• Title/Summary/Keyword: Sodium-cooled Fast Reactor (SFR)

Search Result 81, Processing Time 0.033 seconds

Investigation on Design Requirements of Feed Water Drain and Hydrogen Vent Systems for the Prototype Generation IV Sodium Cooled Fast Reactor (소듐냉각고속로 원형로 소듐-물 반응 압력완화계통의 급수배출 및 수소방출 설계 요건 연구)

  • Park, Sun Hee;Ye, Huee-Youl;Lee, Tae-Ho
    • Korean Chemical Engineering Research
    • /
    • v.55 no.2
    • /
    • pp.170-179
    • /
    • 2017
  • We investigated design requirements of feed water drain and hydrogen vent systems for the sodium-water reaction pressure relief system (SWRPRS) of the prototype generation IV sodium cooled fast reactor (PGSFR). We evaluated the areas of the gas vent pipe of the water dump tank and the length of the water drain pipe of the steam generator to rapid drain of the water steam inside the steam generator for the normal and refueling operations, respectively. We also calculated the diameter of the gas vent pipe of the sodium dump tank which met its design pressure.

FEASIBILITY OF AN INTEGRATED STEAM GENERATOR SYSTEM IN A SODIUM-COOLED FAST REACTOR SUBJECTED TO ELEVATED TEMPERATURE SERVICES

  • Koo, Gyeong-Hoi;Lee, Jae-Han
    • Nuclear Engineering and Technology
    • /
    • v.41 no.8
    • /
    • pp.1115-1126
    • /
    • 2009
  • As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

Preliminary Leak-before Break Assessment of Intermediate Heat Transport System Hot-Leg of a Prototype Generation IV Sodium-cooled Fast Reactor (소듐냉각고속로 원형로 중간열전달계통 고온배관의 파단전누설 예비평가)

  • Lee, Sa Yong;Kim, Nak Hyun;Koo, Gyeong Hoi;Kim, Sung Kyun;Kim, Yoon Jea
    • Transactions of the Korean Society of Pressure Vessels and Piping
    • /
    • v.12 no.1
    • /
    • pp.126-133
    • /
    • 2016
  • Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.

Feasibility Study on Ultrasonic Waveguide Sensor for Under-Sodium Viewing of Reactor Internals in Sodium-Cooled Fast Reactor (소듐냉각고속로 원자로 내부구조물의 소듐내부가시화를 위한 웨이브가이드 초음파센서의 적용 가능성 연구)

  • Joo, Young-Sang;Lim, Sa-Hoe;Park, Chang-Gyu;Lee, Jae-Han
    • Journal of the Korean Society for Nondestructive Testing
    • /
    • v.28 no.4
    • /
    • pp.364-371
    • /
    • 2008
  • Ultrasonic waveguide sensor has been developed for under-sodium viewing of reactor internal structures of a sodium-cooled fast reactor (SFR). The structure design concept of a waveguide sensor assembly was suggested and evaluated for the application in SFR. A 10 m long ultrasonic waveguide sensor assembly has been manufactured and the experimental feasibility tests were carried out. The 10 m long distance propagation performance of zero-order antisymmetric $A_0$ Lamb wave has been verified. The feasibility of ultrasonic waveguide sensor has been demonstrated by the C-scanning resolution performance test.

Dynamic data validation and reconciliation for improving the detection of sodium leakage in a sodium-cooled fast reactor

  • Sangjun Park;Jongin Yang;Jewhan Lee;Gyunyoung Heo
    • Nuclear Engineering and Technology
    • /
    • v.55 no.4
    • /
    • pp.1528-1539
    • /
    • 2023
  • Since the leakage of sodium in an SFR (sodium-cooled fast reactor) causes an explosion upon reaction with air and water, sodium leakages represent an important safety issue. In this study, a novel technique for improving the reliability of sodium leakage detection applying DDVR (dynamic data validation and reconciliation) is proposed and verified to resolve this technical issue. DDVR is an approach that aims to improve the accuracy of a target system in a dynamic state by minimizing random errors, such as from the uncertainty of instruments and the surrounding environment, and by eliminating gross errors, such as instrument failure, miscalibration, or aging, using the spatial redundancy of measurements in a physical model and the reliability information of the instruments. DDVR also makes it possible to estimate the state of unmeasured points. To validate this approach for supporting sodium leakage detection, this study applies experimental data from a sodium leakage detection experiment performed by the Korea Atomic Energy Research Institute. The validation results show that the reliability of sodium leakage detection is improved by cooperation between DDVR and hardware measurements. Based on these findings, technology integrating software and hardware approaches is suggested to improve the reliability of sodium leakage detection by presenting the expected true state of the system.

