초록 ▼

○ LEU 기반 장주기 노심 개념 개발
- PGSFR의 장주기 가능성 평가
- 장주기 노심을 위한 핵연료 개념 도출
- 고연소도 장주기 SFR 노심을 위한 핵연료 및 반사체 개념 도출
- 핵분열가스 배출식 금속연료 개념 개발
○ LEU 및 금속화된 PWR 사용후연료기반 장주기 SFR 노심개념 도출
- 파이로프로세싱을 통해 금속화된 SNF 조성 분석 및 SNF 장전 노심 평가
○ 장주기 SFR 노심 기포반응도 저감연구
- 냉각재 온도계수 및 기포반응도 개선 방안 도출
○ 고속로용 몬테칼

○ LEU 기반 장주기 노심 개념 개발
- PGSFR의 장주기 가능성 평가
- 장주기 노심을 위한 핵연료 개념 도출
- 고연소도 장주기 SFR 노심을 위한 핵연료 및 반사체 개념 도출
- 핵분열가스 배출식 금속연료 개념 개발
○ LEU 및 금속화된 PWR 사용후연료기반 장주기 SFR 노심개념 도출
- 파이로프로세싱을 통해 금속화된 SNF 조성 분석 및 SNF 장전 노심 평가
○ 장주기 SFR 노심 기포반응도 저감연구
- 냉각재 온도계수 및 기포반응도 개선 방안 도출
○ 고속로용 몬테칼로-하이브리드 방법론 개발
- 하이브리드 방법론 개선 및 검증 평가
○ 장주기 SFR 노심 안전인자 평가
- 장주기 노심 주요 안전인자 평가
- 고유안전성 방안 연구
○ 혁신 SFR 요소기술 개발 계획수립
- 혁신 SFR 요소기술 도출 및 평가 및 기술개발 전략 수립
(출처:요약서 3p)

Abstract ▼

IV. Results and Discussion
○ Development of a small long-life SFR based on LEU fuel
- Viability assessment of a long-life SFR using existing PGSFR fuel design
• Depletion calculation performed using the core model loaded with maximum Uranium enrichment of existing PGSFR LEU fuel
• Mon

