NRC Generic Letter 1998-04
UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D.C. 20555-0001July 14, 1998NRC GENERIC LETTER 98-04:POTENTIAL FOR DEGRADATION OF THE EMERGENCYCORE COOLING SYSTEM AND THE CONTAINMENTSPRAY SYSTEM AFTER A LOSS-OF-COOLANTACCIDENT BECAUSE OF CONSTRUCTION ANDPROTECTIVE COATING DEFICIENCIES AND FOREIGNMATERIAL IN CONTAINMENT -All holders of operating licenses for nuclear power reactors, except those who havepermanently ceased operations and have certified that fuel has been permanently removedfrom the reactor vessel.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter for severalreasons. It alerts addressees that foreign material continues to be found inside operatingnuclear power plant containments. During a design basis loss-of-coolant accident (DB LOCA),this foreign material could block an emergency core cooling system (ECCS) or safety-relatedcontainment spray system (CSS) flow path or damage ECCS or safety-related CSS equipment.In addition, construction deficiencies and problems with the material condition of ECCSsystems, structures, and components (SSCs) inside the containment continue to be found.Design deficiencies also have been found which could degrade the ECCS or safety-relatedCSS. No action or information Is requested regarding these issues. The NRC has Issued manyprevious generic communications on this subject, as discussed later In this generic leter, andassumes that addressees have had adequate prior notice to consider possible actions at theirfacilities to address these concerns.The NRC expects addressees to ensure that the ECCS and the safety-related CSS remaincapable of performing their intended safety functions. Due to the importance of these systems,the NRC may conduct Inspections to ensure compliance.The NRC is also Issuing this generic letter to alert the addressees to the problems associatedwith the material condition of Service Level I (see definitions of Service Levels In Attachment 3)protective coatings Inside the containment and to request information under 10 CFR 50.54(f) toevaluate the addressees' programs for ensuring that Service Level I protective coatings Insidecontainment do not detach from their substrate during a DB LOCA and Interfere with the87129 ltDIL fRficeJ 05000X3 , 980'V 1 on 0flqt S- T-' ,'7 419Q' -t~- t42 L 98-04July 14, 1998 operation of the ECCS and the safety-related CSS. The NRC intends to use this information toassess whether current regulatory requirements are being correctly implemented and whetherthey should be revised.BackgroundForeign Materal Exclusion, Construction Deficiencies, and Design DeficienciesIn some recent events (discussed in Attachment 1 to this generic letter), foreign material whichcould have affected the operation of the ECCS was discovered inside the containment. As partof its review of these events, the NRC staff found several types of continuing problems.(1) Foreign material has been found In areas of the containment where It could betransported to the sump(s) or the suppression pool and potentially affect the operationof the ECCS or safety-related CSS. Such material has also been found inpressurized- water reactor (PWR) sumps, In boiling-water reactor (BWR) suppressionpools and downcomers, and in safety-related pumps and piping.(2) Deficiencies have been found in the construction of the ECCS sumps and strainers.These deficiencies, which could have Impaired the operation of the ECCS or thesafety-related CSS, Include missing screens, unintended openings in screens, andincorrectly sized screens.(3) Problems have also been found with the material condition of sumps and suctionstrainers. These problems, potentially Impairing the operation of the ECCS or safety-related CSS, Include deformed suction strainers and unintentional flow paths createdby missing grout.(4) Design deficiencies have been found, including flow line valves with clearancessmaller than the sump screen mesh size and strainers with a flow area smaller thanrequired.(5) There have been two incidents, described in licensee event reports (LERs), In whichdoors to emergency sump structures were left open when ECCS and safety-relatedCSS operability was required by the technical specifications.The Discussion section of this generic letter describes the regulatory and safety bases for theseconcerns.A more complete list of the above events is provided In Attachment 2. As discussed InAttachment 1, almost all of these events have been the subject of previous NRC genericcommunications and LERs. Apparently, past NRC generic communications have not beencompletely effective In focusing licensee attention to the potential areas of concern to controlthese problems. Nevertheless, the NRC expects that licensees will ensure that the ECCS andsafety-related CSS remain capable of performing their intended safety functions. The NRCplans to further emphasize this Issue by conducting inspections to ensure compliance with GL 98-04July 14, 1998 existing plant licensing bases. The NRC intends to take enforcement action for discoveredinadequacies consistent with NRC Enforcement Policy.Protective CoatingsProtective coatings Inside nuclear power plant containments serve three general purposes.Protective coatings are applied to carbon and low alloy steel and, less commonly, to aluminumand galvanized surfaces to control corrosion, control radioactive contamination levels, and toprotect surfaces from wear. (Although aluminum and galvanized surfaces are not commonlycoated, nothing in NRC requirements or industry standards prevent these surfaces from beingcoated.) A discussion on protective coatings inside the containment and the regulatoryrequirements and guidance for their use are discussed In Attachment 3.It has been assumed that qualified protective coatings are capable of adhering to their substrateduring a DB LOCA in order to minimize the amount of material which can reach the emergencysump screens or suction strainers and dog them. The NRC is aware that not all coatings Insidethe containment are qualified and, therefore, the amount of unqualified coatings must becontrolled since the unqualified coatings are assumed to detach from their substrates during aDB LOCA and may be transported to the emergency sump screens or suction strainers. Oncein contact with sump screens or suction strainers, coating chips may adversely impact the netpositive suction head (NPSH) available to the ECCS or CSS pumps.Additionally, In some cases, coatings which were qualified failed during normal operation. Someof these events are discussed in Attachment 4.DiscussionNRC regulations in 10 CFR 50.46 require that licensees design their ECCS to provide long-termcooling capability so that the core temperature can be maintained at an acceptably low valueand decay heat can be removed for the extended period required by the long-lived radioactivityremaining in the core. Licensees are required to demonstrate this capability while assuming themost conservative single failure. Some addressees may credit CSSs in the licensing basis forradioactive-source-term and pressure reduction.Foreign materials, degraded coatings Inside the containment that detach from their substrate,and ECCS components not consistent with their design basis, along with LOCA-generateddebris, are potential common-cause failure mechanisms which may clog suction strainers,sump screens, filters, nozzles, and small-clearance flow paths In the ECCS and safety-relatedCSS and thereby interfere with the long-term cooling function, source-term and pressurereduction features of plant design.Qualified coatings used inside containment should be capable of withstanding theenvironmental conditions of a postulated DB LOCA. Although small, localized areas ofdegraded coatings may not be Indicative of widespread failure of the coatings, the condition ofthe coatings should be evaluated by suitable means. The Electric Power Research Institute(EPRI) has prepared a guidance document for containment coatings titled Guidelines on theElements of a Nuclear Safety-Related Coatings Program." Licensees may find EPRI
GL 98-04July 14, 1998 TR-109937, Final Report, dated April 1998 useful when evaluating coatings, although it has notbeen endorsed by the staff. The LERs and NRC inspection reports described in Attachment 4to this generic letter provide evidence of weaknesses In addressee programs with regard toapplications of protective coatings for Service Level 1. These weaknesses Include deficienciesin addressee programs to: (1) control the preparation and cleanliness of the substrate beforethe coatings are applied, (2) control the preparation of a coating before its application,(3) control the dry film thickness of coatings applied to the substrate, (4) monitor for, and controlthe use of, excessive amounts of unqualified coatings inside the containment, (5) monitor thestatus of "qualified" coatings already applied to the surfaces of the containment structure and toother equipment Inside the containment, and (6) assess the safety significance of coatingsinside containment that have been determined to detach from their substrate and to repairthese coatings If necessary.The NRC has Issued a number of generic communications on problems with protectivecoatings, and the potential for the loss of the ECCS and safety-related CSS as a result ofstrainer clogging and debris blockage. These generic communications are listed InAttachment 5. They apply to both PWRs and BWRs. The events discussed in these genericcommunications, and similar events described in LERs and NRC inspection reports,demonstrate the need for a strong foreign material exclusion (FME) program in all areas ofPWRs and BWRs that may contain materials that could interfere with the successful operationof the ECCS and the safety-related CSS. Other events demonstrate the need to ensure thecorrect design and to maintain the material condition of emergency core cooling system andsafety-related containment spray system SSCs, Including the suppression pools, ECCSstrainers and sumps, and the protective coatings inside containment.Applicable Regulatory RequirementsThe requirements of 10 CFR 50.46(b)(5) and Appendix A to 10 CFR Part 50, Criterion 35,address long term cooling capability and emergency core cooling, respectively. The NRC staffconsiders that the requirements of 10 CFR Part 50, Appendix B, are germane to this issue forsafety related containment coatings.Section 50.65 of 10 CFR, "Requirements for monitoring the effectiveness of maintenance atnuclear power plants," (maintenance rule) includes In Its scope all safety-related SSCs andthose non-safety-related SSCs that fall Into the following categories: (1) those that are reliedupon to mitigate accidents or transients or are used In plant emergency operating procedures,(2) those whose failure could prevent safety-related SSCs from fulfilling their safety-relatedfunction, and (3) those whose failure could cause a reactor scram or an actuation of a safety-related system.The PWR sumps and BWR strainers are included within the scope of the maintenance rule.To the extent that protective coatings meet these scoping criteria, they are within the scope ofthe maintenance rule. The maintenance rule requires that licensees monitor the effectivenessof maintenance for these protective coatings (as discrete systems or components or as part ofany SSC) in accordance with paragraph (a)(1) or (a)(2) of 10 CFR 50.65, as appropriate.
