European expert network for the reduction of uncertainties in severe accident safety issues (EURSAFE)

EURSAFE thematic network was a concerted action in the sixth framework programme of the European Commission. It established a large consensus among the main actors in nuclear safety on the severe accident issues where large uncertainties still subsist. The conclusions were derived from a first-of-kind phenomena identification and ranking tables (PIRT) on all aspects of severe accident also realised in the frame of the project. Starting from a list of all severe accident phenomena containing approximately 1000 entries and established by the twenty partner organisations, 106 phenomena were retained eventually as both important for safety and still lacking sufficient knowledge. Ultimately, 21 research areas for addressing these phenomena regrouped according to their similarities were identified. A networking structure for implementing and executing the necessary research was proposed, which promotes integration and harmonisation of the different national programmes. A severe accident database structure was proposed to ensure preservation of experimental data and enhanced communication for data exchange and use for severe accident codes assessment. The final product, named EURSAFE, is a website network, http://asa2.jrc.it/eursafe, connecting nodes located at partner sites. As the result of an action involving R&D governmental institutions, regulatory bodies, nuclear industry, utilities and universities from six EU Member States (Finland, France, Germany, Spain, Sweden, UK) plus JRC, three European third countries (Czech Republic, Hungary, Switzerland), and USA, EURSAFE represents a significant step towards harmonisation and credibility of the approaches, and resolution of the remaining severe accident issues.     

Development and analysis of LOCA sequences for severe accident risk database

Although a Level 2 PSA has been performed for the Korean Standard Power Plants (KSNPs), and that it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access of the results. At present, KAERI is intensively calculating the severe accident sequences for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviors. The developed database (DB) system includes a graphical display for a plant and equipment status, previous research results by a knowledge-based technique, and the expected plant behavior. The plant model used in this paper is oriented to the cases of LOCAs related to severe accident phenomena and thus can simulate the plant behaviors of a severe accident. Therefore, the developed system may play a central role as an information's source during the decision-making for a severe accident management, and be used as a training simulator for a severe accident management.     

Evaluation of plutonium oxide aerosol release from an LMFBR in a hypothetical accident

This work was undertaken to prepare a computer code for the hazard evaluation of plutonium oxide aerosol released to the atmosphere in the event of a hypothetical accident in a 50 MW(th) scale LMFBR. The reactor building structure consists of semi-double containments as follows: the primary containment has a large volume in comparison with the secondary annular containment in which a part is connected to the atmosphere through an emergency filter system. Sodium oxide aerosol containing PuO₂---UO₂ fuel, fission products and structural steel agglomerates quickly by coagulation due to its high concentration. Simultaneously, the aerosol concentration decreases due to settling, plating and thermophoresis. Using the present code, the amount of PuO₂ aerosol leakage to the atmosphere was evaluated.     

Development of a risk-informed accident diagnosis and prognosis system to support severe accident management

RISARD, risk-informed severe accident risk diagnosis system, is a computerized tool developed to improve a severe accident management (SAM) for a nuclear power plant and to effectively support the MCR and the TSC in executing the relevant SAM activities. In order to provide a diagnostic capability to a state of the plant and a prognostic capability for an anticipated accident progression, the system examines (a) a symptom-based diagnosis of a plant damage state (PDS) sequence in a risk-informing way and (b) a PDS sequence-based prognosis of key plant parameter behavior, through a prepared database (DB) containing plant-specific severe accident risk (SAR)-related information. For a given accident, the replicated use of these two processes makes it possible to obtain information about the functional states of the plant and containment safety systems expected at the time of a severe accident as well as future trend of the key plant parameters that are essentially required for taking the relevant SAM action, eventually leading to an answer about the best strategy for SAM. The foregoing concept for an accident diagnosis and prognosis can give the SAM practitioners more time to take action for mitigating the consequences of the potential accident scenarios since they are made in a simple, fast, and efficient way through a prepared SAR database and it is useful especially when the plant information available for SAM is incomplete and limited. The main purpose of this paper is to (a) introduce the concept of the RISARD system proposed to support SAM and its implementation through a prepared OPR1000 plant- and MAAP code-specific SAR database and (b) assess prediction capabilities of major events expected during the evolution of a severe accident through the system.     

