Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
October 7, 1997
NRC GENERIC LETTER 97-04: 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 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
(GL) to request that addressees submit information necessary to confirm the
adequacy of the net positive suction head (NPSH) available for emergency core
cooling (including core spray and decay heat removal) and containment heat
removal pumps.
Background
As a result of recent inspection activities, licensee notifications, and
licensee event reports (LER), the NRC has identified a safety-significant
issue that has generic implications and warrants action by the NRC to ensure
that the issue is 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 inadequacy may be a result of changes in plant
configuration, operating procedures, environmental conditions, or other
operating parameters over the life of the plant. In other cases, a plant's
NPSH analysis may not bound all postulated events for a sufficient time, or
assumptions used in the analysis may be non-conservative or inconsistent with
assumptions and methodologies traditionally considered acceptable by the
staff. For example, some licensees have recently discovered that they must
take new or additional credit for containment overpressure to meet the NPSH
requirements of the emergency core cooling system and containment heat removal
pumps. In the examples the NRC staff is familiar with, the need for crediting
this overpressure in NPSH analyses has arisen because of changes in plant
configuration and operating conditions, and/or errors in prior NPSH
calculations. As a result, the overpressure being credited by licensees may
be inconsistent with the plant's respective licensing basis.
9710060324. GL 97-04
October 7, 1997
Page 2 of 10
Current NPSH analyses (including any corresponding containment pressure
analyses) may not be available to the staff in docketed material (such as
final safety analysis reports) because some licensees have changed their
analyses. Consequently, this generic letter requests that addressees provide
current information regarding the NPSH analyses for emergency core cooling and
containment heat removal pumps. This generic letter applies only to ECCS and
containment heat removal pumps that meet the following criteria:
(1) pumps that take suction from the containment sump or
suppression pool following a design-basis loss-of-coolant
accident (LOCA) or secondary line break, or
(2) pumps used in "piggyback" operation that are necessary for
recirculation cooling of the reactor core and containment
(that is, pumps that are supplied by pumps which take suction
directly from the sump or suppression pool).
New NPSH analyses are neither requested nor required to be performed to
respond to this information request. However, new NPSH analyses may be
warranted if an addressee determines that changes in plant design or
procedures have occurred which may have reduced the available NPSH. In such
cases, each affected addressee must take appropriate corrective action to
restore its facility to compliance, in accordance with the requirements stated
in Appendix B to 10 CFR Part 50.
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 for which the NPSH
available for 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. As a result, that analysis erroneously
indicated that containment overpressure was not needed to satisfy NPSH
requirements for the pumps in the recirculation mode of operation. A
subsequent analysis showed that the licensee needed to take credit for 41.36
kPa (6 psig) of containment overpressure. In another analysis conducted in
1995 using increased service water temperature, the licensee found that
additional containment overpressure was necessary to meet NPSH requirements
for the same pumps. This additional overpressure constituted a significant
fraction of the peak calculated containment accident pressure. . GL 97-04
October 7, 1997
Page 3 of 10
On August 30, 1996, the licensee reported in LER 96-016 that calculations
recently performed to determine the NPSH available for the RHR pumps may have
been in error for the alternate, short-term recirculation flow path, because
of insufficient containment overpressure for a period of pump operation. The
licensee attributed this error to its failure to fully analyze the containment
pressure and sump temperature responses under design-basis accident
conditions.
Maine Yankee
In July and August 1996, an NRC Independent Safety Assessment Team (ISAT)
conducted an inspection to determine if Maine Yankee was operating in
conformance with its design and licensing bases. During that inspection, the
ISAT identified potential weaknesses in the NPSH analysis conducted by the
licensee for the containment spray pumps. 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 for a power level of
2700 thermal megawatts (MWt) and which does not include the hot-fluid
correction factor, indicates that 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. When the licensee repeated the analysis using the hot-fluid correction
factor (the use of which the ISAT viewed as a non-conservative assumption as
implemented by Maine Yankee), the available NPSH was only slightly greater
than the required NPSH for the same 5-minute period. For the remainder of the
transient, the licensee's analysis showed that NPSH available to the
containment spray pumps would exceed the amount required. As a basis for the
contention that the containment spray pumps were operable despite the 5-minute
period with available NPSH below the required NPSH, the licensee cited recent
pump tests showing 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 be available 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.
Pilgrim
As indicated in the NRC safety evaluation for licensing of the Pilgrim plant,
and in documents referenced by that evaluation, containment overpressure was
not necessary to satisfy RHR and core spray pump NPSH requirements at the time
of licensing. When the plant was modified in 1984, the licensee's safety
evaluation related to the modification stated that the available NPSH was
determined assuming (1) maximum debris loading conditions on the sump
strainers for the RHR and core spray pumps and (2) no credit for containment . GL 97-04
October 7, 1997
Page 4 of 10
overpressure. The licensee reaffirmed this assumption on April 14, 1994, in
its response to NRC Bulletin 93-02, "Debris Plugging of Emergency Core Cooling
Suction Strainers," dated March 23, 1993, stating 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 1995 in support of a
proposal to raise the design seawater injection temperature to 75øF, credit
was needed and taken for containment overpressure. 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. This
example illustrates that the potential exists that other licensees may have
made modifications to their plants that could be inconsistent with the plant's
licensing basis, and could reduce the NPSH available to the ECCS pumps.