GL 98-04July 14, 1998 Although this generic letter concerns coatings within the containment and requests informationabout coatings within containment, addressees should ensure that all coatings which meet themaintenance rule scoping criteria are Included in the programs and procedures forimplementing the maintenance rule.The NRC has conducted numerous inspections In the areas addressed by this generic letter; forexample, the NRC Issued Temporary Instruction (TI) 2515/125, Foreign Material ExclusionControls," on August 25, 1994. Violations discovered during the TI 2515/125 Inspections havebeen identified and appropriate enforcement action has been taken in accordance with theNRC's Enforcement Policy (NUREG-1600, "General Statement of Policy and Procedures forNRC Enforcement Actions: Enforcement Policy").The NRC will continue to conduct inspections In these areas and will consider the long historyof generic communications on the Issues addressed by this generic letter as prior notice tolicensees when assessing civil penalties In accordance with Section VI.B.2 of the EnforcementPolicy. Finally, notwithstanding the normal civil penalty assessment, the NRC will considerwhether the circumstances of the case warrant escalation of enforcement sanctions inaccordance with Section VII.A.1 of the Enforcement Policy.If, in the course of assessing the effectiveness of the plant-specific FME program or preparing aresponse to the required Information, an addressee determines that its facility Is not incompliance with the Commission's requirements, the addressee is expected to take appropriateactions In accordance with both requirements stated In Appendix B to 10 CFR Part 50 and theplant technical specifications to restore the facility to compliance.Licensees are encouraged to work closely with their owners groups and industry associations tocoordinate the responses to this letter to improve the efficiency of the responses. TheInformation submitted in response to this generic letter should be considered to be publicinformation.Required InformationAs a result of NRC findings In these areas and due to the importance of ensuring systemfunctionality, within 120 days of the date of this generic letter, addressees are required tosubmit a written response that Includes the following Information:(1) A summary description of the plant-specific program or programs Implemented toensure that Service Level I protective coatings used inside the containment areprocured, applied, and maintained In compliance with applicable regulatoryrequirements and the plant-specific licensing basis for the facility. Include adiscussion of how the plant-specific program meets the applicable criteria of 10 CFRPart 50, Appendix B, as well as Information regarding any applicable standards, plant-specific procedures, or other guidance used for (a) controlling the procurement ofcoatings and paints used at the facility, (b) the qualification testing of protectivecoatings, and (c) surface preparation, application, surveillance, and maintenanceactivities for protective coatings. Maintenance activities Involve reworking degradedcoatings, removing degraded coatings to sound coatings, correctly preparing thesurfaces, applying new coatings, and verifying the quality of the coatings.
3L 98-04July 14, 1998 (2) Information demonstrating compliance with item (I) or Item (ii):(i) For plants with licensing-basis requirements for tracking the amount of unqualifiedcoatings inside the containment and for assessing the impact of potential coatingdebris on the operation of safety-related SSCs during a postulated DB LOCA, thefollowing information shall be provided to demonstrate compliance:(a) The date and findings of the last assessment of coatings, and the planneddate of the next assessment of coatings.(b) The limit for the amount of unqualified protective coatings allowed in thecontainment and how this limit is determined. Discuss any conservatism Inthe method used to determine this limit.(c) If a commercial-grade dedication program is being used at your facility fordedicating commercial-grade coatings for Service Level I applications insidethe containment, discuss how the program adequately qualifies such a coatingfor Service Level 1 service. Identify which standards or other guidance arecurrently being used to dedicate containment coatings at your facility; or,(ii) For plants without the above licensing-basis requirements, information shall beprovided to demonstrate compliance with the requirements of 10 CFR 50.46b(5),"Long-term cooling" and the functional capability of the safety-related CSS as setforth in your licensing basis. If a licensee can demonstrate this compliancewithout quantifying the amount of unqualified coatings, this Is acceptable. Thefollowing information shall be provided:(a) If commercial-grade coatings are being used at your facility for Service Level 1applications, and such coatings are not dedicated or controlled under yourAppendix B Quality Assurance Program, provide the regulatory and safetybasis for not controlling these coatings in accordance with such a program.Additionally, explain why the facility's licensing basis does not require such aprogram.Address the required written Information to the U.S. Nuclear Regulatory Commission, ATTN:Document Control Desk, Washington, DC 20555-0001, under oath or affirmation pursuant toSection 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). Thisinformation will enable the Commission to determine whether a license should be modified,suspended, or revoked. In addition, submit a copy of the written Information to the appropriateregional administrator.
Backfit Discussion
This generic letter requires information from the addressees under the provisions ofSection 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). Thisgeneric letter does not constitute a backfit as defined in 10 CFR 50.109(a)(1) since It does notimpose modifications or additions to systems, structures, and components or to the design oroperation of an addressee's facility. It also does not it Impose an Interpretation of the GL 98-04July 14, 1998 Commission's rules that is either new or different from a previous staff position. The staff,therefore, has not performed a backfit analysis.Reasons for Information RequirementThis generic letter transmits a requirement to submit information pursuant to the provisions ofSection 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(0 for thepurpose of verifying compliance with applicable regulatory requirements. The requiredInformation will enable the NRC staff to determine whether the addressees' protective coatingsInside the containment comply and conform with the current licensing basis for their facilitiesand whether the regulatory requirements pursuant to 10 CFR 50.46 are being met.Protective coatings are necessary inside containment to control radioactive contamination andto protect surfaces from erosion and corrosion. Detachment of the coatings from the substratemay make the ECCS unable to satisfy the requirement of 10 CFR 50.46(b)(5) to provide long-term cooling and may make the safety-related CSS unable to satisfy the plant-specific licensingbasis of controlling containment pressure and radioactivity following a LOCA.
Paperwork Reduction Act Statement
This generic letter mandates information collections that are subject to the PaperworkReduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approvedby the Office of Management and Budget, approval number 3150-0011, which expires onSeptember 30, 2000.The public reporting burden for this collection of information is estimated to average 400 hour0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />sper response, including the time for reviewing Instructions, searching existing data sources,gathering and maintaining the data needed, and completing and reviewing the collection ofinformation. The NRC Is seeking public comment on the potential Impact of the collection ofinformation requested in the generic letter and on the following issues:(1) Is the proposed collection of Information necessary for the proper performance of thefunctions of the NRC, and will the Information have practical utility?(2) Is the estimate of burden accurate?(3) Is there a way to enhance the quality, utility, and clarity of the information to becollected?(4) How can the burden of the collection of Information be minimized? Can automatedcollection techniques be used?Send comments on any aspect of this collection of information, Including suggestions forreducing this burden, to the Information and Records Management Branch, T-6F33, U.S.Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Intemet electronic mail atBJS1 @NRC.GOV: and to the Desk Officer, Office of Information and Regulatory Affairs,NEOB-1 0202 (3150-0011), Office of Management and Budget, Washington, DC 20503.