Core to surge-line energy transport in a severe accident scenario

The analysis of loss-of-coolant accidents in a nuclear power plant, which progress to the stage where the core is uncovered, poses important safety related questions. One of these questions concerns the rate of energy transport to metal components of the primary system. An experimental program has been conducted at the University of Maryland test facility which quantifies the rate of energy transfer from an uncovered core in a B&W plant (once-through type steam generators). SF₆ is used to simulate the high pressure steam at prototypical conditions. A time-dependent scaling methodology is developed to transpose experimental data to prototypical conditions. To achieve this transformation, a nominal fluid temperature increase rate of 1.0 °C s⁻¹ is inferred from available TMI-2 event data. To bracket the range of potential prototypical transient scenarios, temperature ramps of 0.8 °C s⁻¹ and 1.2 °C s⁻¹ are also considered. Repeated tests, covering a range of test facility conditions, lead to estimated failure times at the surge line nozzle of 1.5–2 h after initiation of the natural circulation phase of the transient.     

Retention factors for fission products from sodium tests to simulate a severe LMFBR accident

In case of an LMFBR whole-core accident, fuel and fission products may be released instantaneously from the HCDA bubble through vessel head leaks or delayed from hot or boiling sodium pools after vessel melt-through. It is necessary to investigate the radiological source terms for both scenarios. In the paper we present and discuss retention factors for simulated fission products and fuel from FAUST under-sodium rupture disk tests on the instantaneous source term and from NALA tests with hot and boiling sodium pools on the delayed source term.     

Core loss during a severe accident (COLOSS)

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H₂ production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO₂ and MOX by molten Zircaloy, (b) simultaneous dissolution of UO₂ and ZrO₂, (c) oxidation of U–O–Zr mixtures, (d) degradation–oxidation of B₄C control rods.Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B₄C control rods and in the TMI-2 accident.Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Break-throughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO₂ and MOX dissolution and oxidation of U–O–Zr and B₄C–metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H₂ production observed during the reflooding of degraded cores under severe accident conditions.The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results.Main results and recommendations for future R&D activities are summarized in this paper.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

小破口引发的严重事故工况及事故缓解的研究

利用MAAP4程序对方家山核电站进行建模,针对事故后果较为严重的小破口事件进行了计算分析,得到了假设事故下电厂系统的反应以及相应的严重事故现象.对事故中发生的DCH(安全壳直接加热)现象和安全壳失效以及裂变产物向环境的释放进行了分析.随后,本文根据相关的严重事故管理导则和该事故的特点,对缓解该事故的策略进行了研究和计算...     

Evaluation of safety of French type B package designs in severe accident environments other than regulatory

Abstract

Packages used to transport radioactive materials in France must be designed to meet the safety performance requirements when subject to the test conditions set forth in the International Atomic Energy Agency (IAEA) Regulations. During actual use, the packages may be subject to quite different accident conditions. The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) has evaluated the behaviour of typical packages designed to transport spent fuel, high activity waste, fresh mixed oxide (MOX) fuel and plutonium oxide powder under realistic conditions of mechanical impact and fire. The studied designs remain safe after impact onto targets present in the real environment of transport. The energy absorption by the package ancillary equipment (transport frame) compensates for the kinetic energy increase by comparison to the energy expended during the regulatory tests. New software was developed to correctly simulate the thermal behaviour of the neutron shielding materials. The studied package designs exhibit large margins of safety concerning resistance to fire. The results obtained have been used to develop tools in support of the appraisal of emergency situations.     

