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UNITED STATES        NUCLEAR REGULATORY COMMISSION        WASHINGTON, D.C. 20555-0001        July 14, 1998

**Addressees**

All holders of operating licenses for nuclear power reactors, except         those who have permanently ceased operations and have certified that fuel         has been permanently removed from the reactor vessel.

**Purpose**

The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic         letter for several reasons. It alerts addressees that foreign material         continues to be found inside operating nuclear 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-related         containment spray system (CSS) flow path or damage ECCS or safety-related         CSS equipment. In addition, construction deficiencies and problems with         the material condition of ECCS systems, structures, and components (SSCs)         inside the containment continue to be found. Design deficiencies also         have been found which could degrade the ECCS or safety-related CSS. No         action or information is requested regarding these issues. The NRC has         issued many previous generic communications on this subject, as discussed         later in this generic letter, and assumes that addressees have had adequate         prior notice to consider possible actions at their facilities to address         these concerns.

The NRC expects addressees to ensure that the ECCS and the safety-related         CSS remain capable 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 associated with the material condition of Service Level 1         (see definitions of Service Levels in Attachment         3) protective coatings inside the containment and to request information         under 10 CFR 50.54(f) to evaluate the addressees' programs for ensuring         that Service Level 1 protective coatings inside containment do not detach         from their substrate during a DB LOCA and interfere with the operation         of the ECCS and the safety-related CSS. The NRC intends to use this information         to assess whether current regulatory requirements are being correctly         implemented and whether they should be revised.

**Background**

*Foreign Material Exclusion, Construction Deficiencies, and Design         Deficiencies*

In some recent events (discussed in Attachment         1 to this generic letter), foreign material which could have affected         the operation of the ECCS was discovered inside the containment. As part         of its review of these events, the NRC staff found several types of continuing         problems.

The Discussion section of this generic letter describes the regulatory         and safety bases for these concerns.

A more complete list of the above events is provided in Attachment         2. As discussed in Attachment 1, almost         all of these events have been the subject of previous NRC generic communications         and LERs. Apparently, past NRC generic communications have not been completely         effective in focusing licensee attention to the potential areas of concern         to control these problems. Nevertheless, the NRC expects that licensees         will ensure that the ECCS and safety-related CSS remain capable of performing         their intended safety functions. The NRC plans to further emphasize this         issue by conducting inspections to ensure compliance with existing plant         licensing bases. The NRC intends to take enforcement action for discovered         inadequacies consistent with NRC Enforcement Policy.**

*Protective Coatings*

Protective coatings inside nuclear power plant containments serve three         general purposes. Protective coatings are applied to carbon and low alloy         steel and, less commonly, to aluminum and galvanized surfaces to control         corrosion, control radioactive contamination levels, and to protect surfaces         from wear. (Although aluminum and galvanized surfaces are not commonly         coated, nothing in NRC requirements or industry standards prevent these         surfaces from being coated.) A discussion on protective coatings inside         the containment and the regulatory requirements and guidance for their         use are discussed in Attachment 3.

It has been assumed that qualified protective coatings are capable of         adhering to their substrate during a DB LOCA in order to minimize the         amount of material which can reach the emergency sump screens or suction         strainers and clog them. The NRC is aware that not all coatings inside         the containment are qualified and, therefore, the amount of unqualified         coatings must be controlled since the unqualified coatings are assumed         to detach from their substrates during a DB LOCA and may be transported         to the emergency sump screens or suction strainers. Once in contact with         sump screens or suction strainers, coating chips may adversely impact         the net positive suction head (NPSH) available to the ECCS or CSS pumps.

Additionally, in some cases, coatings which were qualified failed during         normal operation. Some of these events are discussed in Attachment         4.

**Discussion**

NRC regulations in 10 CFR 50.46 require that licensees design their ECCS         to provide long-term cooling capability so that the core temperature can         be maintained at an acceptably low value and decay heat can be removed         for the extended period required by the long-lived radioactivity remaining         in the core. Licensees are required to demonstrate this capability while         assuming the most conservative single failure. Some addressees may credit         CSSs in the licensing basis for radioactive-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-generated debris, 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-related CSS and         thereby interfere with the long-term cooling function, source-term and         pressure reduction features of plant design.

