Federal Register / Vol. 62, No. 31 / Friday, February 14, 1997 / Notices 7075

Type of Review: Re-clearance, without

change, of a currently approved

collection.

Title: Mergers of Federally Insured

Credit Unions.

Description: Part 708b of NCUA’s

regulations sets forth the procedures for

credit union mergers.

Respondents: Federal and State Credit

Unions.

Estimated No. of Respondents/

Recordkeepers: 200.

Estimated Burden Hours Per

Response: 15.

Frequency of Response: On occasion.

Estimated Total Annual Burden

Hours: 3,000.

Estimated Total Annual Cost:

$44,640.

OMB Number: 3133–0035.

Form Number: None.

Type of Review: Re-clearance, without

change, of a currently approved

collection.

Title: Trustees and Custodians of

Pension Plans.

Description: A federal credit union

acting as trustee for a retirement plan

must maintain individual records for

each participant and provide each

participant with notice of the insurance

status of their account.

Respondents: Federal Credit Unions.

Estimated No. of Respondents/

Recordkeepers: 3,877.

Estimated Burden Hours Per

Response: 1.

Frequency of Response: On occasion.

Estimated Total Annual Burden

Hours: 193,850.

Estimated Total Annual Cost:

$2,884,488.

By the National Credit Union

Administration Board on February 6, 1997.

Becky Baker,

Secretary of the Board.

[FR Doc. 97–3687 Filed 2–13–97; 8:45 am]

BILLING CODE 7535–01–P

NUCLEAR REGULATORY

COMMISSION

Proposed Generic Communication;

Loss of Reactor Coolant Inventory and

Associated Potential for Loss of

Emergency Mitigation Functions While

in a Shutdown Condition (M92635)

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 (1) assess the

susceptibility of their emergency core

cooling system (ECCS) to common-cause

failure as a result of reactor coolant

system (RCS) draindown while in a

shutdown condition, and (2) submit

certain information, pursuant to

§ 50.54(f) of Title 10 of the Code of

Federal Regulations (10 CFR 50.54(f)),

concerning their findings regarding

potential pathways for inadvertent RCS

drain-down and the suitability of

configuration control and operating

practices during reactor shutdown

cooling. This information will enable

NRC staff to verify whether addressees

comply and conform with NRC

regulatory and license requirements;

i.e., are adequately maintaining the

residual heat removal safety function to

transfer fission product decay heat and

other residual heat from the reactor core

(General Design Criterion (GDC) 34 of

Appendix A to 10 CFR 50), and the

ECCS to provide abundant emergency

core cooling when required (GDC 35 of

Appendix A to 10 CFR part 50). 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

17, 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.

ADDRESSES: 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, D.C.

FOR FURTHER INFORMATION CONTACT:

Muhammad M. Razzaque (301) 415–

2882.

SUPPLEMENTARY INFORMATION:

NRC Generic Letter 97–xx: Loss of

Reactor Cooland Inventory and

Associated Potential for Loss of

Emergency Mitigation Functions

While in a Shutdown Condition

Addressees

All holders of operating licenses for

pressurized-water reactors (PWRs),

except those that have certified to the

permanent cessation of operations.

Purpose

The U.S. Nuclear Regulatory

Commission (NRC) is issuing this

generic letter to request that addressees

(1) assess the susceptibility of their

emergency core cooling systems (ECCSs)

to common-cause failure as a result of

reactor coolant system (RCS) draindown

while in a shutdown condition, and (2)

submit certain information, pursuant to

§ 50.54(f) of Title 10 of the Code of

Federal Regulations (10 CFR 50.54(f)),

concerning their findings regarding

potential pathways for inadvertent RCS

drain-down and the suitability of

configuration control and operating

practices during reactor shutdown

cooling. This information will enable

NRC staff to verify whether addressees

comply and conform with NRC

regulatory and license requirements;

i.e., are adequately maintaining the RHR

safety function to transfer fission

product decay heat and other residual

heat from the reactor core (General

Design Criterion (GDC) 34 of Appendix

A to 10 CFR part 50), and the ECCS to

provide abundant emergency core

cooling when required (GDC 35 of

Appendix A to 10 CFR part 50).