Effects of accident management strategy on the severe accident environmental conditions

This paper presents a methodology utilizing an accident management strategy in order to determine accident environmental conditions to be used as inputs to equipment survivability assessments. In the case that there is a well-established accident management strategy for a specific nuclear power plant (NPP), an application of this tool can provide a technical rationale on equipment survivability assessment so that plant-specific and time-dependent accident environmental conditions could be practically and realistically defined in accordance with the equipment and instrumentation required for the accident management strategy or appropriate actions. For this work, three different tools are introduced; probabilistic safety assessment (PSA) outcomes, major accident management strategy actions, and accident environmental stages (AESs). In order to quantitatively investigate an applicability of accident management strategy on equipment survivability, the accident simulation for most likely scenario in Korean standard nuclear power plants (KSNPs) is performed with the MAAP4 code. The accident management guideline (AMG) actions such as the reactor coolant system (RCS) depressurization, water injection into the RCS, the containment pressure and temperature control, and hydrogen concentration control in containment are applied. The effects of these AMG actions on the accident environmental conditions are investigated by comparison to actions from previous normal accident simulation, especially focusing on equipment survivability assessment. As a result, the AMG-involved case shows the higher accident consequences along the accident environmental stages. This implies that plant-specific AMG actions need to be considered in order to determine accident environmental conditions in equipment survivability assessments.     

Quantitative assessment of severe accident management strategies in a nuclear power plant

Nuclear power plants in Korea are preparing improvement countermeasures for severe accidents including mobile gas turbine generators, passive autocatalytic recombiners, containment-filtered venting systems, and external injection using portable pumps. However, these improvement countermeasures have only been determined by expert judgment, and detailed validation of their effects has not been performed. In this paper, the quantitative safety impact of these improvement countermeasures was evaluated for the Westinghouse 3-loop pressurized water reactor. Our evaluation of four improvement countermeasures using the at-power internal event probabilistic safety assessment models revealed that all containment failure modes have positive effects, except for the containment isolation failure and the containment bypass. Therefore, post-Fukushima action plans for coping with a severe accident in Korea have been appropriately evaluated and established.     

Use of a fuzzy decision-making method in evaluating severe accident management strategies

In developing severe accident management strategies, an engineering decision would be made based on the available data and information that are vague, imprecise and uncertain by nature. These sorts of vagueness and uncertainty are due to lack of knowledge for the severe accident sequences of interest. The fuzzy set theory offers a possibility of handling these sorts of data and information. In this paper, the possibility to apply the decision-making method based on fuzzy set theory to the evaluation of the accident management strategies at a nuclear power plant is scrutinized. The fuzzy decision-making method uses linguistic variables and fuzzy numbers to represent the decision-maker's subjective assessments for the decision alternatives according to the decision criteria. The fuzzy mean operator is used to aggregate the decision-maker's subjective assessments, while the total integral value method is used to rank the decision alternatives. As a case study, the proposed method is applied to evaluating the accident management strategies at a nuclear power plant.     

Classification method of severe accident condition for the development of severe accident instrumentation and monitoring system in nuclear power plant

Instrumentation and monitoring systems in a nuclear power plant are very important to monitor plant conditions for safe operations and a plant shutdown. The severe accident at TOKYO ELECTRIC POWER COMPANY's Fukushima Daiichi Nuclear Power Station (hereinafter called as TF1) in March 2011 caused several severe situations such as core damage, hydrogen explosion, etc. Lessons learned from the severe accident at TF1 show that an appropriate operable instrumentation and monitoring system for a severe accident should be developed so that the system will deliver an appropriate performance for mitigation of severe accident condition in a nuclear power plant.

This paper proposes the classification method of severe accident condition for the development of an appropriate operable instrumentation and monitoring system for a severe accident based on the problem analysis of monitoring variables during the severe accident at TF1. The classification is formed on the basis of the integrity of boundary for plant safety and the successful (or unsuccessful) condition of the cooling water injection, and is used for an establishment of defining severe accident environmental conditions for the instrumentation and monitoring system. Examples of the establishment method are also shown in this paper.     

Quantification of severe accident source terms of a Westinghouse 3-loop plant

Integrated severe accident analysis codes are used to quantify the source terms of the representative sequences identified in PSA study. The characteristics of these source terms depend on the detail design of the plant and the accident scenario. A historical perspective of radioactive source term is provided. The grouping of radionuclides in different source terms or source term quantification tools based on TID-14844, NUREG-1465, and WASH-1400 is compared. The radionuclides release phenomena and models adopted in the integrated severe accident analysis codes of STCP and MAAP4 are described. In the present study, the severe accident source terms for risk quantification of Maanshan Nuclear Power Plant of Taiwan Power Company are quantified using MAAP 4.0.4 code. A methodology is developed to quantify the source terms of each source term category (STC) identified in the Level II PSA analysis of the plant. The characteristics of source terms obtained are compared with other source terms. The plant analyzed employs a Westinghouse designed 3-loop pressurized water reactor (PWR) with large dry containment.     

