Seven papers were presented in Session 1. Taking benefit of the lessons from the Barsebäck accident, many countries have improved their units since 1992. In Sweden, new strainers were developed for PWR installations which included large sacrificial strainers and self-cleaning “wing-strainer” to provide robust debris handling. The Canadian Nuclear Safety Commission (CNSC) and Canadian industry worked closely together to solve the strainer issue. AECL performed extensive tests and developed finned strainers to provide added strainer areas. Despite the issue of an official recommendation in 1998, German technical support organisations and utilities performed further experiments to demonstrate the function of the sump ...

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 Barsebäck accident occurred in 1992 and the corresponding questioning about the appropriate character of the requirements of this ...

All Belgian NPPs are PWRs. These PWRs have different containment designs and different sump designs. The four most recent units (Doel 3 &4 and Tihange 2 &3) 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 (Doel 1 & 2 and Tihange 1) 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&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 ...

ÜV Süddeutschland, located in Munich/Bavaria in the southern part of Germany, is a technical support organisation (TSO) in conventional and nuclear technology in Germany. It is an organisation of independent experts. The nuclear division of this organisation supports the Bavarian and Hessian state authorities in charge of nuclear facilities. The task of the TÜV’s nuclear division is to deliver expert’s opinions on licensing and supervisory procedures. To fulfill this task corresponding to up to date knowledge (so called “state of science and technology” within the German framework of rules ...

When the Canadian Nuclear Safety Commission (CNSC) became aware of concerns relating to the collection of debris at suction strainers for emergency core cooling (ECC) systems following the incident in Barsebäck, Sweden, it issued a notice to Canadian utilities requiring them to review their ECC strainer capability in view of the potential increase in pressure drop, and address any deficiencies. During this review period, a number of uncertainties were identified as needing resolution, principally those items directly related to the head loss across the strainer (both in the short and long term), as well as a few secondary issues such as the likelihood of air ingestion. Canadian utilities contracted Atomic Energy of Canada Limited (AECL), through the CANDU owners group (COG), to perform extensive fundamental testing to establish the important parameters governing ECC strainer performance. AECL expanded upon this knowledge base with additional tests to confirm proposed designs for specific applications. As a result of all the testing, a substantial body of ...

NRC regulations in the US Code of Federal Regulations require that the emergency core cooling systems in a nuclear power plant must provide the capability for long-term cooling of the reactor core. As set forth in 10 CFR 50.46(b)(5), the emergency core cooling systems must have the capability to remove decay heat so that the core temperature is maintained at an acceptably low value for the extended period of time required by the long-lived radioactivity remaining in the core. For US plants that are licensed to the General Design Criteria in Appendix A to 10 CFR Part 50, General Design Criterion 35 specifies additional emergency core cooling system requirements, such as fuel and clad damage that could interfere with continued effective core cooling. Similarly, for plants licensed to the General Design Criteria, General Design Criterion 38 provides requirements for containment heat removal systems, and General Design Criterion 41 provides requirements for containment atmosphere cleanup. Many licensees credit a containment ...

Outline: History of GSI-191; Research to date: Technical assessment; Regulatory guide and evaluation guidance; Model validation; Current and planned tests: Chemical effect and calcium silicate tests; Latent debris and downstream effect tests; Integrated chemical effect tests; EPRI coatings study...

The paper describes briefly activities performed by Vattenfall Utveckling AB at Älvkarleby Laboratory as part of the qualification programme for the new ECCS strainers at PWR plant Ringhals 2 based on the “robust solution” with large sacrificial strainers and a self-cleaning “wing-strainer” of same type as used for the five modified Swedish BWR plants. With the new knowledge gained from several BWR strainer projects following the Barsebäck strainer incident in 1992, the functioning of ECCS strainers for PWR was re-evaluated. The upgrading at Ringhals 2, a 3-loop Westinghouse plant having fibreglass and mineral wool as insulation, was the ...

The 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 primary system break in the containment, has been performed by the “Institut de radioprotection et de sûreté nucléaire” (IRSN) for the French pressurised reactors (58 reactors). Those one have been designed according with the RG 1.82 (rev. 1). The IRSN has focused in particular on the CPY series, 900 MWe 3 loops pressurised water reactors (28 reactors). A general overview of the literature has been conducted between October 1999 and November 2000, which resulting in defining of an approach methodology and the writing of technical ...

