Proposed Generic Letter; Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in the Containment (TAC N0. M97146)

NUCLEAR REGULATORY COMMISSION

Proposed Generic Letter; Potential for Degradation of the

Emergency Core Cooling System and the Containment Spray System After a

Loss-of-Coolant Accident Because of Construction and Protective Coating

Deficiencies and Foreign Material in the Containment (TAC N0. M97146)

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of opportunity for public comment.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to issue

a generic letter to licensees of operating nuclear power reactors

regarding the potential for degradation of the emergency core cooling

system (ECCS) and the containment spray system (CSS) after a loss-of-

coolant accident (LOCA) because of construction and protective coating

deficiencies and foreign material that may be present in the

containment. The NRC is issuing this generic letter to alert licensees

to the fact that foreign material continues to be found inside

operating nuclear power plant containments. During a design basis LOCA,

this foreign material could block the ECCS or safety-related 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 potentially degrade the ECCS or safety-related CSS. No actions or

information are requested regarding these issues. The NRC has issued

many previous generic communications on this subject and expects

licensees to have considered possible actions at their facilities to

address these concerns.

The NRC is also issuing this generic letter to alert licensees to

the problems associated with the material condition of protective

coatings inside the containment and to request information under 10 CFR

50.54(f) to evaluate their programs for ensuring that protective

coatings do not detach from their substrate during a design basis 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.

The NRC expects addressees to ensure that the ECCS and the safety-

related CSS remain capable of performing their intended safety

functions. The NRC will conduct inspections to ensure compliance with

existing licensing bases and respond to discovered inadequacies with

aggressive enforcement consistent with its enforcement policy.

The NRC is seeking comment from interested parties regarding both

the technical and regulatory aspects of the proposed generic letter

presented under the SUPPLEMENTARY INFORMATION heading.

The proposed generic letter was endorsed by the Committee to Review

Generic Requirements (CRGR) on May 5, 1997. The relevant information

that was sent to the CRGR will be placed in the Public Document Room.

The NRC will consider comments received from interested parties in the

final evaluation of the proposed generic letter. The final evaluation

by the NRC will include a review of the technical position and, as

appropriate, an analysis of the value/impact on licensees. Should this

generic letter be issued by the NRC, it will become available for

public inspection in the Public Document Room.

DATES: Comment period expires June 27, 1997. Comments submitted after

this date will be considered if it is practical to do so; assurance of

consideration can only be given for those comments received on or

before this date.

ADDRESSES: Submit written comments to Chief, Rules Review and

Directives Branch, U.S. Nuclear Regulatory Commission, Washington, DC

20555. Written comments may also be delivered to 11545 Rockville Pike,

Rockville, Maryland, from 7:30 am to 4:15 pm, Federal workdays. Copies

of written comments received may be examined at the NRC Public Document

Room, 2120 L Street, NW, (Lower Level), Washington, DC.

FOR FURTHER INFORMATION CONTACT: Richard M. Lobel (301) 415-2865 or

James A. Davis (301) 415-2713.

SUPPLEMENTARY INFORMATION:

NRC Generic Letter 97-XX: Potential for Degradation of the Emergency

Core Cooling System and the Containment Spray System After a Loss-of-

Coolant Accident Because of Construction and Protective Coating

Deficiencies and Foreign Material in the Containment

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

[[Page 26332]]

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 potentially degrade the ECCS or

safety-related CSS. No actions or information are requested regarding

these issues. The NRC has issued many previous generic communications

on this subject, as discussed later in this generic letter, and expects

the addressees to have considered possible actions at their facilities

to address these concerns.

The NRC is also issuing this generic letter to alert the addressees

to the problems associated with the material condition of protective

coatings inside the containment and to request information under 10 CFR

50.54(f) to evaluate the addressees' programs for ensuring that

protective coatings 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.

The NRC expects addressees to ensure that the ECCS and the safety-

related CSS remain capable of performing their intended safety

functions. The NRC will conduct inspections to ensure compliance with

existing licensing bases and respond to discovered inadequacies with

aggressive enforcement consistent with its enforcement policy.