-GL 98-04July 14, 1998 The NRC may not conduct or sponsor, and a person is not required to respond to, a collectionof information unless it displays a currently valid OMB control number.If you have any questions about this matter, please contact one of the technical contacts or thelead project manager listed below, or the appropriate Office of Nuclear Reactor Regulationproject manager.4k W. Roe, Acting Directorbi'ision of Reactor Program ManagementOffice of Nuclear Reactor RegulationTechnical contacts: James Davis, NRR301- 415-2713E-mail: Jadnrc.govRichard Lobel, NRR301-415-2865E-mail: rml@nrc.govLarry L. Campbell301-415-2976E-mail: llc3@nrc.govLead Project Manager: John Hickman, NRR301-415-3017E-mail: jbhinrc.govAttachments:1. ECCS sump and strainer events Involving foreign material inside the containment andconstruction and design deficiencies2. Operational events Involving debris In ECCS recirculation flow paths3. Background on regulatory basis for protective coatings4. Chronology of incidents and activities related to protective coatings5. Generic communications Issued by the NRC on ECCS and safety-related CSS sump andstrainer blockage6. List of recently Issued Generic Letters Attachment 1GL 98-04July 14, -1998 ECCS SUMP AND STRAINER EVENTS INVOLVING FOREIGN MATERIAL INSIDE THECONTAINMENT AND CONSTRUCTION AND DESIGN DEFICIENCIESOn November 16, 1988, the NRC issued Information Notice (IN) 88-87, "Pump Wear andForeign Objects in Plant Piping Systems," concerning several Incidents in which the potentialexisted for a flow reduction as a result of pump wear and foreign objects In plant pipingsystems. In one of these Incidents, the licensee found foreign objects In a temporary pumpdischarge cone strainer. The licensee Investigated further and found foreign objects, dating toearly construction modifications, in the sump. In addition, various deficiencies were found in thesump screens.On November 21, 1989, the NRC issued IN 89-77, "Debris in Containment Emergency Sumpsand Incorrect Screen Configurations," which discussed loose parts and debris in thecontainment sumps of three pressurized-water reactors (PWRs), Surry Units 1 and 2 andTrojan. At Surry Units 1 and 2, some of the debris was large enough to cause pump damage orflow degradation. In addition, some of the screens had gaps large enough to allow additionalloose material to enter the sump. The licensee found that screens that separate the redundanttrains of the recirculation spray system were missing at both units. At Trojan, the licenseediscovered debris in the sump. Some debris was found after containment closeout. In addition,still later, before startup, the NRC identified missing portions of the sump top screen and innerscreen. IN 89-77 also reported that In 1980 the Trojan licensee found a welding rod jammedbetween the impeller and the casing ring of a residual heat removal (RHR) pump.On December 23, 1992, the NRC issued IN 9245, "Potential Failures of Emergency CoreCooling Systems Caused by Foreign Material Blockage," which alerted licensees to events attwo PWRs. In these events, foreign material blocked flow paths within the ECCS safetyinjection and containment spray pumps so that the pumps could not produce adequate flow.On April 26, 1993, and May 6, 1993, the NRC Issued IN 93-34, Potential for Loss ofEmergency Cooling Function Due to a Combination of Operational and Post-LOCA Debris inContainment," and its supplement. In these information notices, the NRC described severalinstances of clogged ECCS pump strainers, including two events at the Perry Nuclear PowerPlant, a domestic boiling-water reactor (BWR). In the first Perry event, residual heat removal(RHR) strainers were clogged by operational debris consisting of "general maintenance-typematerial and a coating of fine dirt.' After cleaning the strainers in January 1993, the licenseediscovered that RHR A and B strainers were deformed. The strainers were replaced. Thesecond Perry event Involved an RHR pump test which was run after a plant transient In March1993. Pump suction pressure dropped to 0 KPa (0 psig). No change In pump flow rate wasobserved. Material found on the strainer screen was analyzed and found to consist of glassfibers from temporary drywell cooling filters that had been inadvertently dropped Into thesuppression pool and corrosion products that had been filtered from the pool by the glass fibersadhering to the surface of the strainer. The material significantly Increased the pressure dropacross the strainer.
Attachment IGL 98-04July 14, 1998 In response to these two events, the licensee for Perry increased the suction strainer area,provided suction strainer backflush capability, and Improved measures to keep the suppressionpool clean.On May 11, 1993, the NRC issued Bulletin 93-02, Debris Plugging of Emergency Core CoolingSuction Strainers," which requested that both PWR and BWR addressees (1) identify fibrous airfilters and other temporary sources of fibrous material in containment not designed to withstanda loss-of-coolant accident (LOCA) and (2) take prompt action to remove the foreign matter andensure the functional capability of the ECCS. All addressees have responded to the bulletin,and the NRC staff has completed Its review of their responses.The licensee for Arkansas Nuclear One, Unit 2, reported by Licensee Event Report (LER) 93-002-00, dated November 22, 1993, that the containment sump integrity was inadequate to keepforeign material out. Holes in the masonry grout below the sump screen assembly would havelet water into the sump without being screened. The licensee attributed this condition to failureto implement design basis requirements for the sump during initial plant construction. Theholes were difficult to detect. The holes appeared to be part of the design because of theiruniform spacing and because they were "somewhat recessed...such that to see the holes theymust be viewed from near the floor or from a significant distance away from the sump."On August 12, 1994, the NRC issued IN 94-57, 'Debris in Containment and the Residual HeatRemoval System," which alerted operating reactor licensees to additional instances ofdegradation of ECCS components because of debris. At River Bend Station, the licenseefound a plastic bag on an RHR suction strainer. At Quad Cities Station, Unit 1, on July 14,1994, the remains of a plastic bag were found shredded and caught within the anti-cavitationtrim of an RHR test return valve. Subsequent to that event at Quad Cities, Unit 1, the licenseeobserved reduced flow from the C RHR pump and, upon further investigation, found a 10-cm(4-in.) diameter wire brush wheel and a piece of metal wrapped around a vane of the pump.On January 25, 1995, the NRC issued IN 95-06, 'Potential Blockage of Safety-RelatedStrainers by Material Brought Inside Containment," which discussed a concern that plastic orfibrous material, brought inside the containment to reduce the spread of loose contamination, toidentify equipment, or for cleaning purposes, may collect on screens and strainers and blockcore cooling systems. Several examples were cited.On October 4, 1995, the NRC Issued IN 95-47, 'Unexpected Opening of a Safety/Relief Valveand Complications Involving Suppression Pool Cooling Strainer Blockage,' which discussed anevent on September 11, 1995, at the Limerick Generating Station, Unit 1, during which asafety/relief valve discharged to the suppression pool. The operators started an RHR pump inthe suppression pool cooling mode. After 30 minutes, fluctuating motor current and flow wereobserved. Subsequent inspection of the strainers found them covered with a 'mat' of fibrousmaterial and sludge (corrosion products) from the suppression pool. The licensee removedapproximately 635 kg (1400 lb) of debris from the Unit I pool. A similar amount of debris hadbeen removed earlier from the Unit 2 pool. A supplement to IN 95-47 was issued onNovember 30, 1995.