Qualified coatings used inside containment should be capable of withstanding         the environmental conditions of a postulated DB LOCA. Although small,         localized areas of degraded coatings may not be indicative of widespread         failure of the coatings, the condition of the coatings should be evaluated         by suitable means. The Electric Power Research Institute (EPRI) has prepared         a guidance document for containment coatings titled "Guidelines on the         Elements of a Nuclear Safety-Related Coatings Program." Licensees may         find EPRI TR-109937, Final Report, dated April 1998 useful when evaluating         coatings, although it has not been endorsed by the staff. The LERs and         NRC inspection reports described in Attachment         4 to this generic letter provide evidence of weaknesses in addressee         programs with regard to applications of protective coatings for Service         Level 1. These weaknesses include deficiencies in addressee programs to:         (1) control the preparation and cleanliness of the substrate before the         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 control the use of, excessive amounts         of unqualified coatings inside the containment, (5) monitor the status         of "qualified" coatings already applied to the surfaces of the containment         structure and to other equipment inside the containment, and (6) assess         the safety significance of coatings inside containment that have been         determined to detach from their substrate and to repair these coatings         if necessary.

The NRC has issued a number of generic communications on problems with         protective coatings, and the potential for the loss of the ECCS and safety-related         CSS as a result of strainer clogging and debris blockage. These generic         communications are listed in Attachment 5.         They apply to both PWRs and BWRs. The events discussed in these generic         communications, and similar events described in LERs and NRC inspection         reports, demonstrate the need for a strong foreign material exclusion         (FME) program in all areas of PWRs and BWRs that may contain materials         that could interfere with the successful operation of the ECCS and the         safety-related CSS. Other events demonstrate the need to ensure the correct         design and to maintain the material condition of emergency core cooling         system and safety-related containment spray system SSCs, including the         suppression pools, ECCS strainers and sumps, and the protective coatings         inside containment.

**Applicable Regulatory Requirements**

The 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 staff considers that the requirements of         10 CFR Part 50, Appendix B, are germane to this issue for safety related         containment coatings.

Section 50.65 of 10 CFR, "Requirements for monitoring the effectiveness         of maintenance at nuclear power plants," (maintenance rule) includes in         its scope all safety-related SSCs and those non-safety-related SSCs that         fall into the following categories: (1) those that are relied upon 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-related function, 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 of the maintenance rule. The maintenance rule requires         that licensees monitor the effectiveness of maintenance for these protective         coatings (as discrete systems or components or as part of any SSC) in         accordance with paragraph (a)(1) or (a)(2) of 10 CFR 50.65, as appropriate.

Although this generic letter concerns coatings within the containment         and requests information about coatings within containment, addressees         should ensure that all coatings which meet the maintenance rule scoping         criteria are included in the programs and procedures for implementing         the maintenance rule.

The NRC has conducted numerous inspections in the areas addressed by         this generic letter; for example, the NRC issued Temporary Instruction         (TI) 2515/125, "Foreign Material Exclusion Controls," on August 25, 1994.         Violations discovered during the TI 2515/125 inspections have been identified         and appropriate enforcement action has been taken in accordance with the         NRC's Enforcement Policy (NUREG-1600, "General Statement of Policy and         Procedures for NRC Enforcement Actions: Enforcement Policy").

The NRC will continue to conduct inspections in these areas and will         consider the long history of generic communications on the issues addressed         by this generic letter as prior notice to licensees when assessing civil         penalties in accordance with Section VI.B.2 of the Enforcement Policy.         Finally, notwithstanding the normal civil penalty assessment, the NRC         will consider whether the circumstances of the case warrant escalation         of enforcement sanctions in accordance 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 a response to the required information, an addressee         determines that its facility is not in compliance with the Commission's         requirements, the addressee is expected to take appropriate actions in         accordance with both requirements stated in Appendix B to 10 CFR Part         50 and the plant technical specifications to restore the facility to compliance.

Licensees are encouraged to work closely with their owners groups and         industry associations to coordinate the responses to this letter to improve         the efficiency of the responses. The information submitted in response         to this generic letter should be considered to be public information.

**Required Information**

As a result of NRC findings in these areas and due to the importance         of ensuring system functionality, within 120 days of the date of this         generic letter, addressees are required to submit a written response that         includes the following information:

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 to Section 182a of the Atomic Energy Act         of 1954, as amended, and 10 CFR 50.54(f). This information 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         appropriate regional administrator.

