Assessment of the MARS code for design of water pool-type reactor cavity cooling system in HTGR

This paper reports an experimental and numerical study on the assessment of the MARS code as a tool for analyzing the water pool-type reactor cavity cooling system (RCCS), which was developed by Seoul National University (SNU). A series of experiments were performed to determine the heat removal capability of the proposed RCCS and assess the capability of MARS code to predict the forced convective, natural convective and radiative heat transfer under normal operation conditions and boiling heat transfer during accident conditions in the RCCS. In the loss of forced convection (LOFC) accident experiment performed at the integral effect test facility called RCCS-SNU, the MARS code underestimated the vapor generation rate at the inner wall of the water pool. Therefore, the newly developed models of the bubble departure and lift-off diameters were implemented into the MARS code to make a better prediction of the vapor generation rate. The improved MARS code was assessed again using the experimental data of the LOFC accident conditions in the RCCS-SNU facility.     

辐照条件对光核反应产额的影响

利用２００ＭｅＶ电子束经不同厚度的高纯钨片转换得到的轫致辐射束诱发＾１９Ｆ（γ,ｎ）＾１８Ｆ和＾１２Ｃ（γ,ｎ）＾１１Ｃ反应,通过测定＾１８Ｆ和＾１１Ｃ的产额得到了转换靶钨片的最佳厚度,并与计算值进行了比较。测定了样品与转换靶的间距与产额的关系,推算了利用光子活化的方法分析Ｆ和Ｃ两种元素的探测下限分别为２．０ｎｇ和３．６ｎｇ。     

Development of Systematic Models for Aerosol Agglomeration and Spray Removal under Severe Accident Conditions

Radionuclide behavior during various severe accident conditions has been addressed as one of the important issues to discuss environmental safety in nuclear power plants. The present paper deals with the development of analytical models and their validations for the agglomeration of multiple-component aerosol and spray removal that controls source terms to the environment of both aerosols and gaseous radionuclides during recirculation mode operation in a containment system for a light water reactor. As for aerosol agglomeration, the single collision kernel model that can cover all types of two-body collision of aerosol was developed. In addition, the dynamic model that can treat aerosol and vapor transfer leading to the equilibrium condition under the containment spray operation was developed. The validations of the present models for multiple-component aerosol growth by agglomeration were performed by comparisons with Nuclear Safety Pilot Plant (NSPP) experiments at Oak Ridge National Laboratory (ORNL) and AB experiments at Hanford Engineering National Laboratory (HEDL). In addition, the spray removal models were applied to the analysis of containment spray experiment (CSE) at HEDL. The results calculated by the models showed good agreements with experimental results.     

United States Nuclear Regulatory Commission Research Program on molten core debris interactions in the reactor cavity

The NRC's Research Program on Core-Debris/Cavity Interactions comprises two principal elements: (1) an analytical effort focused primarily on development of computer codes needed to predict the potential consequences of risk-significant severe-accident scenarios; and (2) an experimental component to provide insights into the relevant phenomenological processes and to develop the experimental data base necessary for validation of the codes. The analytical activities at Sandia National Laboratories (SNL) focus primarily on refinement and validation of the CORCON and VANESA codes. Two major experimental activities are also based at SNL: (1) the large-scale SURC tests address the thermal-hydraulic phenomena in the cavity as well as aerosol release associated with prototypical core-melt materials in various types of concrete crucibles, while (2) the WITCH and GHOST experiments are concerned with aerosol generation and radionuclide release phenomena. A program of small-scale special-effects tests at Brookhaven National Laboratory (BNL) is coupled to a concomitant model-development and code-validation activity. In addition, measurements are being made at Battelle Columbus Laboratory (BCL) to augment the thermochemical data base needed in the VANESA code to permit refined radiological source-term predictions. The current scope and status of this research is reviewed.     

Comparison of ICECO code predictions with flexible vessel experiments

Some comparisons of ICECO code predictions with experimental data concerning transient fluid-structure interaction are given. The test results are taken from flexible vessel experiments conducted by Stanford Research Institute under the direction of Argonne National Laboratory. Two different experiments are considered: one with a rigid core barrel, and one with a flexible core barrel. Both experiments are performed in simple reactor vessels with a well-defined energy source and simple boundary conditions. Correlations of pressures and impulses are made at all available gauge stations. The permanent deformations of the core barrel and the cylindrical vessels are compared with ICECO predictions. The effects of core barrel flexibility on the wave propagation and vessel deformation are also investigated. The agreement between the analysis and experiments is found to be quite good.     