Attachment 1GL 98-04July 14, 1998 On October 17, 1995, the NRC issued NRC Bulletin 95-02, Potential Clogging of a ResidualHeat Removal (RHR) Pump Strainer While Operating In Suppression Pool Cooling Mode,"which discussed the Limerick Unit 1 event and requested that BWR addressees review theoperability of their ECCS pumps and other pumps that draw suction from the suppression poolwhile performing their safety function. The addressees' evaluations were to take Intoconsideration suppression pool cleanliness, suction strainer cleanliness, and the effectivenessof the addressees' foreign material exclusion (FME) practices. In addition, BWR addresseeswere requested to Implement appropriate procedural modifications and other actions (e.g.,suppression pool cleaning), as necessary, in order to minimize the amounts of foreign materialin the suppression pool, drywell, and containment. BWR addressees were also requested toverify their operability evaluation through appropriate testing and Inspection.On February 10, 1996, the NRC issued IN 96-10, "Potential Blockage by Debris of SafetySystem Piping Which Is Not Used During Normal Operation or Tested During Surveillances,"which discussed debris blockage In ECCS lines taking suction from the containment sumps at aPWR in Spain. In one of the two partially blocked lines, almost half the flow area of the pipewas blocked; the other line was less blocked. Upon further investigation, Spanish regulatorsfound that many sections of piping in both PWRs and BWRs are only called upon to functionduring accident conditions and are not used during normal operation or tested during functionalsurveillance tests. The licensee in this case concluded that the safety significance was lowbecause the partial blockage of the lines would not have prevented the ECCS from providingsufficient core cooling. However, it was also noted that some of the debris could have beenentrained in the water flow and adversely affected other parts of the system (e.g., pump andvalve components and heat exchangers).In addition, in LER 96-005, the licensee for the H.B. Robinson Steam Electric Plant, Unit 2,reported finding in a pipe in the sump an item of debris larger than the 0.95-cm (3/8-in.)diameter of the holes In the containment spray nozzle.In LER 96-007, the licensee for Diablo Canyon Nuclear Power Plant, Unit 1, reported aradiograph inspection finding that openings in the Diablo Canyon plant's 3.81-cm (1-1/2 in.)centrifugal-charging-pump runout-protection manual throttle valves and In the 5.08-cm (2 in.)safety-Injection (SI) to cold-leg manual throttle valves were less than the 0.673-cm (0.265 In.)diagonal opening In the containment recirculation sump debris screen. Therefore, debris couldpotentially block charging or Si flow through these throttle valves during the recirculation phaseof a LOCA. The licensee concluded that even with a postulated blockage of the throttle valves,the RHR system flow by Itself would be sufficient to maintain adequate core cooling duringrecirculation following a postulated accident. As a corrective action, the Diablo Canyon licenseestated in LER 96-007 that the system would be modified to ensure that the throttle valveclearance Is greater than the maximum sump screen opening.After reviewing an Institute of Nuclear Power Operations (INPO) operational experience reporton this event, the licensee for Millstone Nuclear Station, Unit 2, determined that eight throttlevalves In the high-pressure safety Injection (HPSI) system injection lines were susceptible to thefailure mechanism described in Diablo Canyon Nuclear Power Plant LER 96-007. This situationIs discussed in NRC IN 96-27, 'Potential Clogging of High Pressure Safety Injection Throttle Attachment IGL 98-04July 14, -1998 Valves During Recirculation," dated May 1, 1996. The Millstone Unit 2 licensee concluded thatthe type of debris that would pass through the screen openings would tend to be of low densityand low structural strength and that material of this type would be reduced In size as It passedthrough the HPSI and containment spray pumps. In addition, the differential pressure acrossthe HPSI system injection valves and containment spray nozzles would tend to force throughthe valves or nozzles any material that is "marginally capable" of obstructing flow. Theseconclusions may be plant-specific and may not be applicable to other designs. However, inresponse to IN 96-27, the Millstone Unit 2 licensee committed to replace the sump screen withone that is consistent with the original design.On May 6, 1996, the NRC Issued Bulletin 96-03, "Potential Plugging of Emergency CoreCooling Suction Strainers by Debris In Boiling-Water Reactors," which requested actions byBWR addressees to resolve the Issue of BWR strainer blockage because of excessive buildupof debris from Insulation, corrosion products, and other particulates, such as paint chips andconcrete dust. The bulletin proposed four options for dealing with this Issue: (1) install large-capacity passive strainers, (2) install self-cleaning strainers, (3) Install a safety-related backflushsystem that relies on operator action to remove debris from the surface of the strainer to keep Itfrom clogging, or (4) propose another approach that offers an equivalent level of assurance thatthe ECCS will be able to perform its safety function following a LOCA. BWR addressees wererequested to Implement the requested actions of Bulletin 96-03 by the end of the first refuelingoutage beginning after January 1, 1997.On October 30, 1996, the NRC issued IN 96-59, "Potential Degradation of Post Loss-of-CoolantRecirculation Capability as a Result of Debris," to alert addressees that the suppression pooland associated components of two BWRs, LaSalle County Station, Unit 2, and Nine Mile PointNuclear Station, Unit 2, had been found to contain foreign objects that could have Impairedsuccessful operation of emergency safety systems that used water from the suppression pool.In particular, debris was found in the downcomers (large-diameter pipes connecting the drywellto the suppression pool). Although the licensee for Nine Mile Point, Unit 2, had previouslycleaned the suppression pool, the downcomers had not been inspected. In addition, thelicensee found debris covers in place on seven of the eight downcomers located in the pedestalarea directly under the reactor vessel. These debris covers had been In place sinceconstruction. LER 96-11-00 attributes this oversight to inadequate managerial methods and toenvironmental conditions: the "accessibility of the pedestal area downcomers requires removalof grating In the under vessel area and climbing down to the dimly lit subpile floor. The plasticcovers on the downcomers are not visible from the grating elevation because of the missileshield plates above the downcomer floor penetrations. Furthermore, since the first refuelingoutage, access to this area has been limited because of the high contamination levels andgeneral ALARA [as low as reasonably achievable radiation dose] considerations."Although the NRC has not previously discussed the subject in a generic communication,licensee event reports have been submitted regarding the loss of control of containment sumpaccess hatches, leaving them open during periods when ECCS sump Integrity was required.For example, in LER 89-014-01, the licensee for Diablo Canyon Nuclear Power Plant, Unit 1,discussed the opening of the sump access hatch at various times at power "without adequateconsideration of ECCS operability." In LER 96-006, the licensee for Watts Bar Nuclear Plant,Unit 1, reported that an operator observed a containment sump (trash screen) door open whileECCS operability was required.
-Attachment 2GL 98-04July 14, 1998 OPERATIONAL EVENTS INVOLVING DEBRIS IN ECCS RECIRCULATION FLOW PATHSPLANT/REPORT PROBLEMS DISCUSSEDHaddam Neck In July 1975, six 55-gallon drums of sludge with varyingNRC Inspection Report amounts of debris removed from ECCS sump.50-213/96-08North Anna Units 1/2 Galvanized ductwork painted with unqualified paint.LER 84-006-00Millstone Unit 1 Existing suction strainers too small when criteria of RG 1.82LER 88-004-00 Rev. I applied. Strainers will be replaced with larger strainers.Surry Power Station 1. Foreign material from construction activities found in coneUnits 1/2 strainer of recirculation spray system. Material could haveLER 88-017-01 rendered system inoperable.IN 88-87 2. Gaps in sump screens since Initial construction.IN 89-77Trojan Nuclear Plant 1. Wire mesh screen on top of sump trash rack not Installed.LER 89-016-01 2. Screen damage.IN 89-77 3. Significant amount of debris discovered in the sump. Couldhave caused loss of part or all of ECCS.Diablo Canyon Unit 1 1. Debris In sump.LER 89-014-01 2. As-built sump configuration not In accordance with design.IN 89-77 3. Safety function would not have been impaired.TMI Unit 1 1. Modification of sump access hatches left holes in top ofLER 90-002-00 sump screen cage.2. Could damage pumps or clog spray nozzles.McGuire Unit I Loose material discovered In upper containment before entryLER 900112-00 Into Mode 4. Items found would not have made ECCSInoperable.Calvert Cliffs Units 1/2 Unit 2 sump found to contain 11.3 kg (25 lb) dirt, weld slag,NRC Inspection Report pebbles, etc. Inspection of Unit I found less than1 lb debris. Possible minor damage to ECCS pumps.Diablo Canyon Unit 2 1. Numerous instances of material left unattended orLER 91-012-00 abandoned In sump level of containment (tools, plastic toolbags, clothing, etc.).2. Material would not have prevented ECCS recirculationl __ function.