Extensive investigation and tests on this issue were performed in Germany, and the conclusion was that no major backfitting was necessary on their Siemens PWRs. This was largely due to the use of strong cassette type insulation only on the primary system; special efforts to ensure break preclusion for the main coolant lines thus allowing application of reduced break sizes in their debris generation calculations; the fact that no containment spray systems exist in these plants; high water levels in the sump allowing increased sedimentation; and the enforcement of containment cleanliness after refuelling. The general sense of the session was that because of the different containment designs and ...

Assessment of the operational characteristics of the filtration function used during the recirculation phase of safety injection system (ECCS) and containment spray system (CSS) in the event of a break of the primary system in the containment has been performed by the “Institut de radioprotection et de sûreté nucléaire” [1] for the French pressurised reactors (58 reactors) in particular the CPY series, 900 MWe pressurised water reactors (30 reactors) designed according Regulatory Guide 1.82 (Rev. 1). To estimate the risk involved, the following points were studied: ...

The Canadian nuclear industry, including Atomic Energy of Canada Limited (AECL) and the four nuclear utilities (New Brunswick Power, Hydro-Québec, Ontario Power Generation and Bruce Power) have been heavily involved in strainer clogging issues since the late 1990s. A substantial knowledge base has been obtained with support from various organisations, including the CANDU Owners Group (COG), AECL and the CANDU utilities. Work has included debris assessments at specific stations, debris characterisation, transport, head loss measurements across strainers, head loss models and investigations into paints and coatings. Much of this work was performed at AECL’s Chalk River Laboratories and has been used to customise strainer solutions for several CANDU (PWR-type) stations. This paper summarises the CANDU experience, describing problems ...

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 ...

Experimental and analytical results are summarised from experiments sponsored by the United States (US) Nuclear Regulatory Commission (NRC) and performed under the direction of Los Alamos National Laboratory (LANL) in facilities operated by the Civil Engineering Department at the University of New Mexico (UNM). The study generated data needed to support the resolution of Generic Safety Issue 191, which addresses debris accumulation on the pressurised water reactor (PWR) sump screen and the consequent loss of the emergency core cooling-system pump’s net positive suction head. These experiments investigated: (1) the accumulation of debris on a screen typical of those found in the PWR plants; and (2) the subsequent head loss associated with debris from calcium silicate (CalSil) insulation. Both of these investigations are key elements in the resolution of ...

The paper includes the description of separate effect test facilities used for investigations with regard to the fragmentation and the transport behaviour of different insulation materials in multidimensional aqueous flow. The instrumentation of the rigs is specified, in particular modern digital image processing technologies. First experimental results are shown and discussed generated at three acrylic glass test facilities. The experimental data could use for CFD-modelling and validation.

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; (2) the chemical composition of the cooling system water (boric acid, lithium, etc.) may affect ...

Several series of experiments investigating the post-LOCA behaviour of fragmented insulant material (mineral wool) were carried out and evaluated in the period between 2001 and 2003. The experiments concentrated on: Fragmentation of mineral wool under as realistic conditions as possible; this was performed by exposing stainless steel-clad mineral wool panels as used in German nuclear power plants to a jet of saturated water (DN 200, approx. 290°C, approx. 110 bar). ...

Session 3 comprised six papers; four of them were presented in the workshop. The main topics dealt with the debris transport in water, the debris impact on pump performance, and break characterisation (break size and location) of pipes using fracture mechanics methods to determine the debris source term. The approaches to investigate water-borne debris transport were the following: calculate debris generation from the break location by using basic hydraulic equations, use computational fluid ...

After the Barsebäck 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 wetwell (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 effectivity 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 ...

The investigation of insulation debris generation, transport and sedimentation gains importance regarding the reactor safety research for PWR and BWR considering the long term behaviour of emergency core coolant systems during all types of LOCA. The insulation debris released near the break during LOCA consists of a mixture of very different particles concerning size, shape, consistence and other properties. Some fraction of the released insulation debris will be transported into the reactor sump where it may affect emergency core cooling. Open questions of generic interest are e.g. the sedimentation of the insulation debris in a water pool, possible resuspension, transport in ...

