[Federal Register Volume 62, Number 34 (Thursday, February 20, 1997)]
[Notices]
[Pages 7806-7809]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-4175]
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NUCLEAR REGULATORY COMMISSION
Proposed Generic Communication; Assurance of Sufficient Net
Positive Suction Head for Emergency Core Cooling and Containment Heat
Removal Pumps (M96537)
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 that will request addressees to submit the analysis
and pertinent assumptions used to determine the net positive suction
head (NPSH) available for emergency core cooling (including core spray
and decay heat removal) and containment heat removal pumps. This
information will enable the NRC to determine if the NPSH analyses for
reactor facilities are consistent with their respective current
licensing basis. 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 has been endorsed by the Committee to
Review Generic Requirements (CRGR). The relevant information that was
sent to the CRGR will be placed in the NRC Public Document Room. The
NRC will consider comments received from interested parties in the
final evaluation of the proposed generic letter. The NRC's final
evaluation 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 NRC Public Document Room.
DATES: Comment period expires March 24, 1997. Comments submitted after
this date will be considered if it is practical to do so, but assurance
of consideration cannot be given except for comments received on or
before this date.
ADDRESSEES: Submit written comments to Chief, Rules Review and
Directives Branch, U.S. Nuclear Regulatory Commission, Mail Stop T-6D-
69, Washington, DC 20555-0001. 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, N.W., (Lower Level),
Washington, DC.
FOR FURTHER INFORMATION CONTACT: Howard (Jack) Dawson, (301) 415-3138.
SUPPLEMENTARY INFORMATION:
NRC GENERIC LETTER 97-XX: ASSURANCE OF SUFFICIENT NET POSITIVE SUCTION
HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL PUMPS
Addressees
All holders of operating licenses for nuclear power plants, except
those who have certified to a permanent cessation of operations.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this
generic letter (GL) to request that addressees submit the analysis and
pertinent assumptions used to determine the net positive suction head
(NPSH) available for emergency core cooling (including core spray and
decay heat removal) and containment heat removal pumps. This
information will enable the NRC to determine if the NPSH analyses for
reactor facilities are
[[Page 7807]]
consistent with their respective current licensing basis.
Background
As a result of recent NRC inspection activities, licensee
notifications, and licensee event reports, a safety-significant issue
has been identified that has generic implications and warrants action
by the NRC to ensure that the issue has been adequately addressed and
resolved. The issue is that the NPSH available for emergency core
cooling system (ECCS) (including core spray and decay heat removal) and
containment heat removal pumps may not be adequate under all design-
basis accident scenarios. In some cases, this may be a result of
changes in plant configuration, operating procedures, environmental
conditions or other operating parameters that have taken place over the
life of the plant.
In other cases, the licensing analysis may not bound all postulated
events for a sufficient time, or assumptions used in the analysis may
be non-conservative or inconsistent with those assumptions and
methodologies traditionally considered acceptable by the staff. For
example, some licensees have recently discovered that they must take
credit for containment overpressure to meet ECCS (including core spray
and decay heat removal) and containment heat removal pump NPSH
requirements. In the examples the NRC staff is familiar with, the need
for crediting this overpressure in ECCS analyses has arisen due to
changes in plant configuration and operating conditions which have
occurred over the life of the plant, and/or errors in prior NPSH
calculations. The overpressure being credited by licensees may be
inconsistent with the licensing basis of the plant.
The current NPSH analyses (including any corresponding containment
pressure analysis) may not be available to the staff in docketed
material (e.g., final safety analysis reports) because some licensees
have changed their analyses. Consequently, this generic letter requests
that addressees submit the analyses and pertinent assumptions used to
determine the NPSH available for emergency core cooling (including core
spray and decay heat removal) and containment heat removal pumps. This
generic letter applies only to ECCS (including core spray and decay
heat removal) and containment heat removal pumps that take suction from
the containment sump or suppression pool following a loss-of-coolant
accident (LOCA) or secondary line break.
