scholarly journals CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

2016 ◽  
Vol 2016 ◽  
pp. 1-12 ◽  
Author(s):  
Pengcheng Zhao ◽  
Kangli Shi ◽  
Shuzhou Li ◽  
Jingchao Feng ◽  
Hongli Chen
Keyword(s):  
Steady State ◽  
Fast Reactor ◽  
Cooling System ◽  
Natural Circulation ◽  
Reactor Core ◽  
Transient State ◽  
Fuel Pin ◽  
Gen Iv ◽  
Steady State Simulation ◽  
Accident Conditions

Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS) transient simulation at beginning of the reactor cycle (BOC) has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Emergency Decay Heat Removal in a GEN-IV Gas-Cooled Fast Reactor

2006 ◽  
Author(s):  
Lap Y. Cheng ◽  
Hans Ludewig ◽  
Jae Jo
Keyword(s):  
Fast Reactor ◽  
Cooling System ◽  
Natural Circulation ◽  
Heat Removal ◽  
Forced Flow ◽  
Decay Heat ◽  
Conversion Unit ◽  
Station Blackout ◽  
Combination Approach ◽  
Gen Iv

A series of transient analyses using the system code RELAP5-3d has been performed to confirm the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor. The accident sequence of interest is a station blackout simultaneous with a small break (10 sq.inch/0.645m2) in the reactor vessel. The analyses cover the three phases of decay heat removal in a depressurization accident: (1) forced flow cooling by the power conversion unit (PCU) coast down, (2) active forced flow cooling by a battery powered blower, and (3) passive cooling by natural circulation. The blower is part of an emergency cooling system (ECS) that by design is to sustain passive decay heat removal via natural circulation cooling 24 hours after shutdown. The RELAP5 model includes the helium-cooled reactor, the ECS (primary and secondary side), the PCU with all the rotating machinery (turbine and compressors) and the heat transfer components (recuperator, pre-cooler and inter-cooler), and the guard containment that surrounds the reactor and the PCU. The transient analysis has demonstrated the effectiveness of passive decay heat removal by natural circulation cooling when the guard containment pressure is maintained at or above 800kPa.

Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code

2021 ◽  
Author(s):  
Filip Osuský ◽  
Branislav Vrban ◽  
Stefan Cerba ◽  
Jakub Luley ◽  
Vladimir Necas
Keyword(s):  
Steady State ◽  
Fast Reactor ◽  
Transient State ◽  
Reactor Design ◽  
Temperature Dependent ◽  
Control Rod ◽  
Fuel Bundle ◽  
Fuel Pin ◽  
Radial Heat ◽  
Steady State Heat

Abstract The paper investigates the influence of the used thermal-hydraulic approximations on the coupled calculations of Gas-cooled Fast Reactor design (hereby GFR 2400). The NESTLE code is used as coupled simulation tool and solves the multigroup neutron diffusion equation by the finite difference method that is internally coupled with a thermal-hydraulic sub-channel code. The in-house developed TEMPIN code and the CFD code FLUENT (from ANSYS code system) are used to prepare the thermal-hydraulic data for the GFR 2400 calculations. The TEMPIN code solves the steady state heat balance equation with flowing coolant in triangular lattice cell together with temperature dependent thermal-hydraulic properties of the fuel, cladding and coolant. Based on the calculated fuel bundle temperature distributions by the TEMPIN code, the thermal-hydraulic material properties (approximations) suitable for the NESTLE coupled code are processed for the GFR 2400 design. The influence of the constant and radial heat generation term within the fuel pin is studied within the paper. The performance of the NESTLE code with thermal-hydraulic approximations processed by both (TEMPIN and FLUENT) methods are compared with the findings of the GoFastR project. Moreover, both the thermal-hydraulic approximations were compared for one steady state and one transient state, related to the rapid withdrawal of one control rod assembly from the core. Changes in thermal-hydraulic distributions are described and visualized in the paper.