Crystal River, Unit 3
In July 1996, an NRC inspection team conducted an Integrated Performance
Assessment of Crystal River, Unit 3. As part of that assessment, the team
reviewed the licensee's calculation which established the minimum post-LOCA
reactor building water level required to ensure that adequate NPSH would be
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 available water level and the
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, the licensee
failed to account for uncertainty in data regarding the required NPSH, as well
as for uncertainties associated with the hydraulic resistance of check valves
in the spray lines. In addition, the licensee used a hot fluid correction
factor to reduce the required NPSH without considering the effects of non-
condensable gases in the pumped fluid. Conservative assumptions included in
the licensee's 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 lack of
credit for containment overpressure.
The team concluded that the non-conservative assumptions used in the
licensee's NPSH calculation raise questions concerning the cavitation-free
operation of reactor building spray pump 1B during the recirculation phase of
operation. 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 used the
maximum expected reactor building water temperature.
Dresden
By letter dated January 13, 1997, the licensee for Dresden submitted a license
amendment request for approval of 13 kPa (2 psig) of containment overpressure
for the first 10 minutes . GL 97-04
October 7, 1997
Page 5 of 10
following a design-basis LOCA. This overpressure is necessary to compensate
for an NPSH deficiency for the low pressure coolant injection (LPCI) and core
spray pumps. The licensee identified the need for overpressure after
discovering that an incorrect value for the ECCS suction strainer head loss
had been used in the design-basis NPSH calculation. As part of a design-basis
review, the licensee determined that the actual head loss across the suction
strainers was 1.8 m (5.8 feet) for clean strainers, rather than the 0.30 m
(1 foot) head loss assumed in Dresden's original design basis as documented in
the final safety analysis report and vendor drawings.
Because the licensee could not determine with certainty if overpressure was
part of the original Dresden licensing basis, the licensee concluded that the
use of overpressure constituted an unreviewed safety question and therefore
requested staff approval to credit overpressure. In a license amendment dated
January 28, 1997, the staff approved the requested use of 13 kPa (2 psig) of
containment overpressure. In a subsequent license amendment issued on
April 30, 1997, the staff approved the use of a maximum of 65 kPa (9.5 psig)
of containment overpressure for NPSH, for the first 240 seconds following a
design-basis LOCA. The need for this greater amount of overpressure arose
primarily because of a higher calculated suppression pool temperature than
that used in the analysis to support 13 kPa (2 psig) of overpressure.
Monticello
In a report submitted to the NRC on April 15, 1997, pursuant to 10 CFR 50.72,
the licensee for Monticello reported that the NPSH available to its core spray
pumps may not meet the required NPSH under all accident conditions. The
licensee discovered this possibility during a review of ECCS pump NPSH
requirements, when a higher head loss than had previously been assumed for the
ECCS suction strainers was calculated. During discussions with the licensee,
the staff learned that the head loss across the suction strainers is
approximately 3.57 m (11.7 feet) per 38,000 liters/minute (10,000 gpm), rather
than the 0.3048 m (1 foot) per 38,000 liters/minute (10,000 gpm) assumed in
the original design-basis analysis.
The licensee determined that for a recirculation line break with a single
failure of the LPCI loop select logic, and with credit for containment
overpressure, the core spray pumps would have an NPSH deficit and the LPCI
pumps would have approximately 0.15 m (0.5 feet) of margin in NPSH. Following
discovery of the NPSH condition, the licensee conducted an operability
evaluation of the LPCI and core spray pumps, and made this evaluation
available to the staff for review. Subsequently, on May 9, 1997, the licensee
for Monticello commenced a voluntary shutdown of the plant because of the
possible NPSH deficit for the ECCS pumps that would occur as a result of
postulated clogging of the ECCS suction strainers under design-basis LOCA
conditions.
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 . GL 97-04
October 7, 1997
Page 6 of 10
licensees of possible scenarios for which the NPSH available for ECCS and
containment heat removal pumps is insufficient. 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," dated September 14, 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 to ensure that the systems can reliably perform their intended
functions under all design-basis LOCA conditions. Inadequate NPSH could
cause voiding in the pumped fluid, resulting in pump cavitation. While some
ECCS and containment heat removal pumps can operate for relatively short
periods of time while cavitating, prolonged operation of any pump under
cavitation conditions can cause pump damage with potential common-mode failure
of the pumps. Such common-mode failure would result in the inability of the
ECCS 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 to the
ECCS 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 available
NPSH or a change in the original design basis of the plant. In particular,
recent examples indicate that licensees have credited containment overpressure
to satisfy NPSH requirements in response to changing plant conditions and
errors discovered in earlier NPSH calculations.