New NPSH analyses are not required or requested to respond to this
information request. However, new NPSH analyses may be warranted if an
addressee determines that a facility is not in compliance with the
Commission's rules and regulations. In such cases, the affected
addressees are expected to take corrective action, as appropriate, in
accordance with the requirements stated in 10 CFR part 50, appendix B,
to restore their facility to compliance.
The following is a sample of the NRC staff's recent findings
concerning the NPSH issues addressed by this generic letter:
Haddam Neck
In 1986 and 1995, the licensee identified conditions where the NPSH
available for the residual heat removal (RHR) pumps may be insufficient
when the pumps are operating in the emergency core cooling mode. In
1986, the licensee determined that the only extant NPSH analysis, which
was performed in 1979 as part of the Systematic Evaluation Program, did
not properly account for hydraulic losses in suction piping, and as a
result, erroneously indicated that containment overpressure was not
needed to satisfy NPSH requirements for the pumps in the recirculation
mode of operation. A new analysis showed that credit had to be taken
for 6 psi of containment overpressure. In another reanalysis conducted
in 1995 for increased service water temperature, the licensee found
that additional containment overpressure, which constituted a
significant fraction of the peak calculated containment accident
pressure, was necessary to meet NPSH requirements for the same pumps.
On August 30, 1996, the licensee reported in Licensee Event Report
(LER) 96-016 that calculations recently performed to determine the NPSH
available for the residual heat removal pumps may have been in error
for the alternate, short-term recirculation flow path, due to
insufficient containment overpressure for a period of pump operation.
The licensee attributed this event to the failure to fully analyze the
containment pressure and sump temperature responses under design-basis
accident conditions.
Maine Yankee
During an inspection conducted in July and August 1996, to
determine if Maine Yankee was in conformance with its design and
licensing bases, an NRC Independent Safety Assessment Team (ISAT)
identified potential weaknesses in the licensee's containment spray
pump NPSH analysis. These potential weaknesses included concerns
regarding the validity of the containment sump temperature analysis,
incorrect calculation of bounding pump suction head losses, and use of
a hot fluid correction factor to reduce NPSH requirements. The
licensee's calculation of record, performed in 1995 and which does not
include the hot fluid correction factor, indicates a condition in which
the available NPSH for the containment spray pumps would be below the
required NPSH for the first 5 minutes after pump suction is switched
from the refueling water storage tank to the recirculation sump. This
analysis was performed for a power level of 2700 thermal megawatts
(MWt). When the hot fluid correction factor was used, the NPSH
available could only be shown to be slightly greater than the NPSH
required for the same 5-minute period. For the remainder of the
transient, the NPSH available to the containment spray pumps was shown
to exceed the amount required.
The basis for the licensee's contention that the containment spray
pumps were operable is that recent pump tests showed that the pumps
could operate for a 15-minute period with NPSH below the required value
without damage to the hydraulic performance or mechanical integrity of
the pumps. The licensee performed another analysis for a power level of
2440 MWt which showed that adequate NPSH margin would exist for the
containment spray pumps in the recirculation mode of operation. This
analysis did not include use of the hot fluid correction factor. The
ISAT concluded that it was appropriate to consider the containment
spray pumps operable at a power level of 2440 MWt. Maine Yankee is
currently prohibited by the NRC from operation above 2440 MWt. The NRC
staff is currently reviewing the licensee's analysis and assumptions in
greater detail.
Pilgrim
The NRC staff's safety evaluation for licensing of the Pilgrim
plant, and documents referenced by the evaluation, indicate that
containment overpressure was not necessary to satisfy RHR and core
spray pump NPSH requirements. When a plant modification was made in
1984, the licensee's safety analysis of the modification stated that
the NPSH available was determined assuming (1) maximum debris loading
conditions on the sump strainers for the residual heat removal and core
spray pumps and (2) no credit for containment over-pressure. On April
14, 1994, in its response to NRC Bulletin 93-02, ``Debris Plugging of
Emergency Core Cooling Suction Strainers'' (March 23, 1993), the
[[Page 7808]]
licensee stated that the NPSH available to the residual heat removal
and core spray pumps was analyzed assuming no overpressure condition in
the torus.
