[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