Committee On The Safety Of Nuclear Installations Research Articles

Development of a fully implicit ODE-solver for containment analysis code

The thermal–hydraulic dynamics in containment are governed by a system of stiff ordinary differential equations (ODEs). A fully implicit discretization scheme is adopted to discretize these ODEs in order to mitigate the effects of stiffness. In comparison with explicit or semi-implicit discretization schemes that are subject to Courant limits on time steps, the fully implicit discretization scheme is more suitable for a containment analysis code that focuses on predicting both short-term and long-term thermal–hydraulic parameters after an accident. This study introduces a general-purpose ODE solver for the containment analysis code. The outline of the solver is as follows: The fully implicit discrete equations lead to a large set of nonlinear equations that need to be solved using Newton’s iterative method. The partial derivative components in the Jacobi matrix are calculated by the perturbation method using finite difference approximation, which avoids the complicated derivation of partial derivatives. The scaling modification technique is incorporated into this ODE solver to deal with significant differences in unknown variable magnitudes, and the line search method is introduced to address the difficulty of obtaining an accurate root estimate with Newton’s method when the initial guess is far from the actual root. This proposed ODE solver was applied to two typical stiff ODE problems to test its stiffness-suppressed ability and to demonstrate that this proposed solver can perform calculations with a very large time step. Then, the CASSIA code, a containment analysis code developed by China Nuclear Power Technology Research Institute Co., Ltd (CNPRI), equipped with this ODE solver, was applied to the CSNI (Committee on the Safety of Nuclear Installations) benchmark problem and the Carolinas Virginia Tube Reactor (CVTR) test 3 problem to preliminarily demonstrate that the proposed ODE solver can perform containment thermal–hydraulic analysis correctly. This study could provide references for the development of a home-made containment analysis code.

OECD/NEA R&D priorities for loss-of-cooling and loss-of-coolant accidents in Spent-Fuel-Pools

Motivated by the 2011 accident at the Fukushima Daiichi Nuclear Power Station, the Nuclear Energy Agency (NEA) Committee on the Safety of Nuclear Installations (CSNI) of the Organisation for Economic Cooperation and Development (OECD) launched several activities with the aim to support the post-Fukushima accident decision making process. One of these activities, discussed in the present paper, consisted in the production of a Phenomena Identification and Ranking Table (PIRT) in order to systematically highlight Spent-Fuel-Pool (SFP) accident-related phenomena that are of both high importance and high uncertainty, and thus of primary interest for further experimental or modeling activities. For that purpose, an international panel of experts identified and ranked more than a hundred physical phenomena with regard to their safety importance and current level of knowledge. Altogether eighteen unique phenomena were identified as having priority research needs. About half of these phenomena are related to thermal-hydraulics and heat transfer in the SFP, and they are judged to be important to the coolability of the spent fuel in loss-of-cooling and/or loss-of-coolant accidents. The other half comprises fuel-related phenomena such as fuel volatilization, cladding oxidation in mixed steam–air environment and nitrogen-assisted oxide breakaway at moderate temperature, judged as important and poorly modeled in the scope of SFP accidents. The final part of the paper discusses the research programmes performed on Spent Fuel Pool loss-of-cooling and loss-of-coolant accidents since the publication of the PIRT report in 2018 and the remaining R&D needs.

Regulatory and safety aspects of ageing in nuclear power plants

Abstract The OECD Nuclear Energy Agency (NEA) is a semi-autonomous body within the OECD established in 1958 with the mandate to promote co-operation among the governments of its participating countries in furthering the development of nuclear power as a safe, environmentally acceptable and economic energy source. While all of groups have detailed programmes involving important aspects, this paper will focus specifically on the work of Committee on Nuclear Regulatory Activities (CNRA) and the Committee on the Safety of Nuclear Installations (CSNI).