A study on compatibility of experimental effective atomic numbers with those predicted by ZXCOM

In this study, effective atomic numbers(Zeff) of materials determined at different experimental conditions by measuring the elastic-to-inelastic γ-ray scattering ratios are compared to ZXCOM predictions. It also presents the experimental data obtained via the transmission technique The agreement and disagreement between ZXCOM and experimental values are investigated. The theoretical basics of determining Zeffby scattering mode are outlined. The study shows that choosing appropriate experimental conditions can provide a good compatibility between the experimental results and theoretical ZXCOM calculations     

Neutron production from polyethylene and common spacecraft materials

We report experimental measurements of neutron production from collisions of neutron beams with polyethylene blocks simulating tissue at the Los Alamos National Laboratory Neutron Science Center and 1 GeV/amu iron nuclei with spacecraft shielding materials at the Brookhaven National Laboratory AGS.     

Pneumatic pressure tests of steel containment models — Recent developments

This paper describes the results of recent pneumatic pressure tests of steel containment models. These tests are part of the Containment Integrity Program whose objective is the qualification of methods for predicting containment response during severe accidents and extreme environments. Sandia National Laboratories is conducting this combined experimental and analytical program for the U.S. Nuclear Regulatory Commission (NRC). The long-range plans for the program include the following three containment loading conditions: static internal pressurization, dynamic internal pressurization, and seismic loadings. Steel, reinforced concrete, and prestressed concrete containment types are being considered.In the present experimental effort, models of steel containment structures are being subjected to static internal pressurization. The first set of models are about the size of hybrid-steel containments. Tests of these models are nearly finished. Testing of a large steel model, about of full size, will complete the static pressure experiments with steel models. Analysis of the models is paralleling the experimental effort.The Containment Integrity Program is being coordinated with other NRC programs on potential leakage of penetrations in containments. The results from all of the programs should provide a basis for predicting the structural and leakage behavior of containments during temperature and internal pressure loadings.     

Improvement of HYCA3D code and experimental verification in rectangular geometry

Concerns about the local hydrogen behavior in a nuclear power plant (NPP) containment during severe accidents have increased with the 10CFR50.34(f) regulation after TMI accident. Consequently, investigations on the local hydrogen behaviors under severe accident conditions were required. An analytical model named HYCA3D was developed at Seoul National University (SNU) to predict the thermodynamics and the three dimensional behavior of a hydrogen/steam mixture, within a subdivided containment volume following hydrogen generation during a severe accident in NPPs. In this study, the HYCA3D code was improved with a steam condensation and spray model, and verified with hydrogen mixing experiments executed in a SNU rectangular mixing facility. Helium was used to simulate hydrogen in both the calculations and the experiments. The calculation results show good agreement with the experimental data.     

Analysis of stresses in GCFR cladding under normal operating conditions

A modified version of the LIFE-III code, LIFE-GCFR, and classical stress analysis techniques have been employed to calculate the stresses in GCFR cladding under normal reactor operating conditions. Several types of loadings on the cladding which occur during normal operation have been considered. These include fuel-cladding mechanical interaction, thermal stresses induced by radial and axial temperature gradients, and stresses induced by swelling gradients. The combined and individual effects of these loadings as well as the effect of creep on cladding stresses have been assessed. Results obtained from this study have provided input to the experimental GCFR cladding development work at Argonne National Laboratory.     

Using the tritium plasma experiment to evaluate ITER PFC safety

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capabilty of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 × 10²³ ions/m².s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures.Prepared for the U.S. Department of Energy, Office of Energy Research under DOE Idaho Field Office Contract DE-AC07-76ID01570.     

