[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
[[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
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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