Background

The NRC issued Information Notice

(IN) 95–03, ‘‘Loss of Reactor Coolant

Inventory and Potential Loss of

Emergency Mitigation Functions While

in a Shutdown Condition,’’ on January

12, 1995, to alert addressees to an

incident at the Wolf Creek Plant

involving the loss of reactor coolant

inventory while the reactor was in a

shutdown condition. In that event,

operators were attempting to reborate

residual heat removal (RHR) train B,

while at the same time maintenance

personnel were repacking an RHR train

A-to-train B crossover isolation valve.

Train B is reborated by recirculating

water through a loop that contains the

RHR system piping, the refueling water

storage tank (RWST), a containment

spray pump, a manual RWST isolation
7076 Federal Register / Vol. 62, No. 31 / Friday, February 14, 1997 / Notices

valve, and an RHR system crossover

line. When the RWST isolation valve

was opened for the reboration process

and the train A-to-train B crossover

isolation valve was opened for stroke

testing, a drain-down path was

inadvertently created from the reactor

coolant system (RCS) to the RWST. This

drain-down path included a suction

header common to all ECCS pumps.

Events of this nature are considered

particularly significant because they can

result in loss of emergency core cooling

capability and involve the potential for

containment bypass. On March 25,

1996, the staff issued a supplement to

IN 95–03 that further analyzed the

event. The NRC has also issued a

number of other communications

describing events at reactor facilities

involving inadvertent loss of reactor

coolant inventory while the reactor was

in a shutdown condition. The Office for

Analysis and Evaluation of Operational

Data (AEOD) published AEOD/E704,

‘‘Discharge of Primary Coolant Outside

of Containment at PWRs While on RHR

Cooling,’’ in March 1987, which

documented six events involving RCS

backflow into the RWST. In Generic

Letter 88–17, ‘‘Loss of Decay Heat

Removal (DHR) 10 CFR 50.54(f),’’ dated

October 17, 1988, the NRC requested

several actions to address loss-of-DHR

events that occurred while reactors were

in a shutdown condition. In IN 91–42,

‘‘Plant Outage Events Involving Poor

Coordination Between Operations and

Maintenance Personnel During Valve

Testing and Manipulations,’’ dated June

27, 1991, the NRC discussed inadvertent

loss-of-inventory events.

Discussion

At Wolf Creek, all ECCS pump

suction lines are tied into a common

suction header. When the draindown

event occurred at Wolf Creek, hot RCS

water was introduced into this common

suction header between the RWST and

the ECCS pumps. This hot water flashed

to steam, resulting in a steam/water

mixture in the header. In the event of an

ECCS actuation, this mixture would

have been introduced into the suction of

the ECCS pumps. If operators had not

been able to terminate the event, the hot

water in the RWST suction piping might

have led to steam binding, which could

have affected all pumps in both ECCS

trains. In addition, water flashing to

steam in the header and the RWST

could have caused serious mechanical

damage to the RHR piping and the

RWST as a result of water hammer.

Finally, steaming through the RWST

establishes a containment bypass path.

The licensee estimated (using actual

plant conditions) that for an

unmitigated event, the reactor vessel

water level could have drained to the

bottom of the hot leg within 5 minutes

and, as a consequence, RHR pump A

would have lost suction, cavitated, and

failed. Shortly thereafter, the common

ECCS suction header could have

reached a 90-percent steam/water ratio.

The licensee also estimated that

continued boil-off could have caused

the pressure vessel water level to drop

to the point of core uncovery in less

than 1 hour.

The AEOD report ‘‘Reactor Coolant

System Blowdown at Wolf Creek on

September 17, 1994,’’ (AEOD/S95–01),

dated March 1995, noted 19 events in

which RCS water was transferred to the

RWST. On the basis of this history and

the potential for containment bypass,

the staff has concluded that additional

information is required to confirm the

adequacy of existing ECCS configuration

control and operating practices

regarding residual heat removal.

Requested Actions

Addressees are requested to

determine whether their ECCSs are

susceptible to common-cause failure,

e.g., as a result of events similar to the

Wolf Creek RCS drain-down event of

September 17, 1994.