Experimental Study of the Effect of Hydrogen Inflow on Passive Core Cooling System With Natural Circulation Flow

2020 ◽  
Author(s):  
Yasunori Yamamoto ◽  
Masayoshi Mori ◽  
Kosuke Ono ◽  
Tetsuya Takada
Keyword(s):  
Steady State ◽  
Cooling System ◽  
Natural Circulation ◽  
Operating Conditions ◽  
Quasi Steady State ◽  
Circulation Flow ◽  
Term Operation ◽  
Accident Conditions ◽  
Core Cooling

Abstract Isolation Condenser (IC) is one of the passive core cooling systems with natural circulation flow, which is effective for safety measures against station black out. Once core uncover occurs, hydrogen generated in the core affects operating condition of ICs. To use ICs as an important safety measure not only for transient conditions but also for accident conditions, robustness of ICs against hydrogen inflow must be understood well. In this study, experiments with high pressure steam were conducted using experimental setup simulating IC, where helium was injected to simulate hydrogen effects. When the pressure in an accumulator increased high enough, natural circulation flow generated in the experimental loop. After the long-term operation, the pressure and the natural circulation flow rate achieved nearly constant. The pressure at quasi-steady state increased with increasing the helium injection amount. The pressure difference in a section including outlet side of a vertical pipe was slightly increased when helium was injected which may have indicated that the helium accumulated in the section and caused increment of the pressure loss. The startup pressure of the IC simulator also increased when helium was injected, where the driving force by the water head difference also decreased. Though long-term operations were performed after helium injection, the effect of injected helium on operating conditions of the IC remained for quasi-steady state conditions.

Preliminary Analysis of Unprotected Loss of Heat Sink in Small Long Life Gas Cooled Fast Reactor

2015 ◽  
Vol 751 ◽  
pp. 268-272
Author(s):  
Su'ud Zaki ◽  
Nuri Trianti ◽  
Rosidah M. Indah
Keyword(s):  
Fast Reactor ◽  
Heat Sink ◽  
Channel Model ◽  
Cooling System ◽  
Hot Spot ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reducing Power ◽  
Secondary System ◽  
Heat Removal System

The failure of the secondary side in Gas Cooled Fast Reactor system, which may contain co-generation system, will cause loss of heat sink (LOHS) accident. In this study accident analysis of unprotected loss of heat sink due to the failure of the secondary cooling system has been investigated. The thermal hydraulic model include transient hot spot channel model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is also plays important role to limit the level of consequence during the accident. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate the failure of the secondary system by reducing power through reactivity feedback and remove the rest of heat through natural circulations based decay heat removal (RVACS system).

Steady-state simulation of a cooling system using turbo pascal

1994 ◽  
Vol 48 (3) ◽  
pp. 201-210 ◽  
Author(s):  
E.E. Halawa ◽  
C.J. Trowbridge ◽  
C.J. Marquand
Keyword(s):  
Steady State ◽  
Cooling System ◽  
Steady State Simulation

scholarly journals SBWR Model for Steady-State and Transient Analysis

2008 ◽  
Vol 2008 ◽  
pp. 1-18 ◽  
Author(s):  
Gilberto Espinosa-Paredes ◽  
Alejandro Nuñez-Carrera
Keyword(s):  
Steady State ◽  
Cooling System ◽  
Natural Circulation ◽  
Convective Heat Transfer Coefficient ◽  
Transient Behavior ◽  
Distributed Parameters ◽  
Plant Data ◽  
Main Components ◽  
Main Steam Line ◽  
Main Steam

This paper presents a model of a simplified boiling water reactor (SBWR) to analyze the steady-state and transient behavior. The SBWR model is based on approximations of lumped and distributed parameters to consider neutronics and natural circulation processes. The main components of the model are vessel dome, downcomer, lower plenum, core (channel and fuel), upper plenum, pressure, and level controls. Further consideration of the model is the natural circulation path in the internal circuit of the reactor, which governs the safety performance of the SBWR. To demonstrate the applicability of the model, the predictions were compared with plant data, manufacturer_s predictions, and RELAP5 under steady-state and transient conditions of a typical BWR. In steady-state conditions, the profiles of the main variables of the SBWR core such as superficial velocity, void fraction, temperatures, and convective heat transfer coefficient are presented and analyzed. The transient behavior of SBWR was analyzed during the closure of all main steam line isolation valves (MSIVs). Our results in this transient show that the cooling system due to natural circulation in the SBWR is around 70% of the rated core flow. According to the results shown here, one of the main conclusions of this work is that the simplified model could be very helpful in the licensing process.