IV. Results and Discussion
○ Development of a small long-life SFR based on LEU fuel
- Viability assessment of a long-life SFR using existing PGSFR fuel design
• Depletion calculation performed using the core model loaded with maximum Uranium enrichment of existing PGSFR LEU fuel
• Monte Carlo code McCARD is used, 5,000,000 neutron history per each depletion step
• It has been figured out that the cycle length of existing PGSFR is shorter than 3 years even loaded with maximum Uranium enrichment so that it is not suitable for the long-life SFR concept.
○ Optimization of long-life core design
- Development of lead-alloy radial reflector
• Radial reflector region is replaced by fuel assembly to increase the fuel inventory, and B4C shield assembly is substituted to radial reflector. The number of incore fuel assemblies is increased from 112 to 190.
• Lead-alloy reflector, which shows outstanding performance in SFR, replaces existing HT9M, and one radial reflector layer is added.
• As a result of a depletion calculation with the Uranium enrichment that makes BOC excess reactivity to be 500pcm, the cycle length is evaluated to be less than 2 years. It has been shown that improvement of fuel design is essential for long-life operaton.
- Optimization of fuel design in long-life SFR core
• Since it would be improper to increase diameter of the core anymore, the method to increase the volume ratio and inventory of fuel without a modification of the diameter has been considered.
• Core height is increased to 150cm, and U-7Zr is used as metallic fuel instead of U-10Zr. Natural Uranium blanket is placed at 30cm lower and upper part of the core. As a result, the cycle length becomes 15 years, but it is still beyond the design goal. New fuel assembly design is needed.
• To increase the fuel ratio with the same fuel assembly size, the number of control rods is decreased while the size of them is increased. As a result of depletion calculation using the improved fuel assembly with 140cm core height, the design goal has been satisfied when lead-alloy reflector is used; 20-year cycle length, 100GWd/MTHM fuel burnup.
• The core design that satisfies the design goal is set as the reference core, and the core characteristics has been analyzed to compare the conversion ratio, delayed neutron fraction, CVR using existing HT9M reflector and lead-alloy reflector.
- Development of fission-gas-vented metallic fuel concept
• The fission-gas-vented metallic fuel concept has been analyzed based on the fuel and core model of KALIMER-600.
• Diffusion of Cs-137 from fuel rod to coolant region is approximated in 1-D scheme with conservative assumptions.
• Diffusion of Cs-137 from fuel rod to coolant region is conservatively analyzed by 2-D approximation, which is more realistic than 1-D approximation.
○ Deriving a long-life SFR core based on LEU and metallized PWR spent fuel
- Characterization of PWR spent fuels
• Composition and characteristics of PWR spent fuel with 45 GWd/MTHM burnup is analyzed
• Applicability of PWR spent fuel to SFR is evaluated
- Characterization of the metallized PWR spent fuel by pyro-processing
• Pyro-processing technology to convert PWR spent fuel to metallic fuel for SFR is studied
• Previous studies about pyro-processing are investigated and the removal of 15 isotopes is confirmed by analyzing each process
• PWR spent fuel composition is calculated with ORIGEN code assuming 50 GWd/MTHM burnup and 10-year cooling. Composition ratio of uranium and TRU in spent fuel based metallic fuel except for the rest fission products is around 95.6% and 1.04%, respectively.
- Analysis of SFR loaded with the PWR spent fuel
• Innovative SFR core configuration is derived, which can achieve the criticality with PWR spent fuel and LEU mixed fuel. Inner and outer core is loaded with 43SNF-50LEU-7ZR and 39SNF-54LEU-7ZR, respectively.
• Annular fuel concept is introduced to adjust the axial blanket height of inner core region and to reduce the axial growth of the nuclear fuel.
• Monte Carlo code McCARD with ENDF/B-VII.0 cross section library is used to simulate the reactor physics accurately with minimum approximation
• It is shown that reactor concept with flat axial blanket can exceed the design objective of longer cycle length than 20EFPY and higher discharge burnup than 100GWd/MTHM.
• Radial power distribution is evaluated and optimization requirement of LEU configuration is confirmed due to high power peaking, 1.279 and 1.339 at BOC and EOC, respectively.
• Core temperature analysis for BOC and EOC is performed using the subchannel code MATRA-LMR. It is confirmed that the maximum temperature of both fuel and cladding in hottest channel are lower than limiting design requirement.
• Uranium mass ratio in metallic fuel is considered to be increased from conventional 90% to 92~93% to reduce the uranium enrichment loaded in initial core of the innovatie SFR. Zirconium hydride coated fuel pin concept is derived to maintain the initial reactivity. It is shown by the depletion calculation result of Monte Carlo code SERPENT that same initial reactivity can be achieved without deteriorating the cycle length with 17.82% uranium enrichment.
○ Research for reduction of void reactivity in long-life SFR
- Derivation of ductless fuel assembly concepts
• Ductless fuel assembly concept is derived. The optimum gap between the peripheral fuel rods of fuel assemblies is evaluated to be 0.2 cm. By removing the duct, fuel assembly pitch and equivalent diameter of active core is reduced to 16.72 cm and 107.9 cm, respectively.
- Analysis of impacts of ductless fuel assembly on the neutronic parameters
• Ductless core concept is derived and reactor physics characteristics are evaluated using Monte Carlo code McCARD. Required uranium enrichment to achieve the initial reactivity is reduced to 10.5% for inner core region and 13.5% for outer core region, which is much lower than conventional core.
• It is shown by the depletion calculation of ductless core that both cycle length and conversion ratio are largely improved due to the absence of neutron capture by duct.
• Reactor physics characteristics of ductless core are analyzed and it is shown that coolant void reactivity is slightly increased.
- Derivation of CVR improvement method for long-life PGSFR
• 6 strategies to improve the coolant temperature coefficient are derived and applicability to innovate SFR is evaluated.
• Although the concept that coating ZrH1.6 inside the fuel rod is derived, design objectives in terms of cycle length and discharge burnup could not be achieved.
• Although inner blanket concept using Th-7Zr is derived, coolant void reactivity reduction at EOC is only about 50 pcm in case cycle length and discharge burnup design objectives are satisfied.
• Autonomous Reactivity Control (ARC) is adopted to innovative SFR and the degree of coolant void reactivity reduction is shown to 7$ to 8$ at EOC. However, optimization is needed due to the concern about (n,T) reaction of Li-6.
• By referring Gas Expansion Module (GEM) concept, Floating Absorber for Safety at Transient (FAST) module concept which is capable of ensuring the reduction of both coolant temperature coefficient and coolant void reactivity is derived.
