불확실성 분석에 포함시킨 격납용기 하중과 성능에 관계되는 문제점들을 아울러 제시하였다. 끝으로, 가상적 증대사고에 대한 대형건식 가압경수로 격납용기의 반응에 대한 전망을 제공하기 위해서 결정적 및 통계학적 격납용기 사건수목 해설결과를 간단히 요약하여 제시하였다.설결과를 간단히 요약하여 제시하였다.

• Nuclear Engineering and Technology
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• 제48권6호
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• pp.1303-1314
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• 2016

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical $CO_2$ cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2002년도 춘계학술대회 논문집
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• pp.684-687
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• 2002

The authors have developed two devices which cuts the spend fuel rod to an optimal size and extracts fuel pellet from the pieces of cut fuel rods. These devices are so important to reduce radioactive wastes that some advanced countries developed their own methods and devices. The authors have benchmarked from these methods and devices. For spent fuel rod cutting, the tube cutting method has been chosen. some mechanical properties of the fuel tube and pellet has been carefully considered for an optimal cutting size. For fuel pellet extraction, a mechanically extracting method has been adopted. The existing chemical method have turned out to be inappropriate because it produced large amount of radioactive wastes, in spite of its high fuel recovery characteristics. The developed method has an advantage that it can be applied to other fuel rods that have different shapes and sizes. The two devices are set up and operated in the hot cell where people can not go in, so that the devices have been designed to be controlled remotely and modulated for easy maintenance. And the performance of the devices has been tested by using simulated fuel rod. From the experimental results, the devices are supposed to be useful for reducing radioactive wastes.