**Backfit Discussion**

This generic letter requires information from the addressees under the         provisions of Section 182a of the Atomic Energy Act of 1954, as amended,         and 10 CFR 50.54(f). This generic letter does not constitute a backfit         as defined in 10 CFR 50.109(a)(1) since it does not impose modifications         or additions to systems, structures, and components or to the design or         operation of an addressee's facility. It also does not it impose an interpretation         of the 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 Requirement**

This generic letter transmits a requirement to submit information pursuant         to the provisions of Section 182a of the Atomic Energy Act of 1954, as         amended, and 10 CFR 50.54(f) for the purpose of verifying compliance with         applicable regulatory requirements. The required information will enable         the NRC staff to determine whether the addressees' protective coatings         inside the containment comply and conform with the current licensing basis         for their facilities and whether the regulatory requirements pursuant         to 10 CFR 50.46 are being met.

Protective coatings are necessary inside containment to control radioactive         contamination and to protect surfaces from erosion and corrosion. Detachment         of the coatings from the substrate may 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 licensing         basis of controlling containment pressure and radioactivity following         a LOCA.

**Paperwork Reduction Act Statement**

This generic letter mandates information collections that are subject         to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These         information collections were approved by the Office of Management and         Budget, approval number 3150-0011, which expires on September 30, 2000.

The public reporting burden for this collection of information is estimated         to average 400 hours per response, including the time for reviewing instructions,         searching existing data sources, gathering and maintaining the data needed,         and completing and reviewing the collection of information. The NRC is         seeking public comment on the potential impact of the collection of information         requested in the generic letter and on the following issues:

Send comments on any aspect of this collection of information, including         suggestions for reducing this burden, to the Information and Records Management         Branch, T-6F33, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,         or by Internet electronic mail at [email protected]: and to the Desk Officer,         Office of Information and Regulatory Affairs, NEOB-10202 (3150-0011),         Office of Management and Budget, Washington, DC 20503.

The NRC may not conduct or sponsor, and a person is not required to respond         to, a collection of 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 the lead project manager listed below, or the appropriate         Office of Nuclear Reactor Regulation project manager.

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**ECCS SUMP AND STRAINER EVENTS INVOLVING FOREIGN         MATERIAL INSIDE THE        CONTAINMENT AND CONSTRUCTION AND DESIGN DEFICIENCIES**

On November 16, 1988, the NRC issued Information Notice (IN) 88-87, "Pump         Wear and Foreign Objects in Plant Piping Systems," concerning several         incidents in which the potential existed for a flow reduction as a result         of pump wear and foreign objects in plant piping systems. In one of these         incidents, the licensee found foreign objects in a temporary pump discharge         cone strainer. The licensee investigated further and found foreign objects,         dating to early construction modifications, in the sump. In addition,         various deficiencies were found in the sump screens.

On November 21, 1989, the NRC issued IN 89-77, "Debris in Containment         Emergency Sumps and Incorrect Screen Configurations," which discussed         loose parts and debris in the containment sumps of three pressurized-water         reactors (PWRs), Surry Units 1 and 2 and Trojan. At Surry Units 1 and         2, some of the debris was large enough to cause pump damage or flow degradation.         In addition, some of the screens had gaps large enough to allow additional         loose material to enter the sump. The licensee found that screens that         separate the redundant trains of the recirculation spray system were missing         at both units. At Trojan, the licensee discovered 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 inner screen. IN 89-77 also reported that in 1980 the Trojan licensee         found a welding rod jammed between the impeller and the casing ring of         a residual heat removal (RHR) pump.

On December 23, 1992, the NRC issued IN 92-85, "Potential Failures of         Emergency Core Cooling Systems Caused by Foreign Material Blockage," which         alerted licensees to events at two PWRs. In these events, foreign material         blocked flow paths within the ECCS safety injection 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 of Emergency Cooling Function Due to a Combination of Operational         and Post-LOCA Debris in Containment," and its supplement. In these information         notices, the NRC described several instances of clogged ECCS pump strainers,         including two events at the Perry Nuclear Power Plant, 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-type         material and a coating of fine dirt." After cleaning the strainers in         January 1993, the licensee discovered that RHR A and B strainers were         deformed. The strainers were replaced. The second Perry event involved         an RHR pump test which was run after a plant transient in March 1993.         Pump suction pressure dropped to 0 KPa (0 psig). No change in pump flow         rate was observed. Material found on the strainer screen was analyzed         and found to consist of glass fibers from temporary drywell cooling filters         that had been inadvertently dropped into the suppression pool and corrosion         products that had been filtered from the pool by the glass fibers adhering         to the surface of the strainer. The material significantly increased the         pressure drop across the strainer.