If ECCSs are found to be susceptible

to common-cause failure, addressees are

expected to take corrective action, as

appropriate, in accordance with the

requirements stated in Section XVI of

Appendix B to 10 CFR Part 50, to ensure

compliance with NRC regulatory and

license requirements.

Requested Information

Within 120 days of the date of this

generic letter, addressees are requested

to submit a written summary report that

includes a description of the evaluation

conducted and the conclusions reached

concerning the susceptibility of the RCS

to drain-down events with a potential

for consequential common-cause ECCS

failure, and the corrective actions that

were taken, or that are planned to be

taken, if any, in response to the above

requested actions. If the RCS is found to

be susceptible to drain-down events,

describe each potential drain-down flow

path (include piping sizes, identify flow

path valves and their normal positions,

and identify valve interlocks and

provisions for valve position indication

in the control room), describe potential

valve testing manipulations or uses, and

describe any administrative controls

that are intended to be used to control

valve manipulations to preclude RCS

drain-down events.

Required Response

Within 30 days of the date of this

generic letter, addressees are required to

submit a written response indicating (1)

whether or not the requested actions

will be taken, (2) whether or not the

requested information will be

submitted, and (3) whether or not the

requested information will be submitted

within the requested time period.

Addressees who choose not to complete

the requested actions, or choose not to

submit the requested information, or are

unable to satisfy the requested

completion date must describe in their

response any alternative course of

action that is proposed to be taken,

including the basis for establishing the

acceptability of the proposed alternative

course of action and the basis for

continued operability of affected

systems and components, as applicable.

Address the required written

responses 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 of the Atomic Energy Act

of 1954, as amended, and 10 CFR

50.54(f). In addition, submit a copy to

the appropriate regional administrator.

Backfit Discussion

The actions requested in this generic

letter, if required, would be backfits in

accordance with NRC procedures and

are necessary to ensure that addressees

are in compliance with existing NRC

rules and regulations. Specifically, 10

CFR 50.46 requires that the ECCS be

designed to provide adequate flow

capability to maintain the core

temperature at an acceptably low value

and to remove decay heat for the

extended period of time required by the

long-lived radioactivity remaining in the

core. The Wolf Creek event has

demonstrated that the adequacy of ECCS

configuration control and operating

practices regarding residual heat

removal can adversely impact ECCS

performance and could prevent the

ECCS from performing its safety

function following events at reactor

facilities involving inadvertent loss of

reactor coolant inventory while the

reactor is shut down. Therefore, this

generic letter is being issued as if the

requested actions were compliance

backfits under the terms of 10 CFR

50.109(a)(4)(i). A full backfit analysis

was not performed. However, in

accordance with NRC procedures, an

evaluation was prepared stating the

objectives of and the reasons for the

requested actions and the basis for

invoking the compliance exception if
Federal Register / Vol. 62, No. 31 / Friday, February 14, 1997 / Notices 7077

the requested actions were to be

required. A copy of this evaluation will

be made available in the NRC Public

Document Room.

Dated at Rockville, Maryland, this 10th 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. 97–3737 Filed 2–13–97; 8:45 am]

BILLING CODE 7590–01–P

Advisory Committee on Reactor

Safeguards Joint Meeting of the ACRS

Subcommittees on Materials and

Metallurgy and on Severe Accidents;

Meeting

The ACRS Subcommittees on

Materials and Metallurgy and on Severe

Accidents will hold a joint meeting on

March 4 and 5, 1997, Room T–2B3,

11545 Rockville Pike, Rockville,

Maryland.

The entire meeting will be open to

public attendance.

The agenda for the subject meeting

shall be as follows:

Tuesday, March 4, 1997—8:30 a.m.

until the conclusion of business

Wednesday, March 5, 1997—8:30 a.m.

until 12:00 Noon

The Subcommittees will review the

regulatory analysis and technical bases

for the steam generator tube integrity

rule, as well as an associated regulatory

guide. The purpose of this meeting is to

gather information, analyze relevant

issues and facts, and to formulate

proposed positions and actions, as

appropriate, for deliberation by the full

Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittees, their

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittees, along with

any of their consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittees will then hear

presentations by and hold discussions

with representatives of the NRC staff

and other interested persons regarding

this review.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements,

and the time allotted therefor can be

obtained by contacting the cognizant

ACRS staff engineer, Mr. Noel F. Dudley

(telephone 301/415–6888) between 7:30

a.m. and 4:15 p.m. (ET). Persons

planning to attend this meeting are

urged to contact the above named

individual one or two working days

prior to the meeting to be advised of any

potential changes to the agenda, etc.,

that may have occurred.

