A Simple mass and heat balance model for estimating plant conditions during the Fukushima Dai-ichi NPP accident

Shibamoto, Yasuteru; Moriyama, Kiyofumi*; Maruyama, Yu; Yonomoto, Taisuke

Journal of Nuclear Science and Technology, 49(8), p.768 - 781, 2012/08

https://doi.org/10.1080/00223131.2012.703942

A simple evaluation method for the analysis of thermal-hydraulic transients in reactor pressure vessel (RPV) and primary containment vessel (PCV) is proposed to support understanding the accident behaviors of the Fukushima Daiichi Nuclear Power Plant (NPP). Since most of the measurements of the plants were unavailable especially in the early stage of the accident, and the accessibility to the plants has been limited by radiation, analytical investigation for the plant is urgently required to understand the plant conditions such as the magnitude of the damages. In order to provide easy-to-use technical tools to support the analytical investigation, we developed a simplified analysis code, named "HOTCB", based on total mass and heat balances in a lamped parameter system. The HOTCB code has capabilities to treat two-phase fluid including water, steam, and non-condensable gas in a wide range of temperature up to highly superheated conditions, and to consider heat structures, i.e. heat capacities and heat transfer to the fluid. The code was provided to Tokyo Electric Power Company (TEPCO) and was practically used for the analysis on the accident. This paper provides the details of the code and simulations of Unit 1 and Unit 2 reactors of Fukushima Dai-ichi NPP as examples to show the usefulness of the code.

Experiment on the gaseous iodine release from irradiated cesium iodide solutions (Contract research)

Moriyama, Kiyofumi; Tashiro, Shinsuke; Chiba, Noriaki; Maruyama, Yu; Nakamura, Hideo; Watanabe, Atsushi*

JAEA-Research 2011-016, 125 Pages, 2011/06

JAEA-Research-2011-016.pdf:2.71MB

The volatile iodine production due to radiation chemical effects in the containment vessel of light water reactors (LWRs) during severe accidents was investigated by experiments in small scale and with well controlled conditions. Cesium iodide solutions, 10M, labeled with I, at controlled pH by boric acid-sodium hydroxide buffer, were -irradiated and swept with a constant gas flow rate. The gaseous iodine released from the solution was collected by species selective filters and quantified separately for I and organic iodines. The influences of pH, temperature, inorganic and organic impurities, oxygen and hydrogen concentrations in the cover gas on the iodine release behavior were examined. Data including time dependent gaseous iodine release fractions, comparison of the final iodine release fractions in terms of the parameter effects, as well as the initial, boundary and interface conditions necessary for simulating the experiments by computer codes are provided.

Kiche; A Simulation tool for kinetics of iodine chemistry in the containment of light water reactors under severe accident conditions (Contract research)

Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo

JAEA-Data/Code 2010-034, 62 Pages, 2011/03

JAEA-Data-Code-2010-034.pdf:1.43MB

An iodine chemistry simulation tool, Kiche, was developed for analyses of chemical kinetics relevant to iodine volatilization in the containment vessel of light water reactors (LWRs) during a severe accident. It consists of a Fortran code to solve chemical kinetics models, reaction databases written in plain text format, and peripheral tools to convert the reaction databases into Fortran codes. Potential advantages of Kiche are the text format reaction database separated from the code that provides flexibility of the chemistry model, and being a Fortran code relatively easily coupled with other codes such as severe accident analysis codes. This document describes the model, solution method, code structure, and examples of application for simulation of experiments. The appendixes give practical information for the usage of Kiche system.

Radiolytic iodine volatilization with organic impurities and low oxygen concentrations

Moriyama, Kiyofumi; Nakamura, Hideo; Nakamura, Koichi*

Transactions of the American Nuclear Society, 103(1), p.463 - 464, 2010/11

A computer code, Kiche, was developed at JAEA for the evaluation of the influence of radiolytic volatilization of iodine in the containment of light water reactors during an accident. It solves a chemical kinetics model including water radiolysis, reactions among iodine species, water radiolysis products, and organic compounds. The reactions and rate constants were collected from the literature. The organic iodine production model of Kiche, which was initially based on LIRIC model, was not applicable to low oxygen concentration conditions like inserted containments of BWRs. We modified the model by considering different reaction paths according to the availability of oxygen. The modified model showed improved agreement with our experimental data with various concentrations of organic impurity and oxygen.

Design and fabrication of an ultra-high input impedance low offset amplifier for electro-chemical measurements

Sagawa, Jun; Moriyama, Kiyofumi; Nishikizawa, Tomotoshi; Nakamura, Hideo

JAEA-Technology 2008-059, 43 Pages, 2008/09

JAEA-Technology-2008-059.pdf:8.45MB

Electro-chemical electrodes including pH probes, ion-selective electrodes (ISEs) etc. generally have very high output impedances. In order to measure their outputs with generic measurement devices like data recorders, we need impedance conversion amplifiers that convert the ultra-high impedance signals of the probes into low-impedance input signals for ordinary measurement devices. Although specially designed measurement devices for the electro-chemical probes are commercially available, there are very few products that can be applied for multi-channel time series data acquisition. Thus, we designed and fabricated an ultra-high impedance low-offset amplifier fit for this purpose. The primary specification of the amplifier is, input impedance 10G, input range 1V, gain 120, response time about 1s, output range 10V, output impedance 50, and it has 5 independent channels. This report describes the originally developed design, selection of the element devices, test on the circuit characteristics, and instruction for fabrication.