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 suppression pool clean.

On May 11, 1993, the NRC issued Bulletin 93-02, "Debris Plugging of Emergency         Core Cooling Suction Strainers," which requested that both PWR and BWR         addressees (1) identify fibrous air filters and other temporary sources         of fibrous material in containment not designed to withstand a loss-of-coolant         accident (LOCA) and (2) take prompt action to remove the foreign matter         and ensure 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 keep foreign material out. Holes in the         masonry grout below the sump screen assembly would have let water into         the sump without being screened. The licensee attributed this condition         to failure to implement design basis requirements for the sump during         initial plant construction. The holes were difficult to detect. The holes         appeared to be part of the design because of their uniform spacing and         because they were "somewhat recessed...such that to see the holes they         must 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 Heat Removal System," which alerted operating reactor licensees         to additional instances of degradation of ECCS components because of debris.         At River Bend Station, the licensee found 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-cavitation         trim of an RHR test return valve. Subsequent to that event at Quad Cities,         Unit 1, the licensee observed 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-Related Strainers by Material Brought Inside Containment," which         discussed a concern that plastic or fibrous material, brought inside the         containment to reduce the spread of loose contamination, to identify equipment,         or for cleaning purposes, may collect on screens and strainers and block         core cooling systems. Several examples were cited.

On October 4, 1995, the NRC issued IN 95-47, "Unexpected Opening of a         Safety/Relief Valve and Complications Involving Suppression Pool Cooling         Strainer Blockage," which discussed an event on September 11, 1995, at         the Limerick Generating Station, Unit 1, during which a safety/relief         valve discharged to the suppression pool. The operators started an RHR         pump in the suppression pool cooling mode. After 30 minutes, fluctuating         motor current and flow were observed. Subsequent inspection of the strainers         found them covered with a "mat" of fibrous material and sludge (corrosion         products) from the suppression pool. The licensee removed approximately         635 kg (1400 lb) of debris from the Unit 1 pool. A similar amount of debris         had been removed earlier from the Unit 2 pool. A supplement to IN 95-47         was issued on November 30, 1995.

On October 17, 1995, the NRC issued NRC Bulletin 95-02, "Potential Clogging         of a Residual Heat Removal (RHR) Pump Strainer While Operating in Suppression         Pool Cooling Mode," which discussed the Limerick Unit 1 event and requested         that BWR addressees review the operability of their ECCS pumps and other         pumps that draw suction from the suppression pool while performing their         safety function. The addressees' evaluations were to take into consideration         suppression pool cleanliness, suction strainer cleanliness, and the effectiveness         of the addressees' foreign material exclusion (FME) practices. In addition,         BWR addressees were requested to implement appropriate procedural modifications         and other actions (e.g., suppression pool cleaning), as necessary, in         order to minimize the amounts of foreign material in the suppression pool,         drywell, and containment. BWR addressees were also requested to verify         their operability evaluation through appropriate testing and inspection.

On February 10, 1996, the NRC issued IN 96-10, "Potential Blockage by         Debris of Safety System 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 a PWR in Spain. In         one of the two partially blocked lines, almost half the flow area of the         pipe was blocked; the other line was less blocked. Upon further investigation,         Spanish regulators found that many sections of piping in both PWRs and         BWRs are only called upon to function during accident conditions and are         not used during normal operation or tested during functional surveillance         tests. The licensee in this case concluded that the safety significance         was low because the partial blockage of the lines would not have prevented         the ECCS from providing sufficient core cooling. However, it was also         noted that some of the debris could have been entrained in the water flow         and adversely affected other parts of the system (e.g., pump and valve         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 a radiograph 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 could potentially block charging or SI flow through         these throttle valves during the recirculation phase of 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 during recirculation following a postulated accident. As         a corrective action, the Diablo Canyon licensee stated in LER 96-007 that         the system would be modified to ensure that the throttle valve clearance         is greater than the maximum sump screen opening.

