[en] Under normal operation, nuclear reactor fuel is cooled by water circulating in the primary circuit. In the case of a loss-of-coolant accident, the reactor is stopped automatically. Residual fuel heat must then be evacuated, typically by use of a safety injection system and a reactor containment spray system. These systems are fed with water recovered from the bottom of the containment through sumps. However, because this water may contain debris (insulating material, concrete particles, paint), sumps are equipped with strainers. These strainers may become clogged, preventing emergency coolant recirculation. This could in turn lead to reactor core overheating, or melting in the most extreme circumstances. Participants at the workshop discussed the most recent research and developments in this field, as well as proposed and implemented solutions. These proceedings contain the papers presented at the workshop as well as a summary of the discussions that took place. (author)

[en] The effect of debris generated during a loss of coolant accident (LOCA) on the emergency core cooling system (ECCS) strainers has been studied via numerous avenues over the last several years. The research described in this manuscript examines the generation and effect of secondary materials -- not debris generated in the LOCA itself, but materials created by chemical reactions between exposed surfaces/debris and cooling system water. The secondary materials studied in the research were corrosion products from exposed metallic surfaces and paint chips that may precipitate out of solution, with a focus on the corrosion products of aluminium, iron, and zinc. The processes of corrosion and leaching of metals with subsequent precipitation is important because: (1) the surface area of exposed metal inside containment represents a large potential source term, even for slow chemical reactions; the chemical composition of the cooling system water (boric acid, lithium, etc.) may affect corrosion or precipitation in ways that have not been studied thoroughly in the past; and (3) an eyewitness report of the presence of gelatinous material in the Three Mile Island containment pool after the 1979 accident suggests the formation of a secondary material that has not been examined under the generic safety issue (GSI)-191 research program. This research was limited in scope and consisted only of small-scale tests. Several key questions were investigated: (1) do credible corrosion mechanisms exist for leaching metal ions from bulk solid surfaces or from zinc-based paint chips, and if so, what are the typical rate constants? (2) can corrosion products accumulate in the containment pool water to the extent that they might precipitate as new chemical species at pH and temperatures levels that are relevant to the LOCA accident sequence? and (3) how do chemical precipitants affect the head loss across an existing fibrous debris bed? A full report of the research is available. (authors)

[en] After the Barsebaeck 2 strainer clogging incident from 28 July 1992, a first review of the design features of a Mark I containment and the thermal insulation typically employed revealed a potential for the transportation of larger amounts of insulation into the wet-well (torus) of the containment during a LOCA. Although Switzerland took a quick decision to increase the strainers of all BWRs till the end of 1993 (as it was performed by the Swiss Utilities) for the meantime it was necessary to develop tools for assessing the effectiveness of accident management actions proposed by the utilities for the existing (old strainer) design. Among others tools a simple evaluation model for assessing the transport velocity of insulation debris caused by the suction of operating strainers was developed, which was applied for a BWR with a Mark I containment and can as well be applied for sump pool conditions of a PWR (submerged sump strainers). (author)

[en] When accounting for the total amount of debris that may be present in a pressurised water reactor (PWR) containment pool during operation of the emergency core cooling system (ECCS), it is important to include a reasonable estimate of the latent dirt and foreign material that can be found in containment in addition to the debris generated by a high-pressure pipe rupture. Past and recent testing has shown that even small volumes of fibrous debris present on an ECCS sump screen can very effectively filter particulates that are present in the sump pool, leading to significant pressure losses across the composite debris bed. Debris present during routine operations that is subjected to containment spray and pool transport may contribute a significant source of particulate and perhaps fiber material. Because the PWR industry is working to estimate the quantity of latent debris present in containment, Los Alamos National Laboratory (LANL) is working, under the direction of the United States Nuclear Regulatory Commission (USNRC), to characterise the material composition and the hydraulic flow properties of actual plant debris samples. Beginning in August 2003 and ending in March 2004, this study is expected to quantify particulate and fiber debris parameters, such as the specific surface area and flow porosity that are critical to the proper application of the NUREG 6224 head-loss correlation. Micro-filtering, optical microscopy, and organic dissolution chemistry tests are being used to fractionate the fibrous and particulate components. All tests are being performed at the geochemistry laboratory at the Isotope and Nuclear Chemistry Facility, Chemistry Division (C-INC), LANL, which has the necessary analytic equipment to make direct measurements of the hydraulic flow properties and to handle low-level radioactive PWR latent debris material. The success of this study is dependent on the participation and cooperation of the US PWR industry, the NRC, and LANL. Approximately six volunteer PWR plants are expected to contribute samples collected during their recent condition assessment surveys. (authors)

