7806 Federal Register / Vol. 62, No. 34 / Thursday, February 20, 1997 / Notices

a toll-free telephone call to Western

Union at 1–(800) 248–5100 (in Missouri

1–(800) 342–6700). The Western Union

operator should be given Datagram

Identification Number N1023 and the

following message addressed to Robert

A. Capra: petitioner’s name and

telephone number, date petition was

mailed, plant name, and publication

date and page number of this Federal

Register notice. A copy of the petition

should also be sent to the Office of the

General Counsel, U.S. Nuclear

Regulatory Commission, Washington,

DC 20555–0001, and to Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603, attorney for the licensee.

Nontimely filings of petitions for

leave to intervene, amended petitions,

supplemental petitions and/or requests

for hearing will not be entertained

absent a determination by the

Commission, the presiding officer or the

presiding Atomic Safety and Licensing

Board that the petition and/or request

should be granted based upon a

balancing of the factors specified in 10

CFR 2.714(a)(1)(i)–(v) and 2.714(d).

For further details with respect to this

action, see the application for

amendments dated December 21, 1995,

as supplemented on October 24, 1996,

which are available for public

inspection at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC,

and at the local public document rooms

located at: for Byron, the Byron Public

Library District, 109 N. Franklin, P.O.

Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

Dated at Rockville, Maryland, this 13th day

of February 1997.

For the Nuclear Regulatory Commission.

George F. Dick, Jr.,

Project Manager, Project Directorate III–2,

Division of Reactor Projects—III/IV Office of

Nuclear Reactor Regulation.

[FR Doc. 97–4177 Filed 2–19–97; 8:45 am]

BILLING CODE 7590–01–P

[Docket No. 50–348]

Southern Nuclear Operating Company,

Inc. Alabama Power Company; Notice

of Withdrawal of Application for

Amendment to Facility Operating

License

The U.S. Nuclear Regulatory

Commission (the Commission) has

granted the request of Southern Nuclear

Operating Company, Inc. (Southern

Nuclear), to withdraw its August 23,

1996, application for proposed

amendment to Facility Operating

License No. NPF–2 for the Farley

Nuclear Plant, Unit 1, located in

Houston County, Alabama.

The proposed amendment would

have revised the technical specifications

by modifying the installation method for

previously licensed steam generator

tube elevated tubesheet laser welded

sleeves in Farley Nuclear Plant, Unit 1.

The Commission had previously

issued a Notice of Consideration of

Issuance of Amendments published in

the Federal Register on September 11,

1996 (61 FR 47982). However, by letter

dated February 7, 1997, Southern

Nuclear withdrew the proposed change.

For further details with respect to this

action, see the application for

amendment dated August 23, 1996, and

the licensee’s letter dated February 7,

1997, which withdrew the application

for license amendment. The above

documents are available for public

inspection at the Commission’s Public

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC,

and at the local public document room

located at the Houston-Love Memorial

Library, 212 W. Burdeshaw Street, P.O.

Box 1369, Dothan, Alabama.

Dated at Rockville, Md., this 13th day of

February 1997.

For the Nuclear Regulatory Commission.

Jacob I. Zimmerman,

Project Manager, Project Directorate II–2,

Division of Reactor Projects—I/II, Office of

Nuclear Reactor Regulation.

[FR Doc. 97–4176 Filed 2–19–97; 8:45 am]

BILLING CODE 7590–01–P

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.

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
Federal Register / Vol. 62, No. 34 / Thursday, February 20, 1997 / Notices 7807

consistent with their respective current

licensing basis.

Background

As a result of recent NRC inspection

activities, licensee notifications, and

licensee event reports, a safetysignificant 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, shortterm 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
7808 Federal Register / Vol. 62, No. 34 / Thursday, February 20, 1997 / Notices

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 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. 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 postaccident 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 noncondensable 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 designbasis 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,
Federal Register / Vol. 62, No. 34 / Thursday, February 20, 1997 / Notices 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 licensingbasis 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 designbasis 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-ofcoolant 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