Background

Foreign Material Exclusion, Construction Deficiencies and Design

Deficiencies

In some recent events, 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 reviewed the history

of such events and identified several related problems.

These events are discussed in Appendix A to this generic letter. A

more complete list of the previous events is provided in Appendix B. As

discussed in Appendix A, almost all of these events have been the

subject of previous NRC generic communications and licensee event

reports (LERs). The following types of problems continue to occur.

(1) Foreign material has been found in areas of the containment

where it could be transported to the sump(s) or the suppression pool

and potentially affect the operation of the ECCS or safety-related CSS.

Such material has also been found in PWR sumps, in BWR suppression

pools and downcomers, and in safety-related pumps and piping.

(2) Deficiencies have been found in the construction of the ECCS

sumps or strainers. These deficiencies, which could have impaired the

operation of the ECCS or the safety-related CSS, include missing

screens, unintended openings in screens, and screens that are

incorrectly sized.

(3) Problems have also been found with the material condition of

sumps or suction strainers, potentially impairing the operation of the

ECCS or safety-related CSS. These problems include deformed suction

strainers and unintentional flow paths created by missing grout.

(4) Design deficiencies have been found, including valves in flow

lines with clearances smaller than the sump screen mesh size and

strainers with a flow area smaller than required.

(5) There have been two incidents, described in LERs, in which

doors to emergency sump structures were left open when ECCS and safety-

related CSS operability was required by the technical specifications.

The Discussion section of this generic letter discusses the

regulatory and safety basis for these concerns.

It is evident that past NRC generic communications have not been

completely effective in achieving an acceptable level of control of

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 the existing plant licensing

basis and to respond to discovered inadequacies with aggressive

enforcement consistent with the NRC enforcement policy.

Protective Coatings

Protective coatings inside nuclear power plant containments serve

three general purposes. Protective coatings are applied to steel,

aluminum, and galvanized surfaces to control corrosion. Protective

coatings are applied to surfaces to control radioactive contamination

levels. Protective coatings are also applied to protect surfaces from

erosion and wear.

Protective coatings inside the containment and the regulatory

requirements and guidance for their use are discussed in Appendix C.

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. Not all coatings inside the containment are qualified. The

amount of unqualified coatings must be limited since the unqualified

coatings are assumed to detach from their substrates during a DB LOCA

or steam line break and may be transported to the emergency sump

screens or suction strainers.

In some cases, coatings which should have been qualified failed

during normal operation. Some of these events are discussed in Appendix

D.

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. This criterion must be

demonstrated while assuming the most conservative single failure. Some

addressees may credit CSSs for pressure and radioactive source term

reduction as part of the licensing basis. These CSSs may also take

suction from the suppression pools or emergency sumps.

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.

Qualified coatings used inside containment must be demonstrated to

be capable of withstanding the environmental conditions of a postulated

DB LOCA without detaching from their substrates (detached coatings may

then be transported to the sumps or strainers and cause or contribute

to flow blockage). The LERs and NRC inspection reports described in

Appendix D of this generic letter provide evidence of weaknesses in

addressee programs with regard to applications of protective coatings

for Class I service. 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 paint before its application, (3) control the dry film thickness of

coatings applied to the substrate, (4) monitor for and control the

[[Page 26333]]

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 various

aspects of 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 Appendix E. The basic safety concern

applies to both PWRs and BWRs. These events, discussed in these generic

communications, as well as 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. 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.

The requirements of 10 CFR Part 50, Appendix B, are germane to this

issue.

The maintenance rule, 10 CFR 50.65, ``Requirements for monitoring

the effectiveness of maintenance at nuclear power plants,'' 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.

The NRC expects all addressees to have programs and procedures in

place to ensure that the ECCS and the safety-related CSS are not

degraded by foreign material in the containment, that the ECCS and the

safety-related CSS are consistent with their design and licensing

bases, and that sumps, strainers, and coatings are in good material

condition. The staff may evaluate the condition of sumps, strainers and

protective coatings as a part of maintenance rule inspections.