Interpretation of the results of the CORA-33 dry core boiling water reactor test

All boiling water reactor (BWR) degraded core experiments performed prior to CORA-33 were conducted under ‘wet’ core degradation conditions, in which water remains within the core and continuous steaming feeds metal-steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ‘dry’ core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ‘dry’ core severe accident scenarios and to resolve partially phenomenological uncertainties concerning the behavior of relocating metallic melts that drain into the lower regions of a ‘dry’ BWR core (the ex-reactor experiments at Sandia National Laboratories will further address these uncertainties). CORA-33 was conducted on 1 October 1992, in the CORA test facility at Karlsruhe. A review of the CORA-33 data indicates that the objectives were achieved; i.e. core degradation occurred at a core heat-up rate (characterized by the absence of any temperature escalation caused by oxidation) and a test section axial temperature profile (at incipient structural melting) that are prototypic of full-core nuclear power plant simulations under ‘dry’ core conditions. Simulations of the CORA-33 test at Oak Ridge National Laboratory (ORNL) have required the modification of existing control blade-canister materials interaction models to include the eutectic melting of the stainless steel-zircaloy interaction products and the heat of mixing of stainless steel and zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the post-test analyses carried out at ORNL based on the experimental data and the post-test examination of the test bundle at Karlsruhe. The implications of these results with respect to degraded core modelling and the associated safety issues are also discussed.     

Flow-induced plastic collapse of stacked fuel plates

Flow-induced plastic collapse of stacked fuel plate assemblies was first noted in experimental nuclear reactors such as the Oak Ridge National Laboratory High Flux Reactor Assembly and the Engineering Test Reactor (ETR). The ETR assembly is a stack of 19 thin flat rectangular fuel plates separated by narrow channels through which a coolant flows to remove the heat generated by the nuclear fission of the fuel within the plates. The uranium alloyed plates have been noted to buckle laterally and plastically collapse at the system design coolant flow rate of 10.7 m/s, thus restricting the coolant flow through adjacent channels. In this paper a methodology and criterion are developed for predicting the plastic collapse of ETR fuel plates. The criterion is compared to some experimental results and the Miller critical velocity theory.     

Numerical simulation of seismic sloshing of LMR reactors

Numerical simulations of EPRI/CRIEPI sloshing experiments have been performed by Argonne National Laboratory using the ANL-developed FLUSTR computer code. The number of meshes used in the mathematical model for numerical simulation is rather small. Thus, the computing cost is relatively inexpensive. Results of numerical simulations of the sloshing responses of two test configurations (1 and 2) which were performed by CRIEPI are described in detail. Natural frequencies and sloshing wave heights and fluid pressures at locations of sensors are calculated. The predicted values are compared with the experimental data. In all comparisons, the agreement is very good. Thus, these computer codes can be used for numerical simulation of seismic sloshing.     

Failure characteristics of cladding tubes under RIA conditions

In the frame of actions for improving the safety of its nuclear power plants, Electricité de France needs to build the mechanical criteria ensuring the clad integrity for several operating conditions.This paper presents analytical mechanical models used to derive failure criteria for reactivity insertion accidents (RIA) and interpretation of the CABRI REP-Na experimental tests.Building analytical criteria requires an experimental database. Mechanical tests performed on non-irradiated and irradiated cladding tubes have been provided from French and international programmes (PROMETRA, EPRI, …). These tests consist of tube burst and axial tension, and ring tension. Several strain biaxiality ratios are thus available: pure circumferential tension (from ring tension), pure axial tension (from tube axial tension), and plane strain conditions (from tube burst tests). Several strain rates, temperatures, irradiation conditions are also available.The major feature of our study has been to make it possible that these several thermomechanical conditions be representative of “standard” RIA loading conditions.To this aim, we have derived some biaxiality and strain rate corrections to be applied to the results of experimental tests, in such a way that they could be representative of RIA biaxiality conditions (which are assumed to be strain equibiaxiality), and also representative of RIA strain rate conditions (which are assumed to be 5 s⁻¹).The corrections that we derive are based on the fracture properties of hydrided zirconium alloys (especially in terms of anisotropy), and also on an assumed form of the material constitutive equations.Each test of the “homogenized” database has thus been used to calculate a strain energy density, representative of its fracture (the strain energy density is defined as the integral of the stress times strain rate states, over the duration of the mechanical test). The SED values are plotted against the sample's oxide thickness, and a lower bound limit can be established, with respect to oxide thickness.In order to address the problem of representativeness of the laboratory database, an experimental set-up has been developed that aims at characterizing the failure behavior of cladding tubes under RIA conditions. The developed experimental set-up is based on electromagnetic forming.The development of the test and in particular of the die is delicate but leads to repeatable results with a controlled strain biaxiality ratio higher than those obtained through conventional tests such as ring tests or PSU ring tests. The use of electromagnetic forming process allows testing the specimen with very high strain rates. For the next test series other zircaloy alloys at the reception state and in hydrided conditions will be tested in order to look at hydrides influence on fracture strains.A first finite element simulation of the test was engaged. The simulation and the experimental results are in quite good agreement. In future, the consistency of the previously developed analytical mechanical criterion with the electromagnetic forming experimental results will be verified.     