Evaluation of containment failure probability due to ex-vessel steam explosions in LWR severe accidents

Moriyama, Kiyofumi; Takagi, Seiji*; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu

JAEA-Research 2007-072, 54 Pages, 2007/11

JAEA-Research-2007-072.pdf:25.89MB

The containment failure probability due to ex-vessel steam explosions was evaluated for BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6.4E-2 (mean) and 3.9E-2 (median) for the BWR suppression pool case, 2.2E-3 (mean) and 2.8E-10 (median) for the BWR pedestal case, and 6.8E-2 (mean) and 1.4E-2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information. Additionally, the source term significance of the fine particles generated by steam explosions was discussed.

Evaluation of containment failure probability by Ex-vessel steam explosion in Japanese LWR plants

Moriyama, Kiyofumi; Takagi, Seiji*; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu*

Journal of Nuclear Science and Technology, 43(7), p.774 - 784, 2006/07

https://doi.org/10.3327/jnst.43.774

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. The mean conditional containment failure probabilities (CCFPs) were (mean) and (median) for the BWR suppression pool case, (mean) and (median) for the BWR pedestal case, and (mean) and (median) for the PWR cavity case. Note that the specific values of the probability are most dependent on assumed range of melt flow rates and on fragility curves that involve conservatism and uncertainty due to simplified scenarios and limited information. Also, note that these CCFPs were based on the preconditions of failure of accident termination within the reactor vessel, relocation of the core melt into the water pool at the place in question without significant interference, and a strong triggering ofa steam explosion with maximized premixed mass for the given premixing condition. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for these preconditions. Thus, the CCFPs per core damage should be lower than the values given above.

A Strategy for the application of steam explosion codes to reactor analysis

Moriyama, Kiyofumi; Nakamura, Hideo

Proceedings of Technical Meeting on Severe Accident and Accident Management (CD-ROM), 18 Pages, 2006/03

The steam explosion is one of the phenomena that may threat theintegrity of containment vessel during severe accidents of light waterreactors (LWRs), and has been drawing attention in the field of nuclear safety as well as in many other industrial fields for decades. We developed a code, JASMINE, for the assessment of steam explosion impacts on the LWR safety, and applied the code for simulation of steam explosion experiments and also reactor scale parametric studies. Based on the analytical experiences with JASMINE code and aconsideration on the inter-relation among various fundamental aspects in the steam explosion process, we developed a technical view on aproper application of mechanistic steam explosion analysis codes for plant analysis.

Feasibility study of a high-speed X-ray visualization method for multiphase interactions

Usami, Tsutomu; Moriyama, Kiyofumi; Nishikizawa, Tomotoshi; Nakamura, Hideo

JAERI-Tech 2005-028, 37 Pages, 2005/05

JAERI-Tech-2005-028.pdf:16.32MB

The steam explosion during a severe accident in a light water reactor, which may occur by the contact of molten core and coolant, has been known as a potential threat on the integrity of the containment vessel, and has been studied in the nuclear safety research field. Prediction of the intensity of steam explosions needs an understanding of the initial premixture. However, visual observation of the premixture in experiments is usually difficult due to the vapor generation. Thus, we investigated the possibility of a high-speed X-ray visualization with a 4500 f/s high speed video camera with an image-intensifier and three kinds of scintillator materials: CdWO, ZnS(Ag) and CsI(Tl). A modeled premixture consisted of metal objects and bubbles in a water vessel was used. The test result showed that the CsI(Tl) scintillator gave the best image quality among the three and the high-speed visualization at 4500 f/s was possible, though in the still picture which carried out one of the recorded picture, the picture became indistinct.

Simulation of alumina and corium steam explosion experiments with JASMINE v.3

Moriyama, Kiyofumi; Nakamura, Hideo; Maruyama, Yu

Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 18 Pages, 2004/10

A steam explosion simulation code JASMINE is under development at JAERI for the assessment of steam explosion impacts on the integrity of containment vessel during severe accidents in light water reactors. Selected alumina and corium steam explosion experiments, KROTOS-44, 42, 37 and FARO-L33 were simulated with JASMINE code. The experimentally observed difference of the steam explosion intensity with the two materials, alumina and corium, was reproduced in the simulations without changing the model parameters related to the fine fragmentation process, but based on the difference in the premixing behavior predicted by the simulations. The simulation of corium experiments showed more fraction of the melt droplets frozen during premixing, as well as more void fraction, and those two points were likely to be the primary reasons of weak interactions in corium experiments.