17252 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

Attorney for licensee: Michael I.

Miller, Esquire; Sidley and Austin, One

First National Plaza, Chicago, Illinois

60603.

Local Public Document Room

location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois

61021.

NRC Project Director: Robert A. Capra.

Northern States Power Company,

Docket Nos. 50–282 and 50–306, Prairie

Island Nuclear Generating Plant, Unit

Nos. 1 and 2, Goodhue County,

Minnesota

Date of application for amendments:

January 29, 1997, as supplemented

February 11, 12, March 7, 10, 11, 19,

and 20, 1997.

Brief description of amendments: The

amendments authorize Northern States

Power Company to continue operation

of Prairie Island Units 1 and 2 on an

interim basis, through the incorporation

of three license conditions into its

licenses, until a seismically qualified

emergency cooling water source is

provided that will provide the basis to

extend the time for operator postseismic cooling water load management.

This could be done either through a

seismic evaluation of the intake canal,

physical modifications to the intake

canal or plant, or some combination of

the two.

Date of issuance: March 25, 1997.

Effective date: March 25, 1997, with

implementation of License Condition 1

prior to Unit 2 entering Mode 2, with

implementation of the requirements of

License Condition 2 by July 1, 1997, and

December 31, 1998, and with

implementation of License Condition 3

at the next updated safety analysis

report update following completion of

License Condition 2, but no later than

June 1, 1999.

Amendment Nos.: 128 and 120.

Facility Operating License Nos. DPR–

42 and DPR–60: Amendments revised

the licenses.

Public comments requested as to

proposed no significant hazards

consideration (NSHC): Yes (62 FR 5857

dated February 7, 1997). This notice

provided an opportunity to submit

comments on the Commission’s

proposed NSHC determination. The

notice also provided for an opportunity

to request a hearing by March 10, 1997,

but indicated that if the Commission

makes a final NSHC determination, any

such hearing would take place after

issuance of the amendments. Because of

the significant revisions to the licensee’s

original application, NRC also

published a public notice of the

proposed amendments, issued a

proposed finding of no significant

hazards consideration, and requested

that any comments on the proposed

finding be provided to the staff by close

of business on March 20, 1997. The

notice was published in the St. Paul

Pioneer Press on March 15, 1997, the

Minneapolis Star Tribune on March 16,

1997, and the Red Wing Republican

Eagle on March 17, 1997. No comments

have been received. The Commission’s

related evaluation of the amendments,

finding of exigent circumstances, and

final determination of NSHC are

contained in a Safety Evaluation dated

March 25, 1997.

Attorney for licensee: Jay Silberg, Esq.,

Shaw, Pittman, Potts, and Trowbridge,

2300 N Street, NW, Washington, DC

20037.

Local Public Document Room

location: Minneapolis Public Library,

Technology and Science Department,

300 Nicollet Mall, Minneapolis,

Minnesota 55401

NRC Project Director: John N.

Hannon.

Dated at Rockville, Maryland, this 2nd day

of April, 1997.

For the Nuclear Regulatory Commission.

Jack W. Roe,

Director, Division of Reactor Projects III/IV,

Office of Nuclear Reactor Regulation.

[FR Doc. 97–8916 Filed 4–8–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–219, License No. DPR–16]

Oyster Creek Nuclear Generating

Station; Issuance of Final Director’s

Decision Under 10 CFR 2.206

Notice is hereby given that the

Director, Office of Nuclear Reactor

Regulation, U.S. Nuclear Regulatory

Commission (NRC), has granted in part

and denied in part Petitions, dated

September 19, 1994, and supplemented

by a letter dated December 13, 1994,

submitted by Messrs. Paul Gunter and

William deCamp, Jr. (Petitioners) on

behalf of Oyster Creek Nuclear Watch,

Reactor Watchdog Project, and Nuclear

Information and Resource Service.

Petitioners requested that the NRC take

immediate action with regard to Oyster

Creek Nuclear Generating Station

(OCNGS) operated by GPU Nuclear

Corporation (GPU or licensee). By letter

dated December 13, 1994, Petitioners

supplemented the Petition dated

September 19, 1994.

Specifically, the Petition of September

19, 1994, requested that the NRC (1)

immediately suspend the OCNGS

operating license until the licensee

inspects and repairs or replaces all

safety-class reactor internal component

parts subject to embrittlement and

cracking, (2) immediately suspend the

OCNGS operating license until the

licensee submits an analysis regarding

the synergistic effects of through-wall

cracking of multiple safety-class

components, (3) immediately suspend

the OCNGS operating license until the

licensee has analyzed and mitigated any

areas of noncompliance with regard to

irradiated fuel pool cooling as a singleunit boiling water reactor (BWR), and

(4) issue a generic letter requiring other

licensees of single-unit BWRs to submit

information regarding fuel pool boiling

in order to verify compliance with

regulatory requirements, and to

promptly take appropriate mitigative

action if the unit is not in compliance.