Attachment 2GL 98-04July 14,1998 H.B. Robinson Unit 2 B safety Injection pump flow reduced due to blockage inLER 92-013-00 minimum flow recirculation check valve and flow orifice on July8, 1992. A pump OK. Foreign material also found in thel refueling water storage tank.H.B. Robinson Unit 2 On August 24, 1992, following a reactor trip, A and B safetyLER 92-018-00 Injection pumps Inoperable due to reduced flow. Found duringunscheduled surveillance to demonstrate Si operability.Pt. Beach Unit 2 September 18, 1992: During Technical Specification InserviceLER 92-003-01 inspection testing of the A containment spray pump the pumpIN 92-85 was declared Inoperable. A foam rubber plug was blockingpump suction. Plug removed and pump tested satisfactorily.One train of Unit 2 residual heat removal, safety Injection, andcontainment spray systems inoperable for entire operatingcycle. Plug was part of a cleanliness barrier.Perry Nuclear Plant May, 1992: During a refueling outage, foreign objectsLER 93-011-00 discovered In the containment side of the suppression pool.Fouling of RHR strainers found. Strainers not cleaned.January, 1993: RHR ANB strainers found deformed (collapsedinward in the direction of the fluid flow). Strainers replaced.March, 1993: RHR A/B operated In suppression pool coolingmode. Pump suction pressure decreased. Could havecompromised long-term RHR operation.Susquehanna Units 1/2 1. Assessing Impact of debris and corrosion productsLER 93-007-00 adhering to fibrous materials that may be dislodged by pipe(voluntary) break.2. Developing procedures to backflush strainers.Sequoyah Unit 2 Design basis limit for unqualified coatings inside containmentLER 93-026-00 had been exceeded. Additional quantity of unqualifiedcoatings on reactor coolant pump motor platform discovered.Path to ECCS sump. Screens will be Installed before startup.ANO Unit 2 Seven unscreened holes found in masonry grout below screenLER 93-002-00 assembly of ECCS sump. Could potentially degrade bothIN 89-77 Supplement I trains of HPSI and containment spray. Had previouslyInspected sump because of IN 89-77; did not discoverproblem. NRC estimate of incremental increase in coredamage: 3 X1004.
Attachment 2GL 98-04July 14, 1998 ANO Unit 1 1. 22 unscreened 15.2 cm x 7.6 cm (6"X3") pipe openings atLER 93-005-00 base of sump curb, the result of a modification before initialIN 89-77 Supplement I operation.2. Tears In screen.3. Floor drains leading to sump not screened.4. Licensee estimated increase In core damage frequency5X10°*5.San Onofre Units I and 2 1. Irregular annular gap (approximately 15.2 cm [6,)LER 93-010-00 surrounding 20.3 cm (8") low temperature overpressure(voluntary) discharge line penetrating horizontal steel cover plate.2. Engineering analysis concluded both sump trainsoperable.Vermont Yankee 1. LPCS suction strainers smaller than calculations assumed.LER 93-015-00 NPSH calculations performed in 1986 following change toNUKONTm insulation Invalid.2. Strainers replaced with larger strainers.South Texas Units 1/2 Sump screen openings from initial construction discovered.LER 94-001-00 Frame plate at floor warped, creating several openingsapproximately 1.6 cm (5/8"). Additional 0.6 cm (1/4") gapsdiscovered. Based on ECCS pump tests performed by themanufacturer, the licensee concluded the deficiencies had nosafety significance.Point Beach Unit I NRC Inspector found grout deterioration under sump screens.NRC Inspection Report Could result In flow bypass, or particles of grout could enter50-266/94-06 ECCS pumps.LaSalle Unit 1 April 26 and May 11, 1994: Divers inspecting suppression poolIN 94-57 during outage found operational debris.River Bend June 13, 1994: Plant in refueling outage. Foreign materialIN 94-57 found in suppression pool. Plastic bag removed from B RHRpump suction strainer. Other objects: tools, grinding wheel,scaffolding knuckle, stepoff pad.Quad Cities Unit 1 July 14, 1994, post-maintenance test of A loop RHR indicatedIN 94-57 a plugged torus cooling/test return valve. Inspectiondiscovered remains of shredded plastic bag in anti-cavitationtrim installed during a recent outage.July 23, 1994: 4" diameter wire brush and a piece of metall_ found wrapped around a vane of the C RHR pump.
Attachment 2OL 98-04July 14, 1998 Browns Ferry Units 1/2/3 1. Unqualified coatings on T quenchers in suppression pool.May 20, 1994, letter to NRC 2. Continued operation acceptable.3. Will remove coatings next refueling outage.Palisades Plant Signs, adhesive tape, and labels with potential to block theLER 94-014-00 ECCS sump were found in containment. Containment sprayand HPSI pumps declared inoperable. Engineering analysisconcluded that the sump screen would not be significantlyblocked.Watts Bar Units 1/2 Screens Installed around RCP motors to catch unqualifiedNRC Inspection Report paint not adequately located to contain all unqualified coatings.50-390 and 50-391/94-59Indian Point Unit 2 Licensee discovered portions of the floor on Elevation 46 inLER 95-005-00 containment had lifted and cracked. In other locations, floorcoating cracked when stepped on. Licensee concluded thatl_ sump function would not be compromised.Susquehanna Units 1/2 Licensee took actions to address concern of dogging ECCSLER 93-007-001 suction strainers. Among these actions: removal of fibrous(Voluntary) insulation from HELB areas, testing to determine whether thedebris could block the strainer, quantification of corrosionproducts on structural steel in wetwell, a comprehensiveanalysis of containment debris effects. Coating and insulationprocedures contain steps to reduce potential for strainerblockage.Prairie Island Unit 2 Broken labels for pipe hangers and labels affixed to wall withNRC Inspection Report degrading adhesive discovered by NRC Inspector after50-282/05-009 licensee closeout inspection. Licensee concluded that thispotential debris would not affect operability of ECCS.Palisades Unsecured material stored on the landings of stairways.NRC Inspection Report Broken glass and pieces of signboard and other50-225/95-008 unauthorized" material found In area designated debris-free.Limerick Unit I Debris was allowed to collect in suppression pool rendering theNRC Inspection Report A RHR pump inoperable when safety/relief valve lifted on50-352/96-04 September 11, 1995.Duane Arnold FME controls inadequate in drywell. Hardhats and debrisNRC Inspection Report noted.50-331/95-003 Attachment 2GL 98-04July 14, 1998 Foreign PWR 1. Operator found debris in the sump.NRC IN 96-10 2. Two of 4 ECCS lines taking suction from the sump werepartially blocked by debris. Debris present since plantconstruction.Millstone Unit 2 10 locations with screens whose mesh size was inconsistentLER 96-008 were identified. Placed plant outside original design basis.Sump screen replaced.Watts Bar Unit I Operator observed containment sump trash screen door wasLER 96-006-00 open when plant in MODE 4 and ECCS required to beoperable.Calvert Cliffs Units 1/2 Several holes identified In each unit's containment sumpLER 96-003-00 screen larger than described In the FSAR. Holes field-Installed for transmitter tubing. Concluded not a threat to plantsafety.Diablo Canyon Unit 1 Various debris that could pass through the containment sumpLER 96-007-00 screen could be larger than openings in the 3.8 cm (1-1/2")centrifugal-charging-pump runout-protection manual throttlevalves and 5.1 cm (2") SI-to-cold leg manual throttle valves.Haddam Neck 1. Discrepancies in sump screen mesh sizing, screen fitup,LER 96-014-00 and method of attachment discovered. Sump screenNRC Inspection Report replaced. Sump will be inspected after every refueling50-213/96-08 outage. Licensee reported this as a condition which alonecould have prevented the fulfillment of a safety function.2. Five 208 L (55-gallon) drums of sludge removed fromECCS sump. Also, plastic sheeting, nuts, and bolts, tewraps, and pencils.Big Rock Point Housekeeping in containment In the area under theNRC Inspection Report emergency condenser and the reactor depressurization50-155/96-004 system Isolation valves was poor."Catawba Unit 1 6 floor drains Inside crane wall were not covered with screenNRC Inspection Report that had a finer mesh than the sump screen. 0.6 cm (1/4")50-413/96-11 holes rather than 0.3 cm (1/8") holes. Crane wall penetrationsclose to containment floor could allow the transport of debris tothe sump screen. Penetrations sealed.Millstone Unit 2LER 96-08NRC Inspection Report50-336/96-08Containment sump screens had been Incorrectly constructedso that larger debris than analyzed could pass through theECCS.