An integral loop test and MARS code analysis for a DVI line break LOCA in the APR1400

The transient thermal-hydraulic phenomena of a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in pressurized water reactor, APR1400, were investigated. In order to understand the phenomena during the LOCA transient, a reduced-height and reduced-pressure integral loop test facility, the SNUF (Seoul National University Facility), was constructed with scaling down the prototype. For the appropriate test conditions in the experiment with the SNUF, the energy scaling method was suggested with scaling the coolant mass inventory and the thermal power for the reduced-pressure condition. According to the conditions determined by the method, the experimental study was performed with the SNUF. The experimental results showed that the phenomenon of the downcomer seal clearing played a dominant role in the reduction of the system pressure and the recovery of the coolant level in the core. That phenomenon occurred when the steam incoming from cold legs penetrates the coolant in the upper downcomer toward the broken DVI line. The experimental results were compared with the prototype analysis to estimate the energy scaling method, so that the experiment reasonably simulated the phenomena in the prototype. For the analytical investigation, the experiment was simulated with MARS code to validate the calculation capability of the code, especially for the downcomer seal clearing, which showed good agreement with the results of experiment.     

Test Facilities for Radioactive Materials Transport Packages (Sandia National Laboratories,USA)

Abstract

Compliance with regulatory requirements for normal conditions and hypothetical accident environments can be demonstrated by testing packages developed to transport radioactive material. A comprehensive testing capability has been developed at Sandia National Laboratories to simulate the required test conditions and provide response data. Several of the major facilities available to support testing of radioactive material packages are described.     

RELAP5/3.2 assessment against low pressure onset of flow instability in parallel heated channels

Best Estimate computer codes have been, so far, developed for safety analysis of nuclear power plants and were extensively validated against a large set of separate effects and integral test facilities experimental data relevant to such kind of reactors. Their application to research reactors is not fully straightforward. Modelling problems generally emerge when applying existing models to low pressure and more particularly to subcooled flow boiling situations. The objective of the present work is to investigate the RELAP5/3.2 system code capabilities in predicting phenomena that could be encountered under abnormal research reactor’s operating conditions. For this purpose, the separate effect related to the static onset of flow instability is investigated. The cases considered herein are the flow excursion tests performed at the Oak Ridge National Laboratory thermal hydraulic test loop (THTL) as well as some representative Whittle and Forgan (W & F) experiments. The simulation results are presented and the capabilities of RELAP5/Mod 3.2 in predicting this critical phenomenon are discussed.     

A numerical study of flow in the Thermal-Hydraulic Out-of-Reactor Safety facility expansion tank

A new, fast-running, two-dimensional computer code was developed to model the flow and temperature patterns in the expansion tank of the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility. The THORS facility, located at the Oak Ridge National Laboratory (ORNL), is an engineering-scale sodium loop used for thermal-hydraulic testing of simulated Liquid Metal Fast Breeder Reactor (LMFBR) subassemblies. In the computer model, the fluid is considered Boussinesq and a simple turbulence model is provided so that a wide range of inlet conditions can be studied. A vorticity-stream function formulation is used on a uniform finite difference grid. The model also includes the thermal response of the tank wall. The results of simulated experiments for natural circulation and forced flow conditions are compared. Inflow boundary conditions were adjusted to simulate boiling in the THORS test section upstream of the expansion tank during some runs made with the code. Streamline and isotherm plots of the results are presented. All cases studied reached thermally stratified conditions in the tank, and regions of buoyancy and convection-dominated flow are observed.