After reviewing an Institute of Nuclear Power Operations (INPO) operational         experience report on this event, the licensee for Millstone Nuclear Station,         Unit 2, determined that eight throttle valves in the high-pressure safety         injection (HPSI) system injection lines were susceptible to the failure         mechanism described in Diablo Canyon Nuclear Power Plant LER 96-007. This         situation is discussed in NRC IN 96-27, "Potential Clogging of High Pressure         Safety Injection Throttle Valves During Recirculation," dated May 1, 1996.         The Millstone Unit 2 licensee concluded that the type of debris that would         pass through the screen openings would tend to be of low density and low         structural strength and that material of this type would be reduced in         size as it passed through the HPSI and containment spray pumps. In addition,         the differential pressure across the HPSI system injection valves and         containment spray nozzles would tend to force through the valves or nozzles         any material that is "marginally capable" of obstructing flow. These conclusions         may be plant-specific and may not be applicable to other designs. However,         in response to IN 96-27, the Millstone Unit 2 licensee committed to replace         the sump screen with one that is consistent with the original design.

On May 6, 1996, the NRC issued Bulletin 96-03, "Potential Plugging of         Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors,"         which requested actions by BWR addressees to resolve the issue of BWR         strainer blockage because of excessive buildup of debris from insulation,         corrosion products, and other particulates, such as paint chips and concrete         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 backflush system that relies on         operator action to remove debris from the surface of the strainer to keep         it from clogging, or (4) propose another approach that offers an equivalent         level of assurance that the ECCS will be able to perform its safety function         following a LOCA. BWR addressees were requested to implement the requested         actions of Bulletin 96-03 by the end of the first refueling outage beginning         after January 1, 1997.

On October 30, 1996, the NRC issued IN 96-59, "Potential Degradation         of Post Loss-of-Coolant Recirculation Capability as a Result of Debris,"         to alert addressees that the suppression pool and associated components         of two BWRs, LaSalle County Station, Unit 2, and Nine Mile Point Nuclear         Station, Unit 2, had been found to contain foreign objects that could         have impaired successful 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 drywell to the suppression         pool). Although the licensee for Nine Mile Point, Unit 2, had previously         cleaned the suppression pool, the downcomers had not been inspected. In         addition, the licensee found debris covers in place on seven of the eight         downcomers located in the pedestal area directly under the reactor vessel.         These debris covers had been in place since construction. LER 96-11-00         attributes this oversight to inadequate managerial methods and to environmental         conditions: the "accessibility of the pedestal area downcomers requires         removal of grating in the under vessel area and climbing down to the dimly         lit subpile floor. The plastic covers on the downcomers are not visible         from the grating elevation because of the missile shield plates above         the downcomer floor penetrations. Furthermore, since the first refueling         outage, access to this area has been limited because of the high contamination         levels and general 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 sump access 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 adequate consideration 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 while ECCS operability         was required.

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**OPERATIONAL EVENTS INVOLVING DEBRIS IN ECCS RECIRCULATION         FLOW PATHS**

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**BACKGROUND ON REGULATORY BASIS FOR PROTECTIVE         COATINGS**

This appendix discusses the regulatory basis, including industry standards         and regulatory guidance, for protective coatings inside the containment.         However, this discussion is only for information. Addressees should continue         to comply with the plant licensing basis.

At nuclear power plants, coatings and paints (1) protect carbon and low         alloy steel, austenitic steel, and less commonly, galvanized steel, and         aluminum surfaces against corrosive environments; (2) protect metallic,         concrete, or masonry surfaces against wear during plant operation; and         (3) allow for ease of decontamination of radioactive nuclides from the         containment 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 Reactor Containment Facilities," and ANSI N101.4, "Quality Assurance         for Protective Coatings Applied to 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 the operation of post-accident fluid systems and, thereby,         impair safe shutdown. With few exceptions, 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 corrosion protection and improve         the ability to decontaminate those areas outside primary containment subject         to radiation exposure and radionuclide contamination.

A Service Level 3 coating is used on any exposed surface area located         outside containment whose 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 ability to recirculate coolant following a LOCA.         Therefore, this generic letter is concerned with Service Level 1 coatings.