The NRC has conducted numerous inspections in the areas addressed

by this generic letter; for example, the NRC issued Technical

Instruction 2515/125, ``Foreign Material Exclusion Controls,'' on

August 25, 1994. Violations 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''). A list

of significant enforcement actions is provided in Appendix F of this

generic letter. The NRC intends to continue to conduct inspections in

order to ensure compliance with the existing licensing basis and to

respond to discovered inadequacies with aggressive enforcement

consistent with the NRC Enforcement Policy.

The NRC will consider violations in this area as significant

regulatory failures and will, accordingly, consider categorizing

inadequacies at least as Severity Level III violations. The NRC will

also consider the long history of generic communications on this issue

as prior notice to licensees when the agency assesses 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-

specifc FME program or preparing a response to the requested

information it is determined that a facility is not in compliance with

the Commission's rules or regulations, the addressees are expected to

take whatever actions are deemed appropriate in accordance with

requirements stated in Appendix B to 10 CFR 50 and as required by the

plant technical specifications to restore the facility to compliance.

Required Information

Within 75 days of the date of this generic letter, addressees are

required to submit a written response that includes the following

information:

(1) A summary description of the plant-specific program implemented

to ensure that Class I protective coatings used inside the containment

are procured, applied, and maintained in compliance with applicable

regulatory requirements and the plant-specific licensing basis for the

facility. Include a discussion of how the plant-specific program meets

the applicable criteria of 10 CFR Part 50, Appendix B, as well as

information regarding any applicable standards, plant-specific

procedures or other guidance used for (a) Controlling the procurement

of coatings and paints used at the facility; (b) the qualification

testing of protective coatings; and (c) surface preparation,

application, surveillance, and maintenance activities for protective

coatings.

(2) Information demonstrating compliance with your plant-specific

licensing basis related to tracking the amount of unqualified coatings

inside the containment and for assessing the impact of potential

coating debris on the operation of safety-related SSCs during a

postulated DB LOCA.

Include the following information in the discussion to the extent

it is available:

(a) The date and findings of the last assessment of coatings, and

the planned date of the next assessment of coatings

(b) The limit for the amount of unqualified protective coatings

allowed in the containment and how this limit is determined. Discuss

any conservatisms in the method used to determine this limit.

(c) If a commercial-grade dedication program is being used at your

facility for dedicating commercial-grade coatings for Class I

applications inside the containment, describe why the program is

sufficient to qualify such a coating for Class I service. Identify what

standards or other guidance are currently being used to dedicate

containment coatings at your facility.

(d) If a commercial-grade dedication program is not being used at

your facility for qualifying and dedicating commercial-grade coatings

for use inside containment for Class I applications, provide the

regulatory and safety basis for not controlling these coatings in

accordance with such a program. Additionally, explain why the

facility's licensing basis does not require such a program.

Address the required written information to the U.S. Nuclear

[[Page 26334]]

Regulatory Commission, ATTN: Document Control Desk, Washington, DC

20555-0001, under oath or affirmation under the provisions of Section

182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). This

information will enable the Commission to determine whether the 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 Part 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 of or additions to systems, structures, and

components or to design or operation of an addressee's facility. It

also does not impose an interpretation of the Commission's rules that

is either new or different from a previous staff position. The staff

has, therefore, not performed a backfit analysis.

Reasons for Information Request

This generic letter transmits an information request 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. Specifically, the requested

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 respective

facilities and whether the regulatory requirements pursuant to 10 CFR

50.46 are 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 make the safety-related CSS unable to satisfy the

plant-specific licensing basis by controlling containment pressure and

radioactivity following a LOCA.

Appendix A--Discussion of Events Related to ECCS Sumps and Strainers

Including 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 inside 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

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. This 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

[[Page 26335]]

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 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 off; 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 could have

detrimental effects on 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 an item of debris

larger than the \3/8\-inch diameter of the holes in the containment

spray nozzle in a pipe in the sump.

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 safety injection (SI) to cold-leg 5.08-cm (2-in.) manual

throttle valves were less than the 0.673-cm (0.265-inch) 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 the

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. 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, were 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 since

the ``accessibility of the pedestal area downcomers requires removal

of grating in the undervessel 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, the licensee for Diablo Canyon Nuclear Power

Plant, Unit 1, in LER 89-014-01, discussed the opening of the sump

access hatch at various times at power ``without adequate

consideration of ECCS operability.'' LER 96-006 (Watts Bar Nuclear

Plant, Unit 1) reported that an operator observed a containment sump

(trash screen) door open while ECCS operability was required.