A study sponsored by the United States (US) Nuclear Regulatory Commission (NRC) was performed to reassess the effects of ingesting loss of coolant accident (LOCA) generated materials into emergency core cooling system (ECCS) pumps and the subsequent impact of this debris on the pumps’ ability to provide long-term cooling to the reactor core.1 ECCS intake systems have been designed to screen out large post-LOCA debris materials. However, small-sized debris can penetrate these intake strainers or screens and reach critical pump components. Prior NRC-sponsored evaluations of possible debris and gas ingestion into ECCS pumps and attendant impacts on pump performance were performed in the early 1980s.2 The earlier study focused primarily on pressurised water reactor (PWR) ECCS pumps. This issue was revisited both to factor in our improved knowledge ...

In 1997 the United States Nuclear Regulatory Commission (NRC) initiated an investigation into the possibility of failure of the recirculation system in nuclear power plants (Generic Safety Issue GSI-191). If a loss of coolant accident (LOCA) were to occur within the containment of a pressurised water reactor, piping thermal insulation and other materials in the vicinity of the break would be dislodged by break jet impingement.1,2 Some of this debris could eventually be transported by the recirculating water and accumulate on the suction sump screens, clogging them and causing the cooling system to fail. The NRC initiated a test programme to investigate the amount of insulation ...

An important safety concern regarding long-term recirculation cooling following a loss of coolant accident (LOCA) is the transport of debris materials to interceptors (i.e. trash racks, debris screens, suction strainers) inside containment and the potential for debris accumulation to result in adverse blockage effects. Debris resulting from a LOCA, together with pre-existing debris, could block the emergency core cooling system (ECCS) debris interceptors and result in degradation or loss of recirculation flow margin. Potential debris sources can be divided into three categories: (1) debris that is generated by the LOCA and is transported by blowdown forces (e.g. insulation, paint); (2) debris that is generated or transported by washdown; and (3) other debris that existed before a LOCA (dust, sand, etc.). Each debris source is separately evaluated to estimate the quantity and other ...

In the event of a loss of coolant accident (LOCA) within containment of a pressurised water reactor (PWR) there is the potential for the generation of debris with the attendant concern of containment sump screen blockage. The debris, consisting piping or equipment insulation, protective coatings or paints, concrete dust or general containment housekeeping materials, may be transported to the containment sump during the recirculation phase of emergency core cooling system (ECCS) and containment spray system (CSS) operations. Unresolved Safety Issue (USI) A-43, “Containment emergency sump performance” had been ...

Five papers were presented in this session which described industry solutions for this issue in Belgium, Switzerland and the US. Two kinds of presentations were given. The first three presentations were from utilities or their engineering support that were studying the sump clogging issue, looking for solutions. The other two presentations concerned researches conducted by the industry, focusing on specific topics related to the debris source term (coatings and insulation materials). ..

The emergency core cooling system (ECCS) of a nuclear power plant supplies cooling water to the reactor vessel in the case of a loss of coolant accident (LOCA). A LOCA generates debris by the force of coolant impinging upon pipe insulation and entraining a wide variety of particulate matter from the reactor building surfaces that the coolant flows over. During the recirculation phase following a LOCA, if a sufficient quantity of debris accumulates on the sump screens, the ECCS pumps’ suction flow path can be reduced significantly, causing a drop in the available NPSH and, eventually, a loss of pump flow. If the ECCS flow is lost for a sufficiently long time, the core may become uncovered and overheat, causing severe damage to the fuel. Since the Barsebäck strainer event (July 1992), studies and experiments are being undertaken all ...

After the Barsebäck 2 strainer clogging incident from 28 July 1992, a short term and a long term action programme was initiated by the Swiss Federal Nuclear Safety Inspectorate to resolve the strainer clogging issue [1]. The issue was closed in 1994 based on implemented modifications of suction strainers (flow area largely increased by a factor of 7 to 30) for BWRs and the approval of the strainer design based on plant specific analysis for PWRs. The key parameters used in the plant specific analysis are still surveyed within the plant specific ageing programme. The work presented ...

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 (disbond) 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 [1]. 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 ...

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 ...

Pressure drop calculations across a LOCA induced fibrous debris bed have been successfully demonstrated to be accurate using the NUREG/CR-6224 semi-theoretical head loss correlation. One of the critical parameters needed for the NURE/CR-6224 correlation to predict the pressure drop across a fibrous debris bed are the characteristics of the debris constituents (density and characteristic size). This paper provides a brief description of the NUREG/CR-6224 head loss correlation and presents suggested debris characteristics of typical sources of debris found in American nuclear power plants...