Protective coatings applied to the interior surfaces of the containment         structure and to SSCs inside the containment are considered qualified         coatings if they have been subjected to physical property (adhesion) tests         under conditions that simulate the projected environmental conditions         of a postulated design basis (DB) LOCA and have been demonstrated to maintain         their adhesive properties under these simulated conditions. These tests         are typically conducted in accordance with the guidelines, practices,         test methods, and acceptance criteria specified in applicable industry         standards for coatings applications (such as those issued by the American         National Standards Institute, Inc. [ANSI], or the American Society for         Testing and Materials [ASTM]). However, the licensing basis for Service         Level I coating applications may contain exceptions to, or provide alternative         means of meeting the intent of, the test methods in these standards. This         requires that an adequate safety basis is given to and accepted by the         NRC staff as to why accepting the exceptions or alternatives would not         affect the performance of the ECCS and safety-related CSS during a postulated         DB LOCA. In regard to protective coatings used for Service Level I service         applications inside the containment, the staff normally concludes that         a coating system is acceptable for service if it has been demonstrated         that the coating 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 applications are implemented         in accordance with a quality assurance (QA) program that meets the requirements         of Appendix B to 10 CFR Part 50.

According to Regulatory Guide (RG) 1.54, protective coatings that have         not been successfully tested in accordance with the provisions in the         applicable ANSI or ASTM standards or have not met the acceptance criteria         of the standards are considered to be "unqualified"; that is, they are         assumed to be incapable of maintaining their adhesive properties during         a postulated DB LOCA. The staff normally assumes that "unqualified" coatings         applied to the interior surfaces of the containment structure and to SSCs         inside the containment structure will form solid debris products under         DB LOCA conditions. These debris products should, therefore, be evaluated         for their 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 DB LOCA.

The NRC issued RG 1.54-1973, "Quality Assurance Requirements for Protective         Coatings Applied to Water-Cooled Nuclear Power Plants," to give the industry         an acceptable method for complying with the QA requirements of 10 CFR         Part 50, Appendix B, as they relate to protective coating 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," as supplemented 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 assurance         provisions 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 QA requirements of 10 CFR Part 50, Appendix         B. These standards delineate the type of tests to be performed to qualify         a given coating for nuclear applications. However, how a licensee implements         its program for controlling activities related to protective coating applications         at a particular nuclear plant depends on the plant's licensing basis.         Neither RG 1.54-1973 nor the applicable ANSI standards are NRC requirements:         they merely delineate acceptable programs and practices for controlling         coating application activities at nuclear power plants.

ANSI Standard N101.4-1972 provides recommended guidelines for implementing         QA programs regarding 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 establishing QA and quality control programs         for coating activities. Such programs should control work conditions,         the ambient environmental conditions for coating applications, selection         and procurement activities for coatings, and preparation of substrate         surfaces; establish QA procedures for coating applications; qualify personnel         involved in coating preparation, application, and inspection activities;         and establish coating inspection guidelines and acceptance criteria. ANSI         Standard N101.4-1972, as endorsed by RG 1.54-1973, also recommends keeping         certain QA records on coatings activities.

ANSI Standard N101.4-1972 states that ANSI Standard N5.9, "Protective         Coatings (Paints) for the Nuclear Industry" (later reissued as ANSI Standard         N512), and ANSI Standard N101.2, "Protective Coatings (Paints) for Light-Water         Nuclear Reactor Containment Facilities," are additional acceptable standards         governing activities related to the selection and evaluation of protective         coatings applied either in the shop (i.e., at vendor or manufacturer facilities)         or in the field.

RG 1.54 is currently undergoing a major revision (it was last revised         in 1973). Many of the documents referenced in RG 1.54 are outdated and         have been replaced by newer ASTM or ANSI standards. ASTM Committee D-33,         "Coatings for Power Generation Facilities," has developed 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         maintenance standard 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         old qualified coatings. When this standard is approved, RG 1.54-1973 will         be revised to reflect current standards. Using more up-to-date industry         standards for protective coatings may require changing a plant's licensing         basis. Use of these standards must conform with existing NRC requirements,         including 10 CFR Part 50, Appendix B.