The supplemental Petition, in

addition to providing more information

on the original request, requested that

the NRC (1) suspend the OCNGS

operating license until the Petitioners’

concerns regarding cracking are

addressed, including inspection of all

reactor vessel internal components and

other safety-related systems susceptible

to intergranular stress-corrosion

cracking and completion of any and all

necessary repairs and modification; (2)

explain the discrepancies between the

response of the NRC staff dated October

27, 1994, to the Petition of September

19, 1994, and time-to-boil calculations

for the FitzPatrick plant; (3) require GPU

to produce documents for evaluation of

the time-to-boil calculation for the

OCNGS irradiated fuel pool; (4) identify

redundant components that may be

powered from onsite power supplies to

be used for spent fuel pool cooling as

qualified Class IE systems; (5) hold a

public meeting in Toms River, New

Jersey, to permit presentation of

additional information related to the

Petition; and (6) treat the Petitioners’

letter of December 13, 1994, as a formal

appeal of the denial of their request of

September 19, 1994, to immediately

suspend the OCNGS operating license.

The Director of the Office of Nuclear

Reactor Regulation has granted requests

(3), with the exception of suspending

OCNGS operating license which was

previously denied, and in part (4) of the

Petition of September 19, 1994, and

requests (2), (3), and (4) of the

supplemental Petition of December 13,

1994. The reasons for these decisions

are explained in the ‘‘Final Director’s

Decision Under 10 CFR 2.206: (DD–97–

08), the complete text of which follows

this notice. The decision and the

documents cited in the decision are

available for public inspection and

copying at the Commission’s Public
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17253

1Specifically, the NRC staff observed that a lossof-coolant accident followed by multiple failures of

emergency core cooling systems would be necessary

to achieve the adverse radiological conditions that

would preclude operator actions to ensure

continued adequate decay heat removal from the

spent fuel pool.

Document Room, the Gelman Building,

2120 L Street, NW., Washington, DC,

and at the local public document room

located at Ocean County Library,

Reference Department, 101 Washington

Street, Toms Rivers, NJ 08753.

A copy of this Final Director’s

Decision will be filed with the Secretary

of the Commission for review in

accordance with 10 CFR 2.206(c). As

provided in that regulation, the decision

will contribute the final action of the

Commission 25 days after the date of its

issuance, unless the Commission, on its

own motion, institutes a review of the

decision within that time.

For the Nuclear Regulatory Commission.

Dated at Rockville, Maryland, this 2nd day

of April 1997.

Attachment: DD 97–08

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

Final Director’s Decision Under 10 CFR

2.206

I. Introduction

By a Petition submitted pursuant to

10 CFR 2.206 on September 19, 1994

(Petition), Reactor Watchdog Project,

Nuclear Information and Resource

Service, and Oyster Creek Nuclear

Watch (Petitioners) requested that the

U.S. Nuclear Regulatory Commission

(NRC) take immediate action with

regard to Oyster Creek Nuclear

Generating Station (OCNGS) operated

by GPU Nuclear Corporation (GPU or

Licensee). By letter dated December 13,

1994, Petitioners supplemented the

Petition.

In the Petition of September 19, 1994,

Petitioners requested that the NRC: (1)

immediately suspend the OCNGS

operating license until the Licensee

inspects and repairs or replaces all

safety-class reactor internal component

parts subject to embrittlement and

cracking, (2) immediately suspend the

OCNGS operating license until the

Licensee submits an analysis regarding

the synergistic effects of through-wall

cracking of multiple safety-class

components, (3) immediately suspend

the OCNGS operating license until the

Licensee has analyzed and mitigated

any areas of noncompliance with regard

to irradiated fuel pool cooling as a

single-unit boiling water reactor (BWR),

and (4) issue a generic letter requiring

other licensees of single-unit BWRs to

submit information regarding fuel pool

boiling in order to verify compliance

with regulatory requirements and to

promptly take appropriate mitigative

action if the unit is not in compliance.

In addition to providing more

information on the original request, the

supplement dated December 13, 1994,

requested that the NRC: (1) suspend the

OCNGS operating license until

Petitioners’ concerns regarding cracking

are addressed, including inspection of

all reactor vessel internal components

and other safety-related systems

susceptible to intergranular stresscorrosion cracking and completion of

any and all necessary repairs and

modifications, (2) explain the

discrepancies between the response of

the NRC staff dated October 27, 1994, to

the Petition and time-to-boil

calculations for the FitzPatrick Plant, (3)

require GPU to produce documents for

evaluation of the time-to-boil

calculations for the OCNGS irradiated

fuel pool, (4) identify redundant

components that may be powered from

onsite power supplies to be used for

spent fuel pool cooling as qualified

Class 1E systems, (5) hold a public

meeting in Toms River, New Jersey, to

permit presentation of additional

information related to the Petition, and

(6) treat Petitioners’ letter of December

13, 1994, as a formal appeal of the

denial of their request of September 19,

1994, to immediately suspend the

OCNGS operating license.