Dated: February 7, 1997.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 97–3738 Filed 2–13–97; 8:45 am]

BILLING CODE 7590–01–P

[DD–97–03]

Director’s Decision Under 10 CFR

2.206

In the Matter of Davis-Besse Nuclear Power

Station and Toledo Coalition for Safe

Energy.

Notice is hereby given that the

Director, Office of Nuclear Material

Safety and Safeguards, has taken action

with regard to the Petition of December

5, 1995, by the Toledo Coalition for Safe

Energy, Alice Hirt, Charlene Johnston,

Dini Schut, and William Hoops

(Petitioners), that the Director of the

Office of Nuclear Material Safety and

Safeguards exercise his authority to

immediately issue orders to prevent the

loading of spent nuclear fuel into the

VECTRA Technologies, Inc. (VECTRA)

NUHOMS dry-shielded canisters (DSCs)

at the Davis-Besse Nuclear Power

Station until an NRC rulemaking and/or

license modification hearing is

conducted on all safety-related changes

which have been made to the DSCs, as

described in the Safety Analysis Report.

Also, the NRC was requested not to

authorize any loading of the DSCs until

a written procedure for unloading, in

both urgent and non-urgent

circumstances, was written, approved,

and field-tested.

Petitioners contend that the safety of

the DSCs has been compromised

because of reduction in the thickness of

the DSC welds. In addition, they claim

that the NRC administrative process by

which permission was granted for

VECTRA to deliver the DSCs to the

Davis-Besse station and for the DSCs to

be used on site are legally suspect,

expressing the belief that agency

rulemaking or some other public

proceeding is necessary for permission

for such a transfer and use to be granted.

The Director of the Office of Nuclear

Material Safety and Safeguards has

determined that the NRC Certificate of

Compliance for VECTRA’s standardized

NUHOMS should be modified to require

a fabrication inspection of the DSC. An

agency rulemaking is, therefore, needed

and should be conducted to accomplish

this modification. However, because the

continued storage of spent fuel in the

DSCs at Davis-Besse does not pose an

unreasonable risk to public health and

safety, there is no technical basis to

require the DSCs to be unloaded

pending completion of this rulemaking.

Further, VECTRA has already been cited

for a nonconformance with NRC

regulations, and there is no basis in the

Petition to take other action in this

regard. Toledo Edison has developed

loading and unloading procedures for

handling spent fuels. These procedures

have been applied for the dry run

testing with NRC’s oversight. Therefore,

there is no basis in the Petition for

requiring halting of the ISFSI operation

at Davis-Besse. Accordingly, the Petition

from the Toledo Coalition for Safe

Energy is granted to the extent that it

requests an agency rulemaking and is

denied in all other respects. The reasons

for this decision are explained in the

‘‘Director’s Decision under 10 CFR

2.206’’ (DD–97–03), which is available

for public inspection in the

Commission’s Public Document Room,

Gelman Building, 2120 L Street, NW,

Washington, DC 20555, and in the Local

Public Document Room, William

Carlson Library, University of Toledo,

2001 West Bancroft Avenue, Toledo,

Ohio 43606.

A copy of this Decision will be filed

with the Office of the Secretary for the

Commission in accordance with 10 CFR

2.206(c). As provided by this regulation,

the Decision will constitute the final

action of the Commission 25 days after

the date of issuance of the Decision

unless the Commission, on its own

motion, institutes a review of the

Decision within that time.

Dated at Rockville, Maryland, this 5th day

of February 1997.

For the Nuclear Regulatory Commission.

Carl J. Paperiello,

Director, Office of Nuclear Material Safety

and Safeguards.

[FR Doc. 97–3739 Filed 2–13–97; 8:45 am]

BILLING CODE 7590–01–P