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**CHRONOLOGY OF INCIDENTS AND ACTIVITIES RELATED         TO PROTECTIVE COATINGS**

In January 1997, Commonwealth Edison Company (ComEd), the licensee for         the Zion Nuclear Plant, Unit 2, discovered flaking and unqualified paint         applied to the containment surfaces (IN 97-13, "Deficient Conditions Associated         With Protective Coatings At Nuclear Power Plants"). The peeling of the         protective coatings was determined to occur at the horizontal junction         lines between the concrete shells that were used in construction of the         Zion Unit 2 containment structure. ComEd estimated that the total weight         of degraded coatings (peeling paint) was approximately 445 N (100 lb).         ComEd also initially estimated that an additional 557-650 m2         (6000-7000 ft2) of coatings on surfaces inside containment         were not qualified to withstand the environmental conditions of a postulated         DB LOCA, in accordance with the testing criteria of ANSI Standard N512-1974.         ComEd determined that the peeling of the qualified coatings on the containment         surfaces was due to improper surface preparation, resulting in inadequate         adhesion of the coating following application.

ComEd corrected the condition of the paint by removing all of the degraded         "qualified" paint inside the Zion Unit 2 containment and all of the additional         "unqualified" paints that were determined to be located within the analytically         determined zone of influence(1). ComEd         also performed 33 random adhesion or "pull" tests on the remaining, intact,         "qualified" paint inside the containment structure. All of these tests         were performed in accordance with the applicable testing requirements         specified in ANSI Standard N512-1974. All of the pull tests exhibited         values in excess of the 890 N (200 lb) required by the standard, thus         demonstrating that the remaining qualified coatings were acceptable for         service during the next operating cycle.

On March 10, 1995, Consolidated Edison Company (ConEd), the licensee         for Indian Point Station, 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 reducing personnel radiation exposures until a more permanent         floor resurfacing could be accomplished. ConEd determined that the following         factors contributed to the cracking and delamination of the paint: (1)         in some areas, the paint had been applied in excess of the dry film thickness         recommended by the manufacturer of the paint; (2) during preparation of         the paint, too much paint thinner was added to the paint, which led to         an excessive amount of coating shrinkage when the paint dried; (3) no         scarification of the floor surface was performed before application of         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. ConEd determined the root cause of the         coatings event to be the painters' failure to follow controlled procedures         for applying the particular brand of paint. To address the nonconforming         condition of the paint, ConEd removed all of the old paint from the 14-m         (46-ft) floor elevation and repainted the floor elevation with a qualified         coating in accordance with the station's procedural requirements and the         manufacturer's recommendations for the paint.

ConEd also retrained the paint specialists to indoctrinate them regarding         the importance of complying 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 of unidentified coatings on the surfaces of the No.         4 reactor coolant pump (RCP) motor housings at the Sequoyah Nuclear Plant,         Units 1 and 2. These coatings were not accounted for in the licensee's         QA Uncontrolled Coatings Log. TVA determined that the No. 4 RCP motor         housings were completely within the zones of influence of the containment         sumps at both Sequoyah units. The unqualified coating on each No. 4 RCP         motor housing amounted to an additional 13.3 m2 (143 ft2);         this amount was not accounted for by TVA in its 1986 assessment of unqualified         coatings on the RCP motor housings. The omission is significant because         the maximum amount of uncontrolled coatings allowed by the Uncontrolled         Coatings Logs for the Sequoyah units is 5.3 m2 (56.5 ft2);         this is the maximum amount of uncontrolled coatings that can be in the         zone of influence of the containment sump without having the potential         to affect the operability of the ECCS and safety-related CSS.

The NRC summarized its review of the safety significance of the amount         of unqualified paint on the No. 4 RCP motor housings in Inspection Reports         (IR) Nos. 50-327/93-42 and 50-328/93-42 and in IR Nos. 50-327/94-25 and         50-328/94-25, 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 the         unqualified coatings on or within the RCP motor housings failed, they         could potentially migrate to the containment sump during a postulated         DB LOCA and impair the performance of the containment ECCS and the containment         spray system during the event. TVA addressed this issue by modifying the         RCP motor housings to include "catch" screens designed to prevent coating         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 performance at Susquehanna Steam Electric Station,         Units 1 and 2, respectively, during a postulated DB LOCA. In its initial         analysis of ECCS performance during a postulated DB LOCA, PP&L determined         that sources of fibrous insulating materials could not impair the operability         of the ECCS at Susquehanna Units 1 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 during a postulated DB LOCA. In the LER, PP&L stated that         its increased awareness of the quantity of unqualified coatings and corrosion         products ("other material") inside the containment was a key factor in         deciding to reassess the sources of debris inside the Susquehanna Unit         1 and 2 containments 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 the debris during the         postulated event contained the following uncertainties: (1) uncertainty         in qualifying the sources of debris within the containment, (2) uncertainty         in determining the amount of debris that could be dislodged during a postulated         DB LOCA, and (3) uncertainty in establishing exactly how the debris would         be transported from its source to the ECCS strainers during the postulated         event. Because of these uncertainties, PP&L stated in the licensee         event report that if unqualified coatings and corrosion products were         included among the materials that could become sources of debris, some         potential existed for complete blockage of the suppression pool strainers         during the event.