Appendix B--Operational Events Involving ECCS and Safety-Related

Containment Spray Recirculation Flow Paths

------------------------------------------------------------------------

Plant/report Problems discussed

------------------------------------------------------------------------

Haddam Neck NRC Inspection Six 55 drums of sludge with varying

Report 50-213/96-08. amounts of debris removed from ECCS sump

(July 1975).

North Anna Units 1 and 2 LER Galvanized ductwork painted with

84-006-00. unqualified paint.

Millstone Unit 1 LER 88-004- Existing suction strainers smaller than

00. allowed by criteria of RG 1.82 Rev.1.

Strainers will be replaced with larger

strainers if Integrated Safety

Assessment Program criteria met.

[[Page 26336]]

Surry Power Station Units 1 1. Foreign material from construction

and 2 LER 88-017-01 IN 88-87 activities found in cone strainer of

IN 89-77. recirculation spray system. Material

could have rendered system inoperable.

2. Gaps in sump screens since initial

construction.

Trojan Nuclear Plant LER 89- 1. Wire mesh screen on top of sump trash

016-01 IN 89-77. rack not installed.

2. Screen damage.

3. Significant amount of debris

discovered in the sump. Could have

caused loss of a portion of ECCS.

Diablo Canyon Unit 1 LER 89- 1. Debris in sump.

014-01 IN 89-77. 2. As-built sump configuration not in

accordance with design.

3. Safety function would not have been

impaired.

TMI Unit 1...................

LER 90-002-00 Modification of sump access hatches left

holes in top of sump screen cage.

Potentially could damage pumps or clog

spray nozzles.

McGuire Unit 1 LER 90-0112-00 Loose material discovered in upper

containment prior to entry into Mode 4.

Items found would not have made ECCS

inoperable.

Calvert Cliffs Units 1 and 2 Unit 2 sump found to contain 25 lbs dirt,

NRC Inspection Report March weld slag, pebbles, etc. Inspection of

5, 1991. Unit 1 found less than 1 lb. debris.

Possible minor damage to ECCS pumps.

Diablo Canyon Unit 2 LER 91- 1. Numerous instances of material left

012-00. unattended or abandoned in sump level of

containment (tools, plastic tool bags,

clothing, etc.).

2. Material would not have prevented ECCS

recirculation function.

H.B. Robinson Unit 2 LER 92- ``B'' safety injection pump reduced flow

013-00. due to blockage in minimum flow

recirculation check valve and flow

orifice on July 8, 1992. ``A'' pump OK.

Foreign material also found in refueling

water storage tank (RWST).

H.B. Robinson Unit 2 LER 92- On August 24, 1992, following a reactor

018-00. trip, ``A'' and ``B'' safety injection

pumps inoperable due to reduced flow.

Found during unscheduled surveillance to

demonstrate safety injection (SI)

operability.

Pt. Beach Unit 2 LER 92-003- September 18, 1992: During technical

01 IN 92-85. specifications (inservice) testing of

the ``A'' containment spray pump, the

pump was declared inoperable. A foam

rubber plug was blocking pump suction.

Plug removed and pump tested

satisfactorily. One train of Unit 2

residual heat removal, safety injection

and containment spray systems inoperable

for entire operating cycle. Plug was

part of a cleanliness barrier.

Perry Nuclear Plant LER 93- May 1992: During refueling outage foreign

011-00. objects discovered in the containment

side of the suppression pool. Fouling of

residual heat removal (RHR) strainers

found. Strainers not cleaned.

January 1993: RHR ``A'' and ``B''

strainers found deformed (collapsed

inward in the direction of the fluid

flow. Strainers replaced.

March 1993: RHR ``A'' and ``B'' operated

in suppression pool cooling mode. Pump

suction pressure decreased. Could have

compromised long-term RHR operation.