• It is shown that the coolant void reactivity reduction impact of FAST module is not more efficient than ARC module, however, it can be improved by optimization research. Optimization research is performed in the part of long-life core inherent safety improvement research.
• Static Absorber Feedback Equipment (SAFE) is derived based on simple and passive principle which is thermal expansion and contraction of metal. As coolant temperature increases, negative reactivity feedback can be induced by inserting fixed absorber to the core passively. Noticeable improvement of coolant temperature coefficient is confirmed by testing initial module loading. Like FAST, optimization research is performed in the part of long-life core inherent safety improvement research.
• Finally, FAST & SAFE system is adopted as coolant void reactivity reduction method of long-life core. It is tested in LEU based and TRU based innovative SFR core and effective improvement of coolant temperature coefficient is confirmed.
○ Development of Monte Carlo-Deterministic Hybrid methodology in SFR analysis
- Research for optimized application of nodal equivalence theory
• To define DFs in a CA region, DFs in the fuel region is also generated in order to satisfy the equivalence theory perfectly. The optimized radius for DF generation model is determined as 25 cm.
• The equivalence theory are applied to a reflector region, but it shows unstable results with 􍾢300~200 pcm errors. it is shown that the application of equivalence theory to a reflector region can not improve the accuracy of the hybrid method in fast reactor analysis.
• CVR are estimated by using the hybrid method. It is shown that CVR estimation using the hybrid method needs the nodal equivalence theory in axial direction to consider axial neutron streaming. It is required to do further study on this issue.
- Verification of the hybrid methodology in SFR analysis
• The hybrid method is applied to PGSFR analysis. In the case of unrodded core with 24 energy group, the hybrid method shows a good agreement with 72 pcm error. To solve more difficult problem using the hybrid method, inner fuel assemblies are replaced with burned fuel assemblies in PGSFR. It is shown that the hybrid method has a good performance by showing ~3 pcm errors in this analysis.
○ Characterization of safety parameters of the long-life SFR
- Evaluation of major safety parameters
• BOC and EOC core of derived LEU based long-life innovative SFR(iSFR) are modelled by depletion calculation results of Monte Carlo code McCARD and major safety parameters are evaluated.
• Among the safety parameters, coolant temperature coefficient and coolant void reactivity are shown to be still strongly positive at EOC, which are inherent concerns at EOC of long-life core. Necessity of passive safety device development is identified.
• For more accurate evaluation, neutronics-TH coupled analysis for the innovative SFR is carried out and radial power and temperature peaking distribution are analyzed. By carrying out pin-wise power distribution analysis in fuel assembly with radial temperature peaking, it is confirmed that maximum power peaking is 1.255.
- Research to enhance the inherent safety of long-life small SFR
• Optimization research for the previously derived FAST and SAFE device is carried out to secure the inherent safety of developed LEU based small innovative SFR. By adopting optimized FAST, sodium void reactivity can be reduced 2~3$ at BOC and 4~5$ at EOC. In case of SAFE, 13~18pcm of negative reactivity is inserted if core temperature rises 100K. Optimization of derived passive safety devices contributed to improvement of core inherent safety.
• Time constant analysis based on balance of reactivity (BOR) is performed for more conservative safety evaluation. Possibility to improve the inherent safety of iSFR by applying FAST and SAFE is confirmed again.
• Possible secondary effects in system with FAST are analyzed. Because of operating characteristics of FAST, bending of the module can cause the problem. Basic analysis for linearly bent and quadratically bent scenarios are carried out. It is shown that currently suggested 1700mm long FAST can effectively prevent the clogging due to these 2 phenomena.
○ Establishment of development of essential technologies for innovative SFR plant
- Identification and prioritization of essential technologies
• For iSFR, 7 essential technologies are introduced; 1) Annular Metallic fuel concept for fission gas release, 2) FAST & SAFE for passive safety system of fast reactor, 3) Micro intermediate heat exchanger concept, 4) Decay heat removal technology with mercury, 5) 10MW super-critical CO₂ power conversion system, 6) Technology for removing reaction product between sodium and CO₂ 7) Material for iSFR and super-critical CO₂ power conversion system coupled system
• The order of priority of 7 element technologies are evaluated by essentiality, safety, strategic technique, economics, reliability, necessity of cooperation development and necessirty of mid & long term development
- Establishment of development plans for the elementary technologies
• The Annular metallic fuel concept is one of essential technologies of iSFR in terms of designed as fission gas is released selectively and designed as annular shape. Base on this, there is necessity to develop this annular metallic fuel concept autonomously.
• As passive safety systems, FAST & SAFE system is developed. FAST use the buoyancy force to insert absorber into the core. In that point, component material of FAST has longer life time than lifetime of iSFR. FAST location is studied when pin is bended and when inlet temperature is changed. SAFE use the thermal expansion of SS rod which contain absorber at the end of rod. In that point, practical operating temperature is considered to optimize SAFE. As SAFE use space near control rod assembly, the performance of control rod assembly should be confirmed.
• Based on PCHE method, it is available to minimize intermediate heat exchanger system. To optimize PCHE intermediate heat exchanger design, design tool is needed. Developing design code for intermediate heat exchanger of iSFR is possible by combining own PCHE design code, power cycle code and neutronic code.
• The passive safety system is developed by RVACS method with mercury thermosyphon. Additional analysis is needed in case of floating trap in sodium pool when outer vessel is cooled.
• Validity of manufacturing and demonstrating super-critical CO₂ conversion system is studied. Based on validation study, demonstrating super-critical CO₂ conversion system is available until 2022.
• For better understanding of reaction production between sodium and CO₂, many experiments are done in various condition. To develop the removal system of reaction production, more experiment is needed.
• Several material candidates are evaluated as essential material of iSFR combined the super-critical CO₂ conversion system. And creep analysis is also evaluated with material candidates.
(출처:SUMMARY 20~28p)