RG 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 LOCAs. NRC Standard Review Plan (SRP) 6.2.2, "Containment Heat
Removal Systems" (NUREG-0800, Revision 4, dated October 1985) 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, in order 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
On the basis of the preceding discussion and examples, addressees are
requested to review, for each of their respective reactor facilities, the
current design-basis analyses used to . GL 97-04
October 7, 1997
Page 7 of 10
determine the available NPSH for the emergency core cooling (including core
spray and decay heat removal) and containment heat removal pumps that meet
either of the following criteria:
(1) pumps that take suction from the containment sump or
suppression pool following a design-basis LOCA or secondary
line break, or
(2) pumps used in "piggyback" operation that are necessary for
recirculation cooling of the reactor core and containment
(that is, pumps that are supplied by pumps which take suction
directly from the sump or suppression pool).
Based on this review, within 90 days from the date of this generic letter,
addressees are requested to provide the information outlined below for each of
their facilities. New NPSH analyses are neither requested nor required.
1. Specify the general methodology used to calculate the head loss associated
with the ECCS suction strainers.
2. Identify the required NPSH and the available NPSH.
3. Specify whether the current design-basis NPSH analysis differs from the
most recent analysis reviewed and approved by the NRC for which a safety
evaluation was issued.
4. Specify whether 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.
5. When containment overpressure is credited in the calculation of available
NPSH, confirm that an appropriate containment pressure analysis was done
to establish the minimum containment pressure.
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 submit
the information within the requested period, 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.
After reviewing responses to this generic letter, the NRC staff will notify
individual addressees if concerns are identified with regard to their
facilities. . GL 97-04
October 7, 1997
Page 8 of 10
Addressees should submit the required written response to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 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 requested information 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 conform with the current licensing basis for their respective
facilities, including the licensing safety analyses and the principal design
criteria which require and/or commit that safety-related components and
systems be provided to mitigate the consequences of design-basis accidents.
In particular, 10 CFR 50.46(a)(1)(i), which addresses the ECCS acceptance
criteria for light-water nuclear power reactors, requires in part that the
calculated cooling performance of the ECCS following a postulated LOCA
conforms to the criteria set forth in 10 CFR 50.46, including provisions for
peak cladding temperature and long-term cooling. The potential for loss of
adequate NPSH for ECCS pumps, and the cavitation that would result, raises the
concern that the ECCS would not be capable of maintaining the peak cladding
temperature below acceptable limits, and/or would not be capable of providing
core cooling over the duration of postulated accident conditions, as required
by 10 CFR 50.46.
Furthermore, the licensing bases of some plants credit the operation of
containment sprays for pressure control as well as for fission product
control. The potential for the loss of adequate NPSH for containment spray
pumps, and the cavitation that would result, raises the concern that
containment spray would not be capable of reducing and maintaining the
containment pressure and temperature below design values and would not be
capable of reducing the radiological dose consequences consistent with plants'
licensing bases.
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) 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.
Federal Register Notification
A notice of opportunity for public comment was published in the Federal
Register on February 20, 1997 (62 FR 7806) to solicit public comments on the
draft of this generic letter. A total of 17 comments were received from
interested parties, including one industry group, one legal group affiliated
with the nuclear power industry, and two licensees. When . GL 97-04
October 7, 1997
Page 9 of 10
redundant comments are considered, 12 distinct comments were identified by the
staff. Copies of the staff evaluation of these comments have been made
available in the NRC Public Document Room.
Paperwork Reduction Act Statement
This generic letter contains information collections that are subject to the
Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information
collections were approved by the Office of Management and Budget, approval
number 3150-0011, which expires on August 31, 2000.
The public reporting burden for this collection of information is estimated to
average 200 hours per response, including the time for reviewing instructions,
searching existing data sources, gathering and maintaining the data needed,
and completing and reviewing the collection of information. The NRC is
seeking public comment on the potential impact of the collection of
information contained in the generic letter and on the following issues:
1. Is the proposed collection of information necessary for the proper
performance of the functions of the NRC, including whether the information
will have practical utility?
2. Is the estimate of burden accurate?
3. Is there a way to enhance the quality, utility, and clarity of the
information to be collected?
4. How can the burden of the collection of information be minimized,
including the use of automated collection techniques?
Send comments on any aspect of this collection of information, including
suggestions for reducing this burden, to the Information and Records
Management Branch, T-6 F33, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001, and to the Desk Officer, Office of Information and Regulatory
Affairs, NEOB-10202 (3150-0011), Office of Management and Budget, Washington,
DC 20503.
The NRC may not conduct or sponsor, and a person is not required to respond
to, a collection of information unless it displays a currently valid OMB
control number.. GL 97-04
October 7, 1997
Page 10 of 10
If you have any questions about this matter, please contact one of the
technical contacts listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project manager.
signed by
Jack W. Roe, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: William O. Long, NRR
301-415-3026
E-mail: wol@nrc.gov
Richard M. Lobel, NRR
301-415-2865
E-mail: rml@nrc.gov
Lead Project Manager: T.J. Kim, NRR
301-415-1392
E-mail: tjk3@nrc.gov
Attachment: List of Recently Issued NRC Generic Letters