Susquehanna Units 1 and 2 LER 1. Assessing impact of debris and

93-007-00 (Voluntary). corrosion products adhering to fibrous

materials that may be dislodged by a

pipe break.

2. Developing procedures to backflush

strainers.

Sequoyah Unit 2 LER 93-026-00 Design basis limit for unqualified

coatings inside containment had been

exceeded. Additional quantity of

unqualified coatings on reactor coolant

pump motor platform discovered. Path to

ECCS sump. Screens will be installed

before startup.

ANO Unit 2 LER 93-002-00 IN Seven unscreened holes found in masonry

89-77 Supplement 1. grout below screen assembly of ECCS

sump. Could potentially degrade both

trains of the high pressure coolant

injection system and containment spray.

Had previously inspected sump because of

IN 89-77. Did not discover problem. NRC

estimate of incremental increase in core

damage: 3 x 10-04.

ANO Unit 1 LER 93-005-00 IN 1. 22 unscreened 6 x 3 pipe openings at

89-77 Supplement 1. base of sump curb. Occurred as a result

of modification prior to initial

operation.

2. Tears in screen.

3. Floor drains leading to sump not

screened.

4. Licensee estimated increase in core

damage frequency 5 x 10-05.

San Onofre Units 1 and 2 LER 1. Irregular annular gap (approximately

93-010-00 (Voluntary). 6) surrounding 8 low temperature

overpressure protection system discharge

line penetrating horizontal steel cover

plate.

2. Engineering analysis concluded both

sump trains operable.

Vermont Yankee LER 93-015-00. 1. Low pressure core spray suction

strainers smaller than calculations

assumed. Net positive suction head

calculations performed in 1986 following

change to NUKONTM insulation invalid.

2. Strainers replaced with larger

strainers.

South Texas Unit 1/2 LER 94- 1. Sump screen openings from initial

001-00. construction discovered. Frame plate at

floor warped, creating several openings

approximately \5/8\''. Additional \1/

4\'' gaps discovered. Licensee concluded

there was no safety significance to

these deficiencies based on ECCS pump

tests performed by the manufacturer.

Point Beach Unit 1 NRC NRC inspector found grout deterioration

Inspection Report May 6, under sump screens. Could result in flow

1994. bypass or particles of grout entering

ECCS pumps.

LaSalle Unit 1 IN 94-57...... April 26 and May 11, 1994: Divers

inspecting suppression pool during

outage found operational debris.

River Bend IN 94-57.......... June 13, 1994: Plant in refueling outage.

Foreign material found in suppression

pool. Plastic bag removed from ``B'' RHR

pump suction strainer. Other objects:

tools, grinding wheel, scaffolding

knuckle, step off pad.

[[Page 26337]]

Quad Cities Unit 1 IN 94-57.. July 14, 1994: Post-maintenance test of

``A'' loop RHR indicated a plugged torus

cooling test return valve. Inspection

discovered remains of shredded plastic

bag in anti-cavitation trim installed

during a recent outage.

July 23, 1994: 4'' diameter wire brush

and a piece of metal found wrapped

around a vane of the ``C'' RHR pump.

Browns Ferry Units 1/2/3 May 1. Unqualified coatings on T quenchers in

20, 1994 Letter to NRC. suppression pool.

2. Continued operation acceptable.

3. Will remove coatings next refueling

outage.

Palisades Plant LER 94-014-00 Signs, adhesive tape, and labels with

potential to block the ECCS sump were

found in containment. Containment spray

and HPSI pumps declared inoperable.

Engineering analysis concluded that the

sump screen would not be significantly

blocked.

Watts Bar Units 1 and 2 NRC Screens installed around reactor coolant

Inspection Report 50-390 and pump motors to catch unqualified paint

50-391/94-59 September 28, not adequately located to contain all

1994. unqualified coatings.

Indian Point Unit 2 LER 95- Licensee discovered portions of floor

005-00. coating on containment Elevation 46 had

lifted and cracked. In other locations,

floor coating cracked when stepped on.

Licensee concluded that sump function

would not be compromised.