Assessing the uncertainties in reactor accident consequences by means of RSM: The case of an ATWS in a PWR

A simulation study of a station black-out ATWS has been performed by applying Response Surface Methodology (RSM) on the data obtained by inspecting the ALMOD code.The case under study has shown that the a priori information which alone could be inadequate, is optimally utilized if coupled with a preliminary sensitivity analysis through RSM techniques. In particular the engineering selection of the model variables and the rank order of the remaining ones had to be modified after an RSM preliminary sensitivity analysis.Another qualifying feature of the exercise is the use of the randomization of the variables not included in the model in order to coherently exploit the methodology in its full efficiency. This procedure is able to give a figure of merit of the global importance of the neglected variables through the analysis of residuals.Results show that the proposed technique is an effective tool for selecting the most important accident variables and that the body of information gained is significant with respect to the number of observations performed.     

Iodine-steel reactions under severe accident conditions in light-water reactors

Owing to large surface areas, the reaction of volatile molecular iodine (I₂) with steel surfaces in the containment may play an important role in predicting the source term to the environment. Both wall retention of iodine and conversion of volatile into non-volatile iodine compounds at steel surfaces have to be considered. Two types of laboratory experiment were carried out at Siemens (KWU) in order to investigate the reaction of I₂ at steel surfaces representative for German power plants.

• 1. 

  (1) For steel coupons submerged in an I2 solution at T = 50, 90 or 140 °C the reaction rate of the I2−I− conversion was determined. No iodine loading was observed on the steel in the aqueous phase tests. I2 reacts with the steel components (Fe, Cr or Ni) to form metal iodides on the surface which are all immediately dissolved in water under dissociation into the metal and the iodide ions. From these experiments, the I2−I− conversion rate constants over the temperature range 50–140 °C as well as the activation energy were determined. The measured data are suitable to be included in severe accident iodine codes such as IMPAIR.

• 2. 

  (2) Steel tubes were exposed to a steam-I2 flow under dry air at T = 120 °C and steam-condensing conditions at T = 120 and 160 °C. In dry air, I2, was retained on the steel surface and a deposition rate constant was measured. Under steam-condensing conditions there is an effective conversion of volatile I2 to non-volatile I− which is subsequently washed off from the steel surface. The I2−T− conversion rate constants suitable for modelling this process were determined. No temperature dependence was found in the range 120–160 °C.

     

Accomplishments and challenges of the severe accident research

This paper briefly describes the progress of the severe accident research since 1980, in terms of the accomplishments made so far and the challenges that remain. Much has been accomplished: many important safety issues have been resolved and consensus is near on some others. However, some of the previously identified safety issues remain as challenges, while some new ones have arisen due to the shift in focus from containment to vessel integrity. New reactor designs have also created some new challenges. In general, the regulatory demands for new reactor designs are stricter, thereby requiring much greater attention to the safety issues concerned with the containment design of the new large reactors, and to the accident management procedures for mitigating the consequences of a severe accident. We apologize for not providing references to many fine investigations that contributed to the great progress made so far in the severe accident research.     

Performance of containment penetrations under severe accident loadings

The paper provides a summary of efforts to date to better understand the leakage behavior of containment penetrations when subjected to severe accident conditions. The research activities discussed herein are a part of the Containment Integrity Programs, which are managed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. Past containment penetration research topics, which are briefly described, include testing of typical compression seals and gaskets, electrical penetration assemblies, and a personnel airlock, as well as an investigation of leakage due to ovalization of penetration sleeves. The primary focus of the paper is on recent or ongoing research programs on the behavior of inflatable seals, bellows, and of pressure unseating equipment hatches.