Scram Tests Research Articles

Investigation of the fast flux test facility transient behavior during a loss of flow without scram test

A Coordinated Research Project (CRP) on the benchmark analysis of Fast Flux Test Facility (FFTF) Loss of Flow Without Scram (LOFWOS) has been held under the sponsorship of the International Atomic Energy Agency. The CRP aims to improve understanding of loss of flow events in fast reactors, as well as to assess capabilities of existing computer codes against experimental data. FFTF is a 400 MWth Sodium-cooled Fast Reactor (SFR) owned by the U.S. Department of Energy, operated from 1980 to 1992. The CRP involves the LOFWOS Test #13, belonging to a series of unprotected transients performed as part of the Passive Safety Testing program. In this framework, the Nuclear Engineering Research Group of Sapienza University of Rome has developed an integrated multiphysics modelling based on a coupled approach with RELAP5-3D© (thermal-hydraulics) and PHISICS (neutron-kinetics) codes. The present paper discusses the results obtained by the authors after the participation in the benchmark open phase. As suggested by the organizers, a two-step methodology was followed. Firstly, calculations were run by imposing the reactor power as boundary condition and focusing on the FFTF thermal–hydraulic behaviour. A good accordance was detected between the numerical results and the experimental data, above all for what concerns the reference core outlet temperatures. Then, the simulations were repeated by using a coupled approach. Calculation outcomes demonstrated the capability of the selected code suite (PHISICS/RELAP5-3D®) to reproduce the reactor transient behavior and its suitability to be used as an effective numerical tool to simulate liquid metal fast reactors accidental scenarios where thermal–hydraulic and neutron-kinetic phenomena are strongly coupled.

Analysis of loss of flow without scram test in the FFTF reactor – Part I: Preparation of neutronics data

This study presents a benchmark analysis of an unprotected loss of flow transient in a sodium-cooled fast reactor at the Fast Flux Test Facility (FFTF), carried out as part of an IAEA coordinated research project. Three codes, namely Serpent (Monte Carlo), DYN3D (3D nodal diffusion) and ATHLET (system thermal hydraulics), were employed in the benchmark exercise. Two distinct modeling approaches were utilized: 1) stand-alone ATHLET with point kinetics (PK) and system thermal hydraulics (TH); and 2) coupled DYN3D/ATHLET with spatial kinetics (SK) and system TH. Neutronics data essential for both approaches were generated using Serpent. The study is organized into three parts.Part I presents a summary of the preparation of neutronics data for PK or coupled SK/TH simulations and includes the outcomes of the static neutronics stage of the benchmark. The main focus lies on verifying the cross-section generation method for DYN3D by comparing its results against the reference Monte Carlo solutions obtained with Serpent.Part II will provide a detailed description of the ATHLET TH model of the system. This model will be thoroughly evaluated by comparing the results with the ANL benchmark solution and experimental data for the transient. Furthermore, a sensitivity analysis will be conducted to explore various modeling options and assess their impact on the simulation results.Part III will showcase the transient calculation results using the two modeling approaches. Additionally, an adaptive decay heat model for nodal codes will be introduced. The performance of both modeling approaches will be assessed by comparing their results to the available experimental data.

Validation of the fast reactor plant dynamics analysis code Super-COPD using FFTF loss of flow without scram test #13

The Japan Atomic Energy Agency (JAEA) has been developing a plant dynamics analysis code named Super-COPD for design studies and safety analyses of sodium-cooled fast reactors. Since there are a few validation studies of a thermal–hydraulic model in a plant, including a point kinetics model considering reactivity feedback for anticipated transient without scram (ATWS) events classified into the beyond design basis events (BDBEs), further validation studies under BDBE conditions are needed to improve the prediction accuracy of the model. Therefore, JAEA has participated in the international benchmark analysis of loss of flow without scram (LOFWOS) test #13 performed at the Fast Flux Test Facility (FFTF) as one of passive safety demonstration tests hosted by the International Atomic Energy Agency. In the first step, the transient calculation with major reactivity feedback mechanisms and imposing measured power history was conducted with Super-COPD. The results were compared with the measured data. Several challenges were identified about the thermal–hydraulic model in the core and heat transport systems and the reactivity model in the point kinetics model. Consequently, the transient calculation was performed with modified models considering the radial heat transfer, flow redistribution, and interwrapper flow in the core. The time-dependent change and the second peak of the outlet temperature in the specified assembly were predicted accurately because the core thermal–hydraulics was reproduced well. Furthermore, the assembly bowing reactivity model was developed to improve the net reactivity, thereby resulting in a good agreement with the measured data on the reactor power, temperature, and flow rate in the primary loop. Thus, this paper proves the validation of Super-COPD for the LOFWOS event in the BDBE.

