[Federal Register Volume 62, Number 92 (Tuesday, May 13, 1997)]
[Notices]
[Pages 26331-26340]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-12467]


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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

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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

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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