Structural Concept Design of KALIMER-600 Sodium Cooled Fast Reactor (소듐냉각 고속로 KALIMER-600 원자로 구조 개념설계)

  • Lee, Jae-Han;Park, Chang-Gyu;Kim, Jong-Bum;Koo, Gyeong-Hoi
    • Proceedings of the KSME Conference
    • /
    • 2007.05a
    • /
    • pp.285-290
    • /
    • 2007
  • KALIMER-600 is a sodium cooled fast reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

  • PDF

NUMERICAL ANALYSIS ON THE REACTOR CORE EXPANSION AND ENERGY BEHAVIORS DURING CDA USING UNDERWATER EXPLOSION THEORY (수중폭발 이론을 사용한 노심폭주사고 시 노심 팽창 및 에너지 거동 수치해석)

  • Kang, S.H.
    • Journal of computational fluids engineering
    • /
    • v.21 no.3
    • /
    • pp.8-14
    • /
    • 2016
  • A numerical analysis is conducted to estimate the core expansion and the energy behaviors induced by a core disruptive accident in a sodium-cooled fast reactor. The numerical formulation based on underwater explosion theory is carried out to simulate the core explosion inside the reactor vessel. The transient pressure, temperature and expansion of the core are examined by solving the equation of state and nonlinear governing equation of momentum conservation in one-dimensional spherical coordinates. The energy balance inside the computation domain is examined during the core expansion process. Heat transfer between the core and the sodium coolant, and the bubble rise during the expansion process are briefly investigated.

MICROSTRUCTURAL CHARACTERIZATION OF U-10WT.%ZR FUEL SLUGS CONTAINING RARE-EARTH ELEMENTS PREPARED BY MODIFIED INJECTION CASTING

  • SANG-HUN LEE;KI-HWAN KIM;SEOUNG-WOO KUK;JEONG-YONG PARK;JI-HOON CHOI
    • Archives of Metallurgy and Materials
    • /
    • v.64 no.3
    • /
    • pp.953-957
    • /
    • 2019
  • U-10wt.%Zr metallic fuel slugs containing rare-earth (RE: a rare-earth alloy comprising 53% Nd, 25% Ce, 16% Pr and 6% La) elements for a sodium-cooled fast reactor were fabricated by modified injection casting as an alternative method. The distribution, size and composition of the RE inclusions in the metallic fuel slugs were investigated according to the content of the RE inclusions. There were no observed casting defects, such as shrunk pipes, micro-shrinkage or hot tears formed during solidification, in the metallic fuel slugs fabricated by modified injection casting. Scanning electron micrographs and energy-dispersive X-ray spectroscopy (SEM-EDS) showed that the Zr and RE inclusions were uniformly distributed in the matrix and the composition of the RE inclusions was similar to that of a charged RE element. The content and the size of the RE inclusions increased slightly according to the charge content of the RE elements. RE inclusions in U-Zr alloys will have a positive effect on fuel performance due to their micro-size and high degree of distribution.

SIMMER-IV application to safety assessment of severe accident in a small SFR

  • H. Tagami;Y. Tobita
    • Nuclear Engineering and Technology
    • /
    • v.56 no.3
    • /
    • pp.873-879
    • /
    • 2024
  • A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

Compatibility Study between 316-series Stainless Steel and Sodium Coolant (316계 스테인리스강과 소듐 냉각재와의 양립성 연구)

  • Kim, Jung Hwan;Kim, Jong Man;Cha, Jae Eun;Kim, Sung Ho;Lee, Chan Bock
    • Korean Journal of Metals and Materials
    • /
    • v.48 no.5
    • /
    • pp.410-416
    • /
    • 2010
  • Studies were carried out to establish the technology for sodium-clad compatibility and to analyze the compatibility behavior of the Sodium-cooled Fast Reactor (SFR) cladding material under a flowing sodium environment. The natural circulation facility caused by the thermal convection of the liquid sodium was constructed and the 316-series stainless steels were exposed at $650{^{\circ}C}$ liquid sodium for 1458 hours. The weight change and related microstructural change were analyzed. The results showed that the quasi-dynamic facility represented by the natural convection exhibited similar results compared to the conventional dynamic facility. Selective leaching and local depletion of the chromium, re-distribution of the carbide, and the decarburization process took place in the 316-series stainless steel under a flowing sodium environment. This process decreased as the sodium flowed along the channel, which was caused by the change in the dissolved oxygen and carbon activity in the liquid sodium.