Attachment 2GL 98-04July 14, 1998 Vogtle Unit 2NRC Inspection Report50-425/96-11LER 96-007-00Loose debris in 'readily accessible areas" Identified by NRCinspectors inside containment. Had the potential to blockemergency sump screens during accident conditions.Licensee's evaluation concluded that debris did not represent"substantial challenge" to ECCS. 0.6 m2 (6 ft2) of debrisestimated. Additional items Identified by licensee and NRCInspector during startup while In Mode 3. Further evaluation bylicensee concluded that RHR pump would not have hadadequate NPSH because of debris.Nine Mile Point Unit 2NRC Inspection Report50-410/96-11NRC Event Report 31172A significant amount of debris was found In the suppressionpool and downcomers during Refueling Outage 5. Licensee'spreliminary evaluation concluded that operability of ECCScould have been compromised.LaSalle Unit 2NRC Event Report 31159LER 96-009-00Foreign material recovered from suppression pool anddowncomers. This material would challenge the operability ofthe ECCS. Approximately 0.7 m2 (7 f:2) per strainer removedfrom suppression pool. Material most likely from constructionor early outages. Special multiple ECCS pump runs performedwith satisfactory results. No apparent transport of the foreignmaterial discovered during this outaqe.Millstone Unit 3 1. Construction debris discovered in containment recirculationLER 96-039-00 spray system (RSS) containment sump and In RSS suctionlines.2. Gaps discovered in RSS sump cover plates.3. Later inspection found other sump enclosure gaps.4. Bolts and clips missing from the vortex suppression grating.5. Debris found in all 4 RSS pump suction lines.H.B. Robinson Unit 2 1. Openings found in sump screens. They could have allowedLER 96-005-00 debris above a certain size to enter the sump or preventedthe screens from performing their'design function.2. An Item of debris in excess of the 1 cm (3/8") diametercontainment spray nozzles was found In 36 cm (14") sumpl_ drain pipe.Zion Unit 1 Two 2.5 cm (1") holes detailed on drawings were not in theLER 97-001-00 sump cover. Holes allow air to escape as sump fills. Potentiall_ .to hinder flow to RHR pump suction during a LOCA.Zion Unit 2 1. Miscellaneous debris found throughout containment.NRC Inspection Report 2. Containment recirculation sump screen damage.50-295/96-20 3. Peeling and flaking paint on containment surfaces.50-304/96-20
Attachment 2GL 98-04July 14, 1998 Sequoyah Unit 1 During shutdown on March 22, 1997, an oil cloth wasNRC Event 32139 Introduced into containment. If It had come free, it could haveblocked one or both refueling drains so that water In uppercontainment might not have flowed freely to lower level ofcontainment, where sump is located.Millstone Unit 1 Most of the coating in the torus is unqualified, which couldNRC Event Report 32161 affect the operability of the low-pressure coolant injection andcore spray systems.Clinton Significant degradation in protective coatings In theNRC Event Report 32633 containment wetwell. Some degradation In the drywell.Licensee concluded that the amount of degraded coatingsfrom the containment and the drywell could have exceeded theECCS strainer loading under accident conditions.St. Lucie Unit 2 Containment sump screens with gaps In screen enclosureLER 50-389/97-002 contrary to design.DC Cook Units 1/2 A 1 cm (0.25") particulate retention requirement for theNRC Event Report 32875 containment recirculation sump was not properly establishedfollowing sump modifications. Inadvertent pathways withopenings greater than 1 cm (0.25") were found, including 3 cm(0.75") vents in roof of sump (see following item). Ucenseeconcluded that the ECCS was outside Its design basis.Turkey Point Units 314 Gaps greater than 1 cm (0.25) found in screens for Unit 3 andNRC Event Report 32910 4 sumps.D.C. Cook Units1/2 Enough fibrous material was found in both Unit I and Unit 2NRC Event Report 32948 containments to potentially cause excessive blockage of thecontainment recirculation sump screen during the recirculationphase of a LOCA. Both units were already shut down for otherreasons. The material was removed from both units beforestartun.
Attachment 3GL 98-04July 14, 1998
BACKGROUND
ON REGULATORY BASIS FOR PROTECTIVE COATINGSThis appendix discusses the regulatory basis, including industry standards and regulatoryguidance, for protective coatings Inside the containment. However, this discussion Is only forinformation.
Addressees
should continue to comply with the plant Aicensing basis.At nuclear power plants, coatings and paints (1) protect carbon and low alloy steel, austeniticsteel, and less commonly, galvanized steel, and aluminum surfaces against corrosiveenvironments; (2) protect metallic, concrete, or masonry surfaces against wear during plantoperation; and (3) allow for ease of decontamination of radioactive nucides from thecontainment wall and floor surfaces. These coatings come in inorganic forms, such as zinc-based paints, and organic forms, such as epoxy coatings.ANSI Standards N101.2, "Protective Coatings (Paints) for Light Water Nuclear ReactorContainment Facilities," and ANSI N101 A, "Quality Assurance for Protective Coatings Appliedto Nuclear Facilities," classify coatings as Service Level 1, Service Level 2, or Service Level 3.Service Level 1 coatings are used in areas where coating failure could adversely affect theoperation of post-accident fluid systems and, thereby, impair safe shutdown. With fewexceptions, Service Level 1 applies to coatings inside primary containment.Service Level 2 coatings are used in areas where coating failure could Impair, but not prevent,normal operating performance. The function of Service Level 2 coatings is to provide corrosionprotection and improve the ability to decontaminate those areas outside primary containmentsubject to radiation exposure and radionuclide contamination.A Service Level 3 coating is used on any exposed surface area located outside containmentwhose failure could adversely affect normal plant operation or orderly and safe plant shutdown.This generic letter concerns the possible detrimental effects of failed coatings on a plant's abilityto recirculate coolant following a LOCA. Therefore, this generic letter is concerned with ServiceLevel I coatings.Protective coatings applied to the interior surfaces of the containment structure and to SSCsInside the containment are considered qualified coatings if they have been subjected to physicalproperty (adhesion) tests under conditions that simulate the projected environmental conditionsof a postulated design basis (DB) LOCA and have been demonstrated to maintain theiradhesive properties under these simulated conditions. These tests are typically conducted inaccordance with the guidelines, practices, test methods, and acceptance criteria specified inapplicable industry standards for coatings applications (such as those Issued by the AmericanNational Standards Institute, Inc. [ANSI, or the American Society for Testing and Materials[ASTM). However, the licensing basis for Service Level I coating applications may containexceptions to, or provide alternative means of meeting the intent of, the test methods in thesestandards. This requires that an adequate safety basis is given to and accepted by the NRCstaff as to why accepting the exceptions or alternatives would not affect the performance of the Attachment 3GL 98-04July 14,1998 ECCS and safety-related CSS during a postulated DB LOCA. In regard to protective coatingsused for Service Level I service applications inside the containment, the staff normallyconcludes that a coating system is acceptable for service if it has been demonstrated that thecoating system is qualified to maintain Its integrity during a postulated DB LOCA and If the.programs for controlling applications of coating systems for Service Level I service applicationsare implemented in accordance with a quality assurance (QA) program that meets therequirements of Appendix B to 10 CFR Part 50.According to Regulatory Guide (RG) 1.54, protective coatings that have not been successfullytested in accordance