Susquehanna Units 1 and 2 LER Licensee took actions to address clogging

93-007-001 September 11, ECCS suction strainers: removal of

1995. fibrous insulation from high energy line

break areas, testing to characterize the

debris threat to strainer blockage,

quantification of corrosion products on

structural steel in wetwell,

establishment of a comprehensive

analysis of containment debris effects.

Coating and insulation procedures

contain steps to reduce potential for

strainer blockage.

Prairie Island Unit 2 NRC Broken labels for pipe hangers and labels

Inspection Report 50-282/05- affixed to wall with degrading adhesive

009. discovered by NRC inspector after

licensee closeout inspection. Licensee

concluded that this would not affect

operability of ECCS.

Palisades NRC Inspection Unsecured material stored on the landings

Report 50-225/95-008. of stairways. Broken glass and pieces of

signboard and other ``unauthorized''

material found in area designated debris-

free.

Limerick Unit 1 NRC Debris was allowed to collect in

Inspection Report 50-352/96- suppression pool so that ``A'' RHR pump

04. was rendered inoperable when safety/

relief valve lifted on September 11,

1995.

Duane Arnold NRC Inspection Foreign material exclusion controls

Report 50-331/95-003. inadequate in drywell. Hardhats and

debris noted.

Foreign PWR NRC IN 96-10..... 1. Operator found debris in the sump.

2. Two of 4 ECCS lines taking suction

from the sump were partially blocked by

debris. Debris present since plant

construction.

Millstone Unit 2 LER 96-008.. Ten locations inconsistent with the

specified screen opening size were

identified. Placed plant outside

original design basis. Sump screen

replaced.

Watts Bar Unit 1 LER 96-006- Operator observed containment sump trash

00. screen door was open when plant was in

MODE 4 and ECCS required to be operable.

Calvert Cliffs Units 1 and 2 Several holes identified in each units'

LER 96-003-00. containment sump screen larger than

described in the Final Safety Analysis

Report. Holes field-installed for

transmitter tubing. Concluded not a

threat to plant safety.

Diablo Canyon Unit 1 LER 96- Various debris that could pass through

007-00. the containment sump screen could be

larger than minimum clearances in the

1\1/2\'' centrifugal charging pump

runout protection manual throttle valves

and 2'' SI cold leg manual throttle

valves.

Haddam Neck LER 96-014-00 NRC 1. Discrepancies in sump screen mesh

Inspection Report 50-213/96- sizing, screen fitup, and method of

08. attachment discovered. Sump screen

replaced. Sump will be inspected after

every refueling outage. Licensee

reported that this condition could have

prevented the fulfillment of a safety

function.

2. Five 55-gallon drums of sludge removed

from ECCS sump. Also, plastic, nuts and

bolts, tie wraps, and pencil.

Big Rock Point NRC Inspection ``Housekeeping in containment in the area

Report 50-155/96-004. under the emergency condenser and the

reactor depressurization system

isolation valves was poor.''

Catawba Unit 1 NRC Inspection Six floor drains inside crane wall were

Report 50-413/96-11. not covered with screen that had a finer

mesh than the sump screen. The holes

were \1/4\'' rather than \1/8\'' holes.

Crane wall penetrations close to

containment floor could allow the

transport of debris to the sump screen.

Penetrations sealed.

Millstone Unit 2 LER 50-336/ Containment sump screens had been

96-08 NRC Inspection Report incorrectly constructed so that larger

50-336/96-08. debris than analyzed could pass through

the ECCS.

Vogtle Unit 2 NRC Inspection Containment integrity was established

Report 50-425/96-11 LER 96- prior to startup. Upon subsequent

007-00. containment entries personnel discovered

various items of loose debris. Material

removed while in MODE 4. Material would

have resulted in inadequate NPSH for the

``B'' train of RHR and containment

spray. NPSH for the ``A'' train of RHR

and containment spray would have been

adequate.

Nine Mile Point Unit 2 NRC A significant amount of debris was found

Inspection Report 50-410/96- in the suppression pool and downcomers

11 NRC Event Report 31172. during refueling outage 5. The

licensee's preliminary evaluation

concluded that operability of ECCS could

have been compromised.