However, in an analysis conducted by the licensee in 1996 in
support of a strainer modification, credit is needed and taken for
containment over-pressure. At the time of this analysis, the licensee
also indicated that the assumption of no overpressure in the torus,
stated in its response to Bulletin 93-02, was incorrect. While the
issue of whether or not credit for over-pressure is part of Pilgrim's
original licensing basis is currently under staff review, the potential
exists that other licensees have made modifications to their plants
that may be inconsistent with their licensing basis and could reduce
the NPSH available to ECCS and core spray pumps.
Crystal River, Unit 3
As part of the NRC's Integrated Performance Assessment of Crystal
River, Unit 3, conducted in July 1996, an NRC inspection team reviewed
the licensee's calculation which established the minimum required post-
LOCA reactor building water level for ensuring adequate NPSH available
for the reactor building spray pumps. When the team compared this level
with the minimum predicted level, they found that for one of the pumps,
there was only a slight difference between the water level available
and the water level required to ensure adequate NPSH during the post-
accident recirculation phase of pump operation.
The team found that the licensee used non-conservative assumptions
in calculating the available NPSH for the spray pump. For example,
uncertainty in data regarding the required NPSH was not accounted for,
a correction factor to reduce the NPSH required was used in the
calculation without considering the effects of non-condensable gases in
the pumped fluid, and uncertainties associated with the hydraulic
resistance of check valves in the spray lines were not fully accounted
for. Conservative assumptions that were included in the calculation
were those detailed in Regulatory Guide (RG) 1.1, ``Net Positive
Suction Head for Emergency Core Cooling and Containment Heat Removal
System Pumps,'' dated November 2, 1970 (originally Safety Guide 1),
regarding the use of maximum reactor building fluid temperature and no
credit for containment overpressure.
The team concluded that the cavitation-free operation of building
spray pump 1B during the recirculation phase of operation is
questionable due to the non-conservative assumptions used in the NPSH
calculation. However, the team also concluded that this issue did not
constitute an immediate safety concern since the licensee's
calculations conservatively assumed no credit for containment
overpressure and use of maximum expected reactor building water
temperature. As a result of the teams findings, the NRC staff is
reviewing the issue of adequate NPSH for the reactor building spray
pumps at Crystal River, Unit 3, in greater detail.
Related Generic Communications
On October 22, 1996, the staff issued Information Notice (IN) 96-
55, ``Inadequate Net Positive Suction Head of Emergency Core Cooling
and Containment Heat Removal Pumps Under Design Basis Accident
Conditions,'' to alert addressees to recent discoveries by licensees
that there may be scenarios for which the NPSH available for emergency
core cooling system and containment heat removal pumps may not be
sufficient. Earlier INs describing similar events include IN 87-63,
``Inadequate Net Positive Suction Head in Low Pressure Safety
Systems,'' dated December 9, 1987, and IN 88-74, ``Potentially
Inadequate Performance of ECCS in PWRs During Recirculation Operation
Following a LOCA,'' issued on September 4, 1988.
Discussion
It is important that the emergency core cooling (including core
spray and decay heat removal) and containment spray system pumps have
adequate NPSH available for all design-basis LOCAs to ensure that the
systems can reliably perform their intended functions under accident
conditions. Inadequate NPSH could cause voiding in the pumped fluid,
resulting in pump cavitation. While some ECCS (including core spray and
decay heat removal) and containment heat removal pumps can operate for
relatively short periods of time while cavitating, prolonged operation
under cavitation conditions for any pump can cause vapor binding,
resulting in reduced pump performance and potential common-mode failure
of the pumps. Common-mode failure would result in the inability of the
emergency core cooling system to provide adequate long-term core
cooling and/or the inability of the containment spray system to
maintain the containment pressure and temperature below design limits.