Effect of power uprate on core damage frequency in a MBLOCA with multiple operator actions

In the present study, the methodology developed by the Committee on the Safety of Nuclear Installations (CSNI) of Nuclear Energy Agency was used to quantify the changes in core damage frequency (CDF) owing to a power uprate. The Maanshan Nuclear Power Station of Taiwan Power Company was the plant selected for the analysis. The plant employs a 3-loop pressurized water reactor designed by Westinghouse. The power level was increased from 102% to 105%. The sequence analyzed corresponded to a medium-break loss-of-coolant accident with a break size of 3in. There were two operator actions in the accident mitigation. These actions were the emergency cooldown and depressurization (CND) of RCS and the lineup the high-head safety recirculation (HHSR) when RWST was empty.In the present study, the success of operator action was judged based on the results of the simulation of system thermal hydraulic code, RELAP5-3D. An uncertainty analysis of code calculation was performed. There were 20 parameters in PIRT table, and 124 sets of input data were generated using the Monte Carlo Sampling technique. The times for the operators to execute CND and HHSR were included as one of the parameter in the uncertainty analysis. The results demonstrated that, compared with the PSA results using the conventional HEP model, the CDF of MBLOCA increased by 188%. The dominant core damage sequence in MBLOCA was also changed.The simulation results showed that the CND failure probability increased by 23% after the power uprate. The sequence CDF increased by 16% after the power uprate. The results of the importance analyses indicated that the time to execute CND was not a significant factor in determining the success of CND. The uncertainty involved in the simulations of reactor coolant system thermal hydraulic responses is important in the determination of human error probability.

OECD/NEA Activities to Support Long Term Operation

The Integrity and Aging of Components and Structures Working Group (IAGE) of the Organisation for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA) was established under the Committee on the Safety of Nuclear Installations (CSNI) for three reasons: (i) to advance the current understanding of those aspects relevant to ensuring the integrity of structures, systems, and components ; (ii) to provide for guidance in choosing the optimal ways to handle challenges to the integrity of operating as well as new nuclear power plants, and (iii) to take an integrated approach to design, safety, and nuclear power plant life management. The group operates through annual plenary meetings and technical workshops and by issuing state-of-the-art reports and topical opinion papers. This paper details some recent IAGE activities and products, focusing on those dealing with the degradationmechanisms of metal and concrete components.

Main Results of Phase IV BEMUSE Project: Simulation of LBLOCA in an NPP

Phase IV of BEMUSE Program is a necessary step for a subsequent uncertainty analysis. It includes the simulation of the reference scenario and a sensitivity study. The scenario is a LBLOCA and the reference plant is Zion 1 NPP, a 4 loop PWR unit. Thirteen participants coming from ten different countries have taken part in the exercise. The BEMUSE (Best Estimate Methods plus Uncertainty and Sensitivity Evaluation) Program has been promoted by the Working Group on Accident Management and Analysis (WGAMA) and endorsed by the Committee on the Safety of Nuclear Installations (CSNI). The paper presents the results of the calculations performed by participants and emphasizes its usefulness for future uncertainty evaluation, to be performed in next phase. The objectives of the activity are basically to simulate the LBLOCA reproducing the phenomena associated to the scenario and also to build a common, well-known, basis for the future comparison of uncertainty evaluation results among different methodologies and codes. The sensitivity calculations performed by participants are also presented. They allow studying the influence of different parameters such as material properties or initial and boundary conditions, upon the behaviour of the most relevant parameters related to the scenario.

International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA)

Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs). These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2) and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium) were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects. ISPs are providing unique material and benefits for some safety-related issues.

Assessment of the computer code NAUA Mod 5 against a small-scale containment aerosol test

The Committee on the Safety of Nuclear Installations (CSNI) of the Organisation for Economic Cooperation and Development (OECD) organized in 1992–1994, an International Standard Problem Exercise Number 34 (ISP-34), on aerosol behaviour in the primary circuit and the containment of a nuclear reactor. The exercise involved carrying out specific experiments in the Falcon Test facility at Winfrith Technology Centre, U.K., analysis of these experiments using different computer codes. Containment analysis of these tests was carried out by the Indian participants BARC using the computer code NAUA Mod 5. The present paper compares the code predictions with experimental measurements. The containment floor deposits are predicted closely, whereas the wall deposits are considerably under-predicted. The air-borne concentrations predicted by the code are well within the uncertainties of measurement. The possible reasons for the differences are brought out.

Severe Accidents in Nuclear Power Plants

Nuclear Energy Agency Group of Experts 1 986 Paris: OECD 32 pp price £3.50 (paperback) ISBN 92 64 12821 2 Commissioned back in 1979 by the Committee on the Safety of Nuclear Installations (CSNI) of the NEA, following the Three Mile Island reactor accident, this report was completed some four years later. It deals with severe, beyond the design basis, accidents in light water reactors (PWRS and BWRS) and discusses the extent to which the design features of current plant may prevent or limit their consequences.