with the provisions in the applicable ANSI or ASTM standards or have notmet the acceptance criteria of the standards are considered to be "unqualified"; that is, they areassumed to be Incapable of maintaining their adhesive properties during a postulated DBLOCA. The staff normally assumes that "unqualified" coatings applied to the Interior surfaces ofthe containment structure and to SSCs inside the containment structure will form solid debrisproducts under DB LOCA conditions. These debris products should, therefore, be evaluated fortheir potential to clog ECCS sump screens and strainers and to affect the operability of safety-related pumps taking suction from ECCS sumps and suppression pools during a postulated DBLOCA.The NRC issued RG 1.54-1973, "Quality Assurance Requirements for Protective CoatingsApplied to Water-Cooled Nuclear Power Plants," to give the industry an acceptable method forcomplying with the QA requirements of 10 CFR Part 50, Appendix B. as they relate to protectivecoating systems applied to carbon and low alloy steel, austenitic stainless steel, aluminum,galvanized steel, and masonry surfaces of water-cooled nuclear power reactors. In RG 1.54-1973, the NRC stated that the guidelines for coating applications in ANSI Standard N101.4-1972, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities," assupplemented in RG 1.54-1973, delineate acceptable QA criteria for providing confidence that"shop or field coating work [will] perform satisfactorily in service." The quality assuranceprovisions stated In ANSI Standard N101.4-1972, as endorsed by the staff In RG 1.54-1973,are considered by the staff to provide an adequate basis for complying with the pertinent QArequirements of 10 CFR Part 50, Appendix B. These standards delineate the type of tests to beperformed to qualify a given coating for nuclear applications. However, how a licenseeImplements Its program for controlling activities related to protective coating applications at aparticular nuclear plant depends on the plant's licensing basis. Neither RG 1.54-1973 nor theapplicable ANSI standards are NRC requirements: they merely delineate acceptable programsand practices for controlling coating application activities at nuclear power plants.ANSI Standard N101.4-1972 provides recommended guidelines for Implementing QA programsregarding coating applications at domestic nuclear power plants. ANSI Standard N101.4-1972,as endorsed In RG 1.54-1973, delineates recommended guidelines and criteria for establishingQA and quality control programs for coating activities. Such programs should control workconditions, the ambient environmental conditions for coating applications, selection andprocurement activities for coatings, and preparation of substrate surfaces; establish QAprocedures for coating applications; qualify personnel involved In coating preparation,application, and Inspection activities; and establish coating Inspection guidelines and Attachment 3GL 98-04July 14, 1998 acceptance criteria. ANSI Standard N101.4-1972, as endorsed by RG 1.54-1973, alsorecommends keeping certain QA records on coatings activities.ANSI Standard NIOI.4-1972 states that ANSI Standard N5.9, "Protective Coatings (Paints) forthe Nuclear Industry (later reissued as ANSI Standard N512), and ANSI Standard N101.2,"Protective Coatings (Paints) for Light-Water Nuclear Reactor Containment Facilities," areadditional acceptable standards governing activities related to the selection and evaluation ofprotective coatings applied either in the shop (i.e., at vendor or manufacturer facilities) or in thefield.RG 1.54 is currently undergoing a major revision (it was last revised In 1973). Many of thedocuments referenced in RG 1.54 are outdated and have been replaced by newer ASTM orANSI standards. ASTM Committee D-33, "Coatings for Power Generation Facilities," hasdeveloped the standards that have replaced many of the standards referenced In RG 1.54-1973. At the request of the NRC staff, this committee is currently developing a maintenancestandard for qualified coatings. This standard will cover inspection of existing coatings,application of new coatings over the original substrate (steel, concrete, galvanized steel,aluminum), new coatings over a substrate-old coating Interface, and new coatings over oldqualified coatings. When this standard Is approved, RG 1.54-1973 will be revised to reflectcurrent standards. Using more up-to-date Industry standards for protective coatings mayrequire changing a plant's licensing basis. Use of these standards must conform with existingNRC requirements, including 10 CFR Part 50, Appendix B.
Attachment 4GL 98-04July 14, 1998 CHRONOLOGY OF INCIDENTS AND ACTIVITIES RELATED TO PROTECTIVE COATINGSIn January 1997, Commonwealth Edison Company (ComEd), the licensee for the Zion NuclearPlant, Unit 2, discovered flaking and unqualified paint applied to the containment surfaces (IN97-13, "Deficient Conditions Associated With Protective Coatings At Nuclear Power Plants").The peeling of the protective coatings was determined to occur at the horizontal junction linesbetween the concrete shells that were used in construction of the Zion Unit 2 containmentstructure. ComEd estimated that the total weight of degraded coatings (peeling paint) wasapproximately 445 N (100 lb). ComEd also Initially estimated that an additional 557-650 i2(6000-7000 ft2) of coatings on surfaces Inside containment were not qualified to withstand theenvironmental conditions of a postulated DB LOCA, in accordance with the testing criteria ofANSI Standard N512-1974. ComEd determined that the peeling of the qualified coatings on thecontainment surfaces was due to Improper surface preparation, resulting in inadequateadhesion of the coating following application.ComEd corrected the condition of the paint by removing all of the degraded "qualified" paintinside the Zion Unit 2 containment and all of the additional "unqualified" paints that weredetermined to be located within the analytically determined zone of Influence'. ComEd alsoperformed 33 random adhesion or opull" tests on the remaining, Intact, "qualified" paint insidethe containment structure. All of these tests were performed in accordance with the applicabletesting requirements specified in ANSI Standard N512-1974. All of the pull tests exhibitedvalues in excess of the 890 N (200 lb) required by the standard, thus demonstrating that theremaining qualified coatings were acceptable for service during the next operating cycle.On March 10, 1995, Consolidated Edison Company (ConEd), the licensee for Indian PointStation, Unit 2, reported In LER 95-005-00 that paint was peeling off the floor at the 14-m (46-ft)elevation of the Indian Point Unit 2 containment structure. The paint was applied to the 14-m(46-ft) floor elevation during the 1993 refueling outage as an Interim measure for reducingpersonnel radiation exposures until a more permanent floor resurfacing could be accomplished.ConEd determined that the following factors contributed to the cracking and delamination of thepaint: (1) in some areas, the paint had been applied In excess of the dry film thicknessrecommended by the manufacturer of the paint; (2) during preparation of the paint, too muchpaint thinner was added to the paint, which led to an excessive amount of coating shrinkagewhen the paint dried; (3) no scarification of the floor surface was performed before applicationof the paint to remove old coatings, greases, or silicone or wax buildups from the floor surface;and (4) the painters had not been trained to apply the particular brand of paint. ConEddetermined the root cause of the coatings event to be the painters' failure to follow controlledprocedures for applying the particular brand of paint. To address the nonconforming conditionof the paint, ConEd removed all of the old paint from the 14-m (46-fl) floor elevation andAll of the unqualified paint within the containment sump's zone of influence was removed, with the exception ofapproximately I cm2(11 ft2) of unqualified paint applied to small components, such as lighting fixtures or name tags.