PP&L addressed this issue, in part, by requiring that DB LOCA qualification         testing be performed on all inorganic zinc paints inside the Susquehanna         containments. PP&L also implemented improved administrative housekeeping         and inventory controls and issued an administrative coating specification         that restricted any coatings applied inside the containment structures         to qualified coatings.

On April 16, 1997, the licensee for Millstone Nuclear Power Station,         Unit 1, a BWR-3 with a Mark I containment, reported to the NRC that a         significant amount of coating work inside the Millstone Unit 1 torus (suppression         pool) was unqualified. Millstone Unit 1 LER 97-026 stated that a number         of different coating materials had been used inside the torus, but the         locations and extent of various coating systems were unclear.

On July 15, 1997, the licensee for Clinton Station, a BWR-6 with a Mark         III containment, reported to the NRC that a significant quantity of degraded         protective coatings was removed from the primary containment and the drywell.         The licensee stated that due to the indeterminate condition of these degraded         coatings, reasonable assurance could not be given that they would not         disbond from their substrates enough to clog the ECCS suction strainers         during accident conditions.

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**GENERIC COMMUNICATIONS ISSUED BY THE NRC ON ECCS         AND        SAFETY-RELATED CSS SUMP AND STRAINER BLOCKAGE**

Generic Letter 85-22, "Potential for Loss of Post Loss of Coolant Accident         Recirculation Capability Due to Insulation Debris Blockage," December         3, 1985.

IN 88-28, "Potential for Loss of Post Loss of Coolant Accident Recirculation         Capability Due to Insulation Debris Blockage," May 19, 1988.

IN 89-77, "Debris in Containment Emergency Sumps and Incorrect Screen         Configurations," 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 Foreign         Material Blockage," December 23, 1992.

IN 93-34, "Potential for Loss of Emergency Core Cooling Function Due         to a Combination of Operational and Post Loss of Coolant Accident Debris         in Containment," April 26, 1993.

IN 93-34, Supplement 1, "Potential for Loss of Emergency Cooling Function         Due to a Combination of Operational and Post Loss of Coolant Accident         Debris in Containment," 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         Cooling Suction 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 Inside Containment," January 25, 1995.

IN 95-47, "Unexpected Opening of a Safety/Relief Valve and Complications         Involving Suppression Pool Cooling Strainer Blockage," October 4, 1995.

Bulletin 95-02, "Unexpected Clogging of a Residual Heat Removal (RHR)         Pump Strainer While Operating in the Suppression Pool Cooling Mode," October         17, 1995.

IN 95-47, Revision 1, "Unexpected Opening of a Safety/Relief Valve and         Complications Involving Suppression Pool Cooling Strainer Blockage," November         30, 1995.

IN 96-10, "Potential Blockage by Debris of Safety System Piping Which         is Not Used During Normal Operation or Tested During Surveillances," February         13, 1996.

Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction         Strainers by Debris in Boiling Water Reactors," May 6, 1996.

IN 96-27, "Potential Clogging of High Pressure Safety Injection Throttle         Valves During Recirculation," May 1, 1996.

IN 96-55, "Inadequate Net Positive Suction Head of Emergency Core Cooling         and Containment Heat Removal Pumps Under Design Basis Accident Conditions,"         October 22, 1996.

IN 96-59, "Potential Degradation of Post Loss of Coolant Accident Recirculation         Capability as a Result of Debris," October 30, 1996

IN 97-13, "Deficient Conditions Associated With Protective Coatings at         Nuclear Power Plants," March 24, 1997.

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1.  All of the unqualified paint within the containment         sump's zone of influence was removed, with the exception of approximately         1 cm2 (11 ft2 ) of unqualified paint applied to         small components, such as lighting fixtures or name tags.