[en] All Belgian NPPs are PWRs. These PWRs have different containment designs and different sump designs. The four most recent units took RG 1.82 rev. 0 into account for the recirculation sump design. Several improvements were made in the framework of the first Periodic Safety Review to the oldest units such as enlargement of the sumps strainer in 1985, taking into account the requirements of RG 1.82 rev. 0. The problem of recirculation sump screen clogging due to the accumulation of insulation debris after an accident was also identified as a safety issue in the first Periodic Safety Review for Doel 3 and Tihange 2 and the second Periodic Safety Review for Doel 1 and 2 and Tihange 1, that started in the nineties. One of the purposes of the Periodic Safety Review being the justification of the safety level of the power plants with regard to the most recent safety rules and practices, the licensee was asked to review the characteristics of the sump strainers according to the revision 1 of the RG 1.82 (which referred to Nureg 0897 rev. 1). Later on, the Barsebaeck incident and the events leading to the NRC IN 93-34 showed that the conclusions of Nureg 0897 could underestimate the potential for loss of NPSH of the ECCS- and CSS-pumps. The issue was further examined and the AIO asked the licensee to continue efforts to recollect information. After issuance of the results from the parametric evaluation for pressurised water reactor recirculation sump performance (GSI-191) performed by LANL under NRC contract, the AIO asked the Belgian licensee in August 2002 to increase its efforts to investigate and to solve this issue. The requirements to the Licensee and the corresponding action plan as approved by the AIO are presented. Some significant lessons learned from the implementation of this action plan are discussed. (authors)

[en] This executive summary presents the main topics of the workshop on Debris on Emergency Coolant Recirculation, which was held from 25 to 27 February 2004 in Albuquerque. The background of the workshop, the specific purposes of the workshop are detailed, as the scope and technical content and the attendance. This summary provides also a presentation of the main findings and recommendations on future research programs. (A.L.B.)

[en] This report presents an assessment of the operational characteristics of the filtration function used during the recirculation phase of safety injection system (SI) and containment spray system (SS) in the event of a break of the primary system in the containment for the French pressurised reactors (58 reactors), which have been designed according with the Regulatory Guide 1.82 (revision 1) published in 1985. In spite of the lessons learned from the Barsebaeck accident occurred in 1992 and the corresponding questioning about the appropriate character of the requirements of this Regulatory Guide, Electricite de France has always proclaimed the compliance of the French plants with revision 1 of the Regulatory Guide 1.82 and has refused to review this position. Consequently, the 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) has decided to perform its own assessment. A general overview of the literature has been conducted between October 1999 and November 2000, questioning the capability of the safeguards systems to operate during recirculation phase which resulted in defining an approach methodology and writing technical specifications related to the French design reactors. After this general overview, the 'Institut de Radioprotection et de Surete Nucleaire' has decided to perform a detailed study on the risk of sump plugging for the French reactors. (author)

[en] The potential for LOCA-generated debris to degrade PWR ECCS performance during recirculation has gained international industry and regulatory attention. Many US PWR owners are still in the early stages of determining the significance of this issue for their particular plants. For a sizable number of these plants, resolution of the ECCS debris blockage issue will require a combination of analytical evaluation, changes to operating or maintenance protocols, and modifications or additions to some plant components. This paper provides an overview of site-specific ECCS debris blockage solutions anticipated to be used by US PWRs. The overview begins with a brief review of the variability in containment floor and ECCS sump configurations found in the US PWR fleet. With this variability in mind, discussion then focuses on the several different resolution modes that may be brought to bear on the ECCS debris blockage issue. Finally, a variety of factors that owners must consider in order to select an optimal means of resolution are surveyed. (authors)

[en] In the framework of research of researches related to the PWR sump clogging in Usa, the author presents the history of GSI-191 (assessment of debris accumulation on PWR sump performance), the research to date (technical assessment, regulatory guide and evaluation guidance, model validation), the current and planned tests (chemical effect and calcium silicate tests, latent debris and downstream effect tests, integrated chemical effect tests, EPRI coatings study). (A.L.B.)

[en] The United States Nuclear Regulatory Commission (USNRC) currently holds the position that 100 percent of design basis accident (DBA) unqualified coating materials located within a pressurised water reactor (PWR) containment will fail (dis-bond) during a DBA (e.g. loss of coolant, main steam line break) and may contribute to the emergency core cooling system (ECCS) sump debris source term. Electrical cabinets, small cranes, electric motors, pipe support components, and other miscellaneous equipment installed within US PWR containments are often coated by the original equipment manufacturers (OEM) using DBA unqualified coating materials (usually a standard shop oil based alkyd system). Little or no documented DBA test data currently exists concerning these OEM coatings. In support of the US industry efforts to resolve Generic Safety Issue 191, EPRI International, PSE Division (EPRI PSE) and the Nuclear Utility Coating Council (NUCC) are jointly conducting research to investigate the actual effect of PWR DBA environmental exposure on OEM coatings applied to components installed in US PWR containments. This paper will present details concerning the study. The EPRI/NUCC study, entitled, 'Unqualified OEM Coatings Testing', will be conducted during 2004 in four major steps: 1. conduct an industry-wide survey to determine which components in US PWR containments are normally coated with DBA-unqualified OEM coatings; 2. determine which of the components identified in Step 1 have been previously EQ tested, and review available EQ test data to determine the performance of the OEM coatings when exposed to DBA environments; 3. determine which of the components identified in Step 1 have not been previously EQ tested, and perform DBA testing on samples cut from actual OEM components in accordance with ASTM D3911-03 and ASTM D4082-02; and 4. publish the results of Steps 2 and 3 in two separate reports. The EPRI/NUCC study plan is described in this paper. Since the actual study will not be completed until mid-2004, updated information concerning the study progress will be presented at the February 2004 OECD conference. (authors)