Attachment 4GL 98-04July 14, 1998 repainted the floor elevation with a qualified coating in accordance with the station's proceduralrequirements and the manufacturer's recommendations for the paint.ConEd also retrained the paint specialists to Indoctrinate them regarding the Importance ofcomplying with the station's procedures and standards for coating applications.On October 18, 1993, the Tennessee Valley Authority (TVA) reported In LER 93-026 the use ofunidentified coatings on the surfaces of the No. 4 reactor coolant pump (RCP) motor housingsat the Sequoyah Nuclear Plant, Units I and 2. These coatings were not accounted for In thelicensee's QA Uncontrolled Coatings Log. TVA determined that the No. 4 RCP motor housingswere completely within the zones of Influence of the containment sumps at both Sequoyahunits. The unqualified coating on each No. 4 RCP motor housing amounted to an additional13.3 m2 (143 ft2); this amount was not accounted for by TVA in its 1986 assessment ofunqualified coatings on the RCP motor housings. The omission is significant because themaximum amount of uncontrolled coatings allowed by the Uncontrolled Coatings Logs for theSequoyah units is 5.3 m2 (56.5 ft2); this Is the maximum amount of uncontrolled coatings thatcan be in the zone of influence of the containment sump without having the potential to affectthe operability of the ECCS and safety-related CSS.The NRC summarized Its review of the safety significance of the amount of unqualified paint onthe No. 4 RCP motor housings in Inspection Reports (IR) Nos. 50-327/9342 and 50-328/93-42and in IR Nos. 50-327/94-25 and 50-328194-25, dated November 9, 1993, and September 12,1994, respectively. In IR Nos. 50-327/94-25 and 50-328/94-25, the NRC concluded that If theunqualified coatings on or within the RCP motor housings failed, they could potentially migrateto the containment sump during a postulated DB LOCA and impair the performance of thecontainment ECCS and the containment spray system during the event. TVA addressed thisissue by modifying the RCP motor housings to Include "catch" screens designed to preventcoating material on the motor housings from reaching the strainers In the containment sumps.On July 2, 1993, and September 11, 1995, the Pennsylvania Power and Light Company (PP&L)Issued LERs 93-007-00 and 93-007-01, to summarize Its reassessment of ECCS performanceat Susquehanna Steam Electric Station, Units 1 and 2, respectively, during a postulated DBLOCA. In Its Initial analysis of ECCS performance during a postulated DB LOCA, PP&Ldetermined that sources of fibrous Insulating materials could not Impair the operability of theECCS at Susquehanna Units I and 2. However, PP&L's initial analysis did not account for"unqualified" coatings as potential sources of debris.In LER 93-007-00, PP&L discussed the effect of debris on the performance of the ECCS duringa postulated DB LOCA. In the LER, PP&L stated that its increased awareness of the quantityof unqualified coatings and corrosion products ("other material") Inside the containment was akey factor in deciding to reassess the sources of debris inside the Susquehanna Unit 1 and 2containments during a postulated DB LOCA. PP&L considered fibrous Insulation material,unqualified coatings, and corrosion products as the sources of debris. PP&L's evaluation of thedebris during the postulated event contained the following uncertainftes: (1) uncertainty In Attachment 4GL 98-04July 14, 1998 qualifying the sources of debris within the containment, (2) uncertainty In determining theamount of debris that could be dislodged during a postulated DB LOCA, and (3) uncertainty inestablishing exactly how the debris would be transported from Its source to the ECCS strainersduring the postulated event. Because of these uncertainties, PP&L stated in the licensee eventreport that If unqualified coatings and corrosion products were Included among the materialsthat could become sources of debris, some potential existed for complete blockage of thesuppression pool strainers during the event.PP&L addressed this issue, In part, by requiring that DB LOCA qualification testing beperformed on all inorganic zinc paints inside the Susquehanna containments. PP&L alsoimplemented Improved administrative housekeeping and inventory controls and Issued anadministrative coating specification that restricted any coatings applied Inside the containmentstructures to qualified coatings.On April 16, 1997, the licensee for Millstone Nuclear Power Station, Unit 1, a BWR-3 with aMark I containment, reported to the NRC that a significant amount of coating work inside theMillstone Unit I torus (suppression pool) was unqualified. Millstone Unit I LER 97-026 statedthat a number of different coating materials had been used inside the torus, but the locationsand extent of various coating systems were unclear.On July 15, 1997, the licensee for Clinton Station, a BWR-6 with a Mark IlIl containment,reported to the NRC that a significant quantity of degraded protective coatings was removedfrom the primary containment and the drywell. The licensee stated that due to theindeterminate condition of these degraded coatings, reasonable assurance could not be giventhat they would not disbond from their substrates enough to clog the ECCS suction strainersduring accident conditions.
Attachment 5GL 98-04July 14,1998 GENERIC COMMUNICATIONS ISSUED BY THE NRC ON ECCS ANDSAFETY-RELATED CSS SUMP AND STRAINER BLOCKAGEGeneric Letter 85-22, "Potential for Loss of Post Loss of Coolant AccidentRecirculation Capability Due to Insulation Debris Blockage," December 3, 1985.IN 88-28, "Potential for Loss of Post Loss of Coolant Accident RecirculationCapability Due to Insulation Debris Blockage," May 19, 1988.IN 89-77, "Debris In Containment Emergency Sumps and Incorrect ScreenConfigurations," November 21, 1989.IN 92-71, "Partial Blockage of Suppression Pool Strainers at a Foreign BWR,"September 30, 1992.IN 92-85, Potential Failures of Emergency Core Cooling Systems by ForeignMaterial Blockage," December 23, 1992.IN 93-34, "Potential for Loss of Emergency Core Cooling Function Due to aCombination of Operational and Post Loss of Coolant Accident Debris InContainment," April 26, 1993.IN 93-34, Supplement 1, "Potential for Loss of Emergency Cooling Function Due to aCombination of Operational and Post Loss of Coolant Accident Debris inContainment," May 6,1993.Bulletin 93-02, "Debris Plugging of Emergency Core Cooling Suction Strainers,"May 11, 1993.NRC Bulletin 93-02, Supplement 1, "Debris Plugging of Emergency Core CoolingSuction Strainers," February 18, 1994.IN 94-57, "Debris In Containment and the Residual Heat Removal System,"August 12, 1994.IN 95-06, "Potential Blockage of Safety Related Strainers by Material Brought InsideContainment," January 25,1995.IN 95-47, "Unexpected Opening of a Safety/Relief Valve and Complications InvolvingSuppression Pool Cooling Strainer Blockage," October 4, 1995.
Attachment 5GL 98-04July 14, 1998 Bulletin 95-02, "Unexpected Clogging of a Residual Heat Removal (RHR) PumpStrainer While Operating in the Suppression Pool Cooling Mode," October 17,1995.IN 95-47, Revision 1, "Unexpected Opening of a Safety/Relief Valve andComplications Involving Suppression Pool Cooling Strainer Blockage,"November 30, 1995.IN 96-10, "Potential Blockage by Debris of Safety System Piping Which is Not UsedDuring Normal Operation or Tested During Surveillances," February 13, 1996.Bulletin 96-03, Potential Plugging of Emergency Core Cooling Suction Strainers byDebris In Boiling Water Reactors," May 6, 1996.IN 96-27, "Potential Clogging of High Pressure Safety Injection Throttle Valves DuringRecirculation," May 1, 1996.IN 96-55, "Inadequate Net Positive Suction Head of Emergency Core Cooling andContainment Heat Removal Pumps Under Design Basis Accident Conditions,"October22, 1996.IN 96-59, "Potential Degradation of Post Loss of Coolant Accident RecirculationCapability as a Result of Debris," October 30, 1996IN 97-13, "Deficient Conditions Associated With Protective Coatings at NuclearPower Plants," March 24, 1997.
--Attachment 6GL 98-04July 14,-1998Page 1 of ILIST OF RECENTLY ISSUED GENERIC LETTERSGENELETTE98-03RICR.SUBJECTNMSS Licensees' and CertificateHolders' Year 2000 ReadinessProgramsDATE OFISSUANCE06/22/98ISSUED TOAll licensees or certificateholders for uraniumhexafluoride productionplants, uranium enrichmentplants, and uranium fuelfabrication plants, exceptthose that have permanentlyceased operations98-0298-0197-06Loss of Reactor CoolantInventory and AssociatedPotential for Loss of EmergencyMitigation Functions While in aShutdown ConditionYear 2000 Readiness ofof Computer Systems atNuclear Power PlantsDegradation of SteamGenerator Intemals05/28/9805/12/9812/30/97All holders of OLS for PWRs,except those who havepermanently ceasedoperations, and havecertified that fuel has beenpermanently removed fromthe reactor vessel.All holders of OLS fornuclear power plants,except those who havepermanently ceasedoperations and havecertified that fuel has beenpermanently removed fromthe reactor vesselAll holders of OLS forpressurized-water reactors,except those who havepernmanently ceasedoperations and have certifiedthat fuel has been perman-ently removed from thereactor vesselOP = Operating LicenseCP = Construction PermitNPR = Nuclear Power Reactors GL 98-04July 14, 1998
3 -The Office of Enforcement reviewed this generic letter and has no objection to it. The Office ofthe General Counsel reviewed this generic letter and has no legal objection to it.The staff intends to issue this generic letter 5 working days after the date of this informationpaper.L. Joseph CallanExecutive Directorfor OperationsAttachments:Proposed NRC Generic Letter OPotential for Degradation of the EmergencyCore Cooling and Containment Spray System Because of Construction andProtective Coating Deficiencies and Foreign Material in Containment'Paul Keene 3/14I98Tech Editor DateDOCUMENT NAME: G:VDAVIS1SECY-GL.WPD *PREVIOUSLY CONCURREDTo receive a copy of this document, indicate in the box C=Copy wlo attachmentlenclosure E=Copy withAttachmentlenclosure N = No copy