LaSalle Unit 2 NRC Event Substantive foreign material recovered

Report 31159 LER 96-009-00. from suppression pool and downcomers

which would challenge the operability of

the ECCS. Items most likely from

construction or early outages.

Millstone Unit 3 LER 96-039- 1. Construction debris discovered in

00. containment recirculation spray system

(RSS) containment sump and in RSS

suction lines.

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.

[[Page 26338]]

H.B. Robinson Unit 2 LER 96- 1. Openings found in sump screens that

005-00. could allow debris above a certain size

to enter the sump. Could have prevented

the screens from performing their design

function.

2. An item of debris in excess of \3/8\''

diameter limit on containment spray

nozzles found in 14'' sump drain pipe.

Zion Unit 1 LER 97-001-00.... Two 1-inch holes were not in the sump

cover as detailed on drawings. Holes

allow air to escape as sump fills.

Potential to hinder flow to RHR pump

suction during a LOCA.

Zion Unit 2 NRC Inspection 1. Miscellaneous debris located

Report 50-295/96-20 50-304/ throughout containment.

96-20 March 24, 1997. 2. Containment recirculation sump screen

damage.

3. Peeling and flaking paint on

containment surfaces.

Sequoyah Unit 1 10 CFR 50.72 During shutdown on March 22, 1997, an oil

Report 32139 April 11, 1997. cloth was introduced to containment

which, if it had come free of its

restraints, could have blocked one or

both refueling drains so that water in

upper containment may not have flowed

freely to lower level of containment

where sump is located.

Millstone Unit 1 10 CFR 50.72 Most of the coating in the torus is

Report 32161 April 16, 1997. unqualified, which could affect the

operability of the low-pressure coolant

injection and core spray systems.

------------------------------------------------------------------------

Appendix C--Background On Regulatory Basis for Protective Coatings

This appendix discusses the regulatory basis for protective

coatings inside the containment. Industry standards and regulatory

guidance are included in this discussion. However, this discussion

is only for information. Addressees should continue to comply with

the plant licensing basis.

At nuclear power plants, coatings and paints serve to (1)

protect ferritic steel, austenitic steel, galvanized (zinc-coated)

steel, or aluminum surfaces against corrosive environments; (2)

protect metallic, concrete, or masonry surfaces against erosion or

wear during plant operation; and (3) allow for ease of

decontamination of radioactive nuclides from the containment wall

and floor surfaces. These coatings may come in inorganic forms, such

as zinc-based paints, or organic forms, such as organic latex,

polyurethane, or epoxy coatings.

There are two kinds of coatings applications at domestic nuclear

power plants:

(1) Class I Service Applications, which are applications of

coatings or paints to SSCs that are essential to prevent or mitigate

the consequences of postulated accidents. Protective coatings

applied to the interior wall and floor surfaces of the containment

structure and to the exterior surfaces of most of the SSCs located

inside the containment structure normally fall into this

category.1

---------------------------------------------------------------------------

\1\ Coatings applied to non-safety-related small-scale

components inside the containment structure, such as small lighting

fixtures or small non-safety-related power buses, are an exception

to this statement.

---------------------------------------------------------------------------

(2) Class II Service Applications, which are applications of

coatings or paints to SSCs that are essential to the achievement of

normal operating performance.

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 demonstrated the capability of maintaining 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 standard procedures (such as those issued by the

American National Standards Institute, Inc. [ANSI], or the American

Society for Testing and Materials [ASTM]) for coatings applications.

However, the licensing basis for Class I coating applications may

contain exceptions to or provide alternative means of meeting the

intent of the test methods in these standards, provided an adequate

safety basis was given to and accepted by the NRC staff as to why

accepting the exceptions or alternatives could not have the

potential to affect the performance of the ECCS and safety-related

CSS during a postulated DB LOCA. In regard to protective coatings

used for Class 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

Class I service applications are implemented in accordance with a

quality assurance (QA) program that meets the requirements of

Appendix B to Part 50 of Title 10 of the Code of Federal Regulations

(10 CFR).

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 or strainers and their effect on the operability

of safety-related pumps taking suction from ECCS sumps and

suppression pools during a postulated DB LOCA.