This generic letter addresses situations in which the NPSH
available for ECCS (including core spray and decay heat removal) and
containment heat removal pumps may be inadequate as a result of
changing plant conditions, and/or errors and non-conservative
assumptions in NPSH calculations. In some cases, NPSH reanalyses
conducted to support plant modifications may result in a substantial
reduction of margin in NPSH available or a change in the original
design basis of the plant. In particular, recent examples have
indicated that containment overpressure has been credited by licensees
to satisfy NPSH requirements in response to changing plant conditions
and errors in prior NPSH calculations.
NRC Regulatory Guide 1.1 establishes the regulatory position that
emergency core cooling and containment heat removal systems should be
designed so that adequate NPSH is provided to system pumps assuming
maximum expected temperatures of pumped fluids and no increase in
containment pressure from that present before any postulated loss-of-
coolant accidents. Standard Review Plan (SRP) 6.2.2, ``Containment Heat
Removal Systems'' (NUREG-0800, Revision 3, July 1981) clarifies RG 1.1
by stating that the NPSH analysis should be based on the assumption
that the containment pressure equals the vapor pressure of the sump
water, to ensure that credit is not taken for containment
pressurization during the transient. As part of licensing and
Systematic Evaluation Plan reviews, the NRC staff has, in the past,
selectively allowed limited credit for a containment pressure that is
above the vapor pressure of the sump fluid (i.e., an overpressure) to
satisfy NPSH requirements on a case-by-case basis.
Requested Information
Addressees are requested to review, for each of their reactor
facilities, the current analyses that are used to determine the
available NPSH for the emergency core cooling (including core spray and
decay heat removal) and containment heat removal pumps which, at any
time following a design-basis accident, take suction from the
containment sump or the suppression pool. No new NPSH analysis is
requested or required. Based on this review, within 60 days from the
date of this generic letter, addressees are requested to provide the
information outlined below for each of their facilities; to the extent
practical, the use of a tabular format is acceptable in presenting the
information.
(1) Provide the NPSH analysis and assumptions for each pump, and,
in particular,
(a) Specify, as a function of time, the required NPSH and the
available NPSH,
[[Page 7809]]
(b) Identify the postulated pipe breaks that were analyzed if a
spectrum of primary and secondary system pipe break sizes and locations
was considered in the NPSH analysis,
(c) Specify the emergency core cooling (including core spray and
decay heat removal) and containment heat removal system configurations
(and associated flow rates) that were considered in the NPSH analysis
for each pump; identify and justify which configurations were not
analyzed,
(d) Specify if the current licensing-basis NPSH analysis is
different from the original licensing-basis analysis, and
(e) Specify any quality assurance procedures and engineering
program controls in place when the current NPSH analysis was performed.
(2) For each pump, specify whether or not containment overpressure,
i.e., containment pressure above the vapor pressure of the sump (or
suppression pool) fluid, was credited in the calculation of available
NPSH. Specify the amount of overpressure needed, and the minimum
overpressure available. Indicate if the overpressure was determined
from the containment pressure at a single point in time, or if the
containment pressure profile over an extended period of time was
considered. If an extended period of time was considered, state how
long and give the rationale for choosing this time period; if only a
single point in time was considered, state the point in time and give
the rationale for selecting this point in time.
(3) When containment overpressure is credited in the calculation of
available NPSH, specify the containment atmosphere heat removal
assumptions that were used in the containment response analysis to
determine the minimum containment overpressure available, and in
particular,
(a) Identify the heat transfer correlations that were used, and
specify whether or not multipliers were used, to calculate the transfer
of energy to the heat sinks in the containment,
(b) Specify how many trains of containment spray were assumed to be
operating, and whether a minimum, maximum, or intermediate value of
spray flow was assumed,
(c) Specify how the service water temperatures for the heat
exchangers that remove energy from the containment atmosphere were
chosen for the NPSH analysis, and specify any special assumptions made
concerning heat transfer across the heat exchangers (e.g., effect of
fouling on heat transfer),
(d) Specify the total number of containment fan coolers at the
plant, and specify how many fan coolers were assumed to be operating.