목차 Contents

• 표지 ... 1제 출 문 ... 2보고서 요약서 ... 3요 약 문 ... 4SUMMARY ... 16CONTENTS ... 30목차 ... 32제 1 장 연구개발과제의 개요 ... 34 제 1 절 연구개발의 필요성 ... 34 제 2 절 연구개발 범위 및 목표 ... 37제 2 장 국내외 기술개발현황 ... 38제 3 장 연구개발수행 내용 및 결과 ... 44 제 1 절 LEU 기반 고연소도 장주기 소형 SFR 노심 개념 개발 ... 44 1. PGSFR 기존 핵연료에 기초한 장주기 노심 가능성 평가 ... 44 2. 납합금 반경방향 반사체 개념 개발 ... 48 3. 고연소도 장주기 노심 핵연료 설계 최적화 ... 51 4. 핵분열가스 배출식 금속연료 개념 개발 ... 85 제 2 절 LEU와 PWR 사용후연료 기반 장주기 노심개념 도출 ... 91 1. 기존 PWR 사용후연료 조성특성 분석 ... 91 2. 파이로프로세싱을 이용한 금속화된 사용후연료 조성 평가 ... 93 3. 사용후연료 기반 장주기 SFR 노심개념 평가 및 성능 예비평가 ... 96 제 3 절 장주기 SFR 노심기포반응도 저감연구 ... 121 1. 덕트(Duct)를 제거한 핵연료집합체 개념 도출 ... 121 2. 핵연료집합체 덕트 제거가 노물리 특성에 미치는 영향 평가 ... 123 3. 장주기 SFR 노심의 냉각재 온도계수 및 기포반응도 개선 방안 도출 ... 126 제 4 절 고속로용 몬테칼로-결정론 하이브리드 방법론의 검증 및 평가 ... 160 1. 하이브리드 방법론 정확도 개선을 위한 노달균등이론 최적 적용 연구 ... 160 2. 고속로 해석 하이브리드 방법론의 검증 및 평가 ... 185 제 5 절 장주기 소형 SFR 노심 안전인자 특성평가 ... 189 1. 장주기 노심 성능 및 주요 안전인자 평가 ... 189 2. 고유안전성 강화 방안 연구 ... 195 제 6 절 혁신 SFR 요소기술 도출 및 개발 계획수립 ... 212 1. 혁신 SFR 요소기술 도출 및 우선순위 평가 ... 212 2. 요소기술 기술개발 전략 수립 ... 218제 4 장 목표달성도 및 관련분야에의 기여도 ... 224 제 1 절 과제 목표 달성도 ... 224 제 2 절 관련분야 기여도 ... 225제 5 장 연구개발결과의 활용계획 ... 226제 6 장 연구개발과정에서 수집한 해외과학기술정보 ... 227제 7 장 참고문헌 ... 228끝페이지 ... 231