Advanced Recycling Reactor Design Study

2009 ◽  
Author(s):  
Makoto Mito ◽  
Shigeru Kunishima ◽  
Kim O. Stein ◽  
Kazumi Ikeda ◽  
Horoyuki Sato
Keyword(s):  
Cooling System ◽  
Natural Circulation ◽  
Scale Up ◽  
Reactor Core ◽  
Development Effort ◽  
Design Study ◽  
Commercial Plant ◽  
Reactor Building ◽  
Set Up ◽  
Recycling Reactor

The Advanced Recycling Reactor (ARR) design study sponsored by DOE of USA has been conducted [1]. This paper presents the pre-conceptual design of the ARR that is a loop-typed sodium cooled reactor with MOX fuel. The International Nuclear Recycling Alliance (INRA) takes advantage of international experience and agreed to use Japan Sodium-cooled Fast Reactor (JSFR) [2] as reference for Funding Opportunity Announcement (FOA) studies [1]. Since the scale-up factor of two is acceptable increase from manufacturing and licensing points of view, INRA proposes 3 evolutions of the ARR; ARR1, a 500 MWe demonstration plant, online in 2025; ARR2, a 1,000 MWe commercial plant, online in 2035; ARR3, a 1,500 MWe full-scale commercial plant, online in 2050. Japan has conducted R&Ds for the JSFR incorporating thirteen technology enhancements expected to improve safety, enhance economics, and increase reactor reliability. The ARR design is based on such the technology enhancements that it can benefit from this development effort and the ARR3 can become cost competitive with the similar sized LWRs. Major features of key technology enhancements are the following: Decay heat can be removed by natural circulation to improve safety. The primary cooling system consists of two-loop system and the integrated IHX/Pump to improve economics. The steam generator with the straight double-walled tube is used to improve reliability. The reactor core of the ARR1 is 70 cm high. The conversion ratio of fissile is set up less than 0.6 and the amount of burned TRU is 45–51 kg/TWeh. The ARR1 consists of a reactor building (including reactor auxiliary facilities and electrical / control systems), a turbine building, and a reprocessing building. The dimensions of the overall reactor building will be 46.1 m (W) × 72.8 m (L) × 70.3 m (H), and the volume of the building will be approximately 180,000 m3.

URANUS: Korean Lead-Bismuth Cooled Small Modular Fast Reactor Activities

2011 ◽  
Author(s):  
Sungyeol Choi ◽  
Il Soon Hwang ◽  
Jae Hyun Cho ◽  
Chun Bo Shim
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Base Isolation ◽  
Cooling System ◽  
Natural Circulation ◽  
High Level Waste ◽  
Isolation System ◽  
Base Isolation System ◽  
Enriched Uranium ◽  
Seismic Base Isolation

Since 1994, Seoul National University (SNU) has developed an innovative future nuclear power based on LBE cooling advanced Partitioning and Transmutation (P&T) approach that leaves no high-level waste (HLW) behind with transmutation reactor named as Proliferation-resistant, Environment-friendly, Accident-tolerant, Continual, and Economical Reactor (PEACER). A small modular lead-bismuth cooled reactor has been designated as Ubiquitous, Robust, Accident-forgiving, Nonproliferating and Ultra-lasting Sustainer (URANUS-40) with a nominal electric power rating of 40 MW (100 MW thermal) that is well suited to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. URANUS-40 is a pool type fast reactor with and an array of heterogeneous hexagonal core, fueled by proven low-enriched uranium dioxide fuels. The primary cooling system is designed to be operated by natural circulation. 3D seismic base isolation system is introduced underneath the entire reactor building allowing an earthquake of 0.5g zero period acceleration (ZPA) for the Safe Shutdown Earthquake (SSE). Also, the proliferation risk can be effectively managed by capsulized core design and a long refueling period (25yr).

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