Benchmark Analysis of FFTF Unprotected Loss of Flow without Scram Test No.13 with Fast Reactor Plant Dynamics Analysis Code Super-COPD

Benchmark Analysis of FFTF Unprotected Loss of Flow without Scram Test No.13 with Fast Reactor Plant Dynamics Analysis Code Super-COPD

Deceleration Performance and Parameter Influence Analysis of the Control Rod Hydraulic Deceleration Device for Control Rod Hydraulic Drive System

Control rod hydraulic drive system (CRHDS), which is invented by INET, Tsinghua University, is a new type of internal control rod drive technology. Control rod hydraulic deceleration device (CRHDD), which consists of the plug, the hydraulic deceleration cylinder, etc., is one of the main components of the CRHDS. The CRHDD performs the rod dropping deceleration function through the interworking of the plug and the deceleration cylinder, which is filled with water, and reduces the rod dropping peak acceleration and the impact force acting upon the control rod to prevent the control rod cruciform blade from being deformed or damaged. The working mechanism of the CRHDD is presented and analyzed. The theoretical model of the control rod dropping process, which is based upon force analysis of the control rod during scram process, three-dimensional flow field analysis, and flow resistance calculation of the hydraulic deceleration cylinder, the kinematics and dynamics analysis of the control rod, is built whose results are compared and validated by the CRHDS scram test results. Then the model is used to analyze the influence of the key parameters, including the fuel case gap, the plug design clearance, the working temperature, etc. on the CRHDD working performance. The research results can give guidance for the design and optimization of the CRHDD.

Physical startup tests calculations for Dukovany NPP using MOBY-DICK macrocode

Abstract Physical startup tests are an important part of the operation of each nuclear reactor fuel cycle. This paper is focused on calculations of these tests for Dukovany NPP using MOBY-DICK diffusion macrocode. In this paper two tests are presented, specifically SCRAM test and “Weight of the 6th group of control assemblies” test. Both tests are described, methodology of calculation is introduced and results of calculations are compared with experimental data.

Study on the decelerating mechanism of the new control rod hydraulic decelerator for the CRHDS based on numerical simulation and experiment

Control rod hydraulic drive system is a new type of internal control rod drive technology for the nuclear heating reactor. Control rod hydraulic decelerator is one of the major components of the CRHDS. The CRHD performs the rod drop speed control function to reduce the rod drop impact force and prevent the control rod cruciform blade from being deformed or damaged. The composition and working principle of the new CRHD are analyzed. The theoretical model of the rod drop dynamic process was developed and the flow field analysis of the CRHD flow channels was carried out to close the model. The full scale rod scram tests have been conducted to get the rod drop dynamic data. The theoretical results agree well with the experimental results, which verify its correctness. The rod drop theoretical model was then used to analyze the influence of the reactor coolant operating temperature and the hydraulic deceleration cylinder sidewall flow area on the dynamic characteristics of the new CRHD. The research results lay the base for the design and further optimization of the CRHD.

소듐냉각고속로 제어봉집합체의 낙하시간 및 충격속도 예측을 위한 CFD 해석

ABSTRACT In a pressurized water reactor (PWR), control rod assembly (CRA) falls into the guide tubes of a fuel assembly due to gravity for scram. Various theoretical approaches and numerical analyses have been performed because its shape is simple and its design was completely developed several decades ago. A control rod assembly for a sodium-cooled faster reactor (SFR) which is geometrically more complicated is being actively developed in Korea nowadays. Drop time and impact velocity of a CRA are important parameters with respect to reactivity insertion time and the mechanical robustness of a CRA and a guide duct. In this paper, computational method considering simultaneously the equation of motion for rigid body and the Navier-Stokes equations for fluid is suggested and verified by comparison with theoretical analysis results. Through this valuable CFD analysis method, drop time and impact velocity of initially designed SFR CRA are evaluated before performing scram tests with it. 1. 서 론

SAC-CFR computer code verification with Experimental Breeder Reactor II loss-of-primary-flow-without-scram tests data

A series of tests in the Experimental Breeder Reactor No.2 (EBR-II) have been conducted to investigate the effects of a complete loss of primary flow without scram (LOFWS). In this paper we use a computer code which we have developed, named System Analysis Code for China Fast Reactor (SAC-CFR), which has a three-dimensional hot pool model to analyze the whole-plant transient behavior of the Experimental Breeder Reactor II (EBR-II) under loss of primary flow without scram (LOFWS) tests. The three-dimensional hot pool model code coupled in SAC-CFR was used to analyze flow patterns of the flow filled in Experimental Breeder Reactor II. Among the LOFWS tests, SHRT 45 and SHRT 39 are the most typical accident tests. In this paper, particular emphasis was given to validation of models for predicting SHRT 45 and SHRT 39 tests by SAC-CFR. The predicted results agreed well with measurements which are taken from Experimental Breeder Reactor II loss of primary flow without scram tests data.

Safety analysis for the loss-of-flow and loss-of-heat sink without scram tests in EBR-II

This paper discusses the safety considerations and the analysis that was done to support the conduct of the Loss-of-Flow and Loss-of-Heat Sink Without Scram tests in EBR-II. The plant safety limits and the analysis codes and methods are presented. The impact of the tests and off-normal conditions on the fuel and plant structures is evaluated. Conclusions are that the test program had no impact on plant operations and the accumulated fuel damage was negligible.