Required Response
Within 30 days from the date of this generic letter, each addressee
is required to submit a written response indicating (a) whether or not
the requested information will be submitted, and (b) whether or not the
requested information will be submitted within the requested time
period. Addressees who choose not to submit the requested information,
or are unable to satisfy the requested completion date, must describe
in their response an alternative course of action that is proposed to
be taken, including the basis for the acceptability of the proposed
alternative course of action.
New NPSH analyses are not required or requested to respond to this
information request. However, new NPSH analyses may be warranted if an
addressee determines that a facility is not in compliance with the
Commission's rules and regulations. In such cases, the affected
addressees are expected to take corrective action, as appropriate, in
accordance with the requirements stated in 10 CFR part 50, appendix B,
to restore their facility to compliance.
NRC staff will review the responses to this generic letter and if
concerns are identified, affected addressees will be notified.
Address the required written response to the U.S. Nuclear
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).
Backfit Discussion
This generic letter only requests information from addressees under
the provisions of section 182a of the Atomic Energy Act of 1954, as
amended, and 10 CFR 50.54(f). The information requested will enable the
staff to determine whether addressees' NPSH analyses for the emergency
core cooling (including the core spray and decay heat removal) and
containment heat removal system pumps comply and conform with the
current licensing basis for their respective facilities, including the
licensing safety analyses and the principle design criteria which
require and/or commit that safety-related components and systems be
provided to mitigate the consequences of design-basis accidents.
With respect to the principle design criteria for nuclear power
reactor facilities, which establish minimum requirements for
structures, systems, and components important to safety, General Design
Criterion (GDC) 35 of appendix A to Title 10 of the Code of Federal
Regulations (10 CFR part 50, appendix A) specifies that there be a
system to provide abundant emergency core cooling. Furthermore, 10 CFR
50.46, which addresses the acceptance criteria for emergency core
cooling systems for light water nuclear power reactors, requires, in
part, that the emergency core cooling system be able to provide long-
term cooling following any loss-of-coolant accident. The potential for
the loss of adequate NPSH for emergency core cooling system pumps, and
the cavitation that would result, raises the concern that the emergency
core cooling system would not be capable of providing core cooling over
the duration of postulated accident conditions as required by GDC 35
and 10 CFR 50.46.
Similarly, GDC 38 of appendix A to 10 CFR part 50 specifies that
there be a system to rapidly remove heat from the reactor containment
in order to reduce the containment pressure and temperature following
any loss-of-coolant accident, and GDC 16 of appendix A to 10 CFR part
50 specifies that reactor containment and associated systems be
provided to assure that the containment design conditions important to
safety are not exceeded for the duration of the accident conditions.
The potential for the loss of adequate NPSH in containment spray pumps,
and the cavitation that would result, raises the concern that
containment spray would not be capable of lowering and maintaining the
containment pressure and temperature below design values as required by
GDC 38 and GDC 16.
Considering the safety significance of removing heat from the
containment atmosphere and cooling the reactor core following a design-
basis accident, the requested information is needed to verify addressee
compliance with licensing basis commitments regarding the performance
of emergency core cooling (including core spray and decay heat removal)
system and containment heat removal system pumps. The evaluation
required by 10 CFR 50.54(f) to justify this information request is
included in the preceding discussion.
Dated at Rockville, Md., this 11th day of February 1997.
For the Nuclear Regulatory Commission.
Thomas T. Martin,
Director, Division of Reactor Program Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 97-4175 Filed 2-19-97; 8:45 am]
BILLING CODE 7590-01-P