The NRC has issued Regulatory Guide (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

ferritic steel, aluminum, stainless steel, zinc-coated (galvanized)

steel, or 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 subject to

the additional regulatory positions 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. Although neither RG 1.54-1973 nor the applicable

ANSI standards are NRC requirements, they do delineate acceptable

programs and practices for controlling coatings 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,

including activities for controlling work conditions, for

controlling the ambient environmental conditions for coating

applications, for controlling selection and procurement activities

for coatings, for controlling preparation of substrates, for

establishing QA procedures for coating applications, for qualifying

personnel involved in coating preparation, application, and

inspection activities, and for establishing coating inspection

guidelines and acceptance criteria. The scope of ANSI Standard

N101.4-1972, as endorsed by RG 1.54-1973, also includes recommended

QA records on coatings activities.

[[Page 26339]]

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 for

governing activities related to the selection and evaluation of

protective coatings applied both 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 replace 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. Utilizing more modern

industry standards for protective coatings may require a change to

the existing licensing basis. Use of these standards must conform

with existing NRC requirements, including 10 CFR 50, Appendix B.

Appendix D--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 located 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 m \2\

(6000-7000 ft \2\) 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 by removing all of the additional ``unqualified'' paints that

were determined to be located within the analytically determined

zone of influence.2 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 tests exhibited

``pulls'' 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.

---------------------------------------------------------------------------

\2\ All of the unqualified paint within the containment sump's

zone of influence was removed, with the exception of approximately

112 ft \2\ of unqualified paint applied to small components, such as

lighting fixtures or name tags.

---------------------------------------------------------------------------

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-meter (46-ft)

elevation of the Indian Point Unit 2 containment structure. The

paint was applied to the 14-meter (46-foot) 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-foot) 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 reindoctrinate 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 are 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,

respectively, to summarize its reassessment of ECCS performance at

Susquehanna Steam Electric Station, Units 1 and 2, 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 would not have the potential to 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 Units 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

[[Page 26340]]

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.

Appendix E--Generic Communications Issued by the NRC on the Subject of

ECCS and Safety-Related CSS Sump and Strainer Blockage

Generic Letter 85-22,''Potential for Loss of Post LOCA

Recirculation Capability Due to Insulation Debris Blockage,''

December 3, 1985.

IN 88-28, ``Potential for Loss of Post LOCA 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 LOCA 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 LOCA

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 LOCA 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.

Appendix F--Enforcement Actions Taken by the NRC Dealing With

Construction and Protective Coatings Deficiencies and Foreign Material

Exclusion

----------------------------------------------------------------------------------------------------------------

Severity

Plant Date of inspection level/civil Description

penalty

----------------------------------------------------------------------------------------------------------------

Surry Unit 1............................ 7/30/88................. 3 Debris in containment sump.

$50,000

Trojan.................................. 8/8/89.................. 2 Inoperable recirculation sump.

$280,000

Diablo Canyon........................... 12/8/89................. 3 1. Gaps in sump screens

$50,000 2. Opening sump access hatches

when sump operability is

required

3. Debris in sump.

Perry................................... 6/23/93................. 3 Clogged RHR strainers.

$200,000

Arkansas Nuclear One Unit 1............. 10/25/93................ 3 Degradation of containment sump

$0 screens.

Browns Ferry Unit 2..................... 5/17/94................. 4 Unqualified protective coatings

$0 applied to safety/relief valve

discharge quenchers.

Point Beach Unit 2...................... 10/12/92................ 3 Foreign material in containment

$75,000 spray.

Sequoyah Units 1 and 2.................. 9/3/94.................. 4 Unqualified coatings on RCP

$0 motor stand.

Nine Mile Point Unit 2.................. April 10, 1997 *........ 3 Debris in suppression pool and

** $200,000 downcomers.

----------------------------------------------------------------------------------------------------------------

* Date enforcement action issued.

** Combined with other enforcement actions.

Dated at Rockville, Maryland, this 8th day of May, 1997.

For the Nuclear Regulatory Commission.

Marylee M. Slosson,

Acting Director, Division of Reactor Program Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 97-12467 Filed 5-12-97; 8:45 am]

BILLING CODE 7590-01-P