On October 27, 1994, the Director of

the Office of Nuclear Reactor Regulation

informed the Petitioners that he was

denying their request for immediate

suspension of the OCNGS operating

license, that their Petition was being

evaluated under 10 CFR 2.206 of the

Commission’s regulations, and that

action would be taken in a reasonable

time. By letter dated April 10, 1995, the

Director denied requests (5) and (6) of

Petitioner’s supplemental Petition. On

August 4, 1995, the Director issued a

Partial Director’s Decision (DD–95–18),

denying requests (1) and (2) of their

Petition of September 19, 1994, and

request (1) of the supplemental Petition

of December 13, 1994. A decision

regarding requests (3) and (4) of the

Petition of September 19, 1994, and

requests (2), (3), and (4) of the

supplemental Petition of December 13,

1994, was deferred pending completion

of our review.

The NRC staff’s review of the Petition

and supplemental Petition is now

complete. For the reasons set forth

below, requests (3), with the exception

of suspending OCNGS operating license

which was previously denied, and (4) of

the Petition of September 19, 1994, are

granted in part and requests (2), (3), and

(4) of the supplemental Petition of

December 13, 1994 are granted as

described below.

II. Background

On November 27, 1992, a report was

filed pursuant to 10 CFR Part 21 by two

contract engineers that notified the

Commission of potential design

deficiencies in spent fuel pool decay

heat removal systems and containment

systems at Susquehanna Steam Electric

Station (SSES). The report noted that

under certain conditions, systems

designed to remove decay heat from the

spent fuel pool would be unable to

perform their intended function, and

that as a result of concurrent plant

conditions it would not be possible for

operators to place backup systems in

service or that backup systems would

otherwise be unable to perform their

intended function. The report

concluded that under such conditions,

the spent fuel pool could reach boiling

conditions and that the adverse

environment created by a boiling pool

would render systems designed to

remove decay heat from the reactor core

and systems designed to limit the

release of fission products to the

environment unable to perform their

intended function. The ultimate

consequence of these conditions could

be the failure (meltdown) of fuel in both

the reactor vessel and the spent fuel

pool and a substantial release of fission

products to the environment that would

cause significant harm to public health

and safety.

Although the issues raised by this Part

21 report appeared to be of low safety

significance, because of the low

probability that the necessary sequence

of events would take place,1 the

complex nature of the issues prompted

the NRC staff to undertake an extensive

evaluation of the matter. The NRC staff

review process, which continued from

November 1992 to June 1995, included

information-gathering trips to the

licensee’s engineering offices and to

SSES, public meetings with the

licensee, public meetings and written

correspondence with the authors of the

Part 21 report, and numerous written

requests for information to the licensee

and corresponding responses.

The staff issued Information Notice

(IN) 93–83, ‘‘Potential Loss of Spent

Fuel Pool Cooling After a Loss-ofCoolant Accident or a Loss of Offsite

Power,’’ on October 7, 1993, which

informed licensees of all operating

reactors of the nature of the issues

raised in the Part 21 report.
17254 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

2Letter to R. Byram, Pennsylvania Power & Light

Company, from J. Stolz, NRC, ‘‘Susquehanna Steam

Electric Station, Units 1 and 2, Safety Evaluation

Regarding Spent Fuel Pool Cooling Issues (TAC No.

M85337),’’ dated June 19, 1995.

3On January 25, 1994, the licensee for Dresden,

Unit 1, a permanently shutdown facility, discovered

approximately 55,000 gallons of water in the

basement of the unheated Unit 1 containment. The

water originated from a rupture of the service water

system that occurred as a result of freeze damage.

The licensee investigated further and found that

although the fuel transfer system was not damaged,

there was a potential for a portion of the fuel

transfer system inside containment to fail and result

in a partial draindown of the spent fuel pool that

contained 660 spent fuel assemblies. The NRC

issued NRC Bulletin 94–01, ‘‘Potential Fuel Pool

Draindown Caused by Inadequate Maintenance

Practices at Dresden Unit 1,’’ on April 8, 1994, to

all licensees with permanently shutdown reactors

that had spent fuel stored in spent fuel pools. The

NRC requested that such licensees take certain

actions to ensure that spent fuel storage safety did

not become degraded.

4Memorandum to the Commission, from J.

Taylor, ‘‘Resolution of Spent Fuel Storage Pool

Action Plan Issues,’’ dated July 26, 1996.

The NRC staff issued a draft safety

evaluation (SE) addressing the issues

raised in the Part 21 report on SSES for

comment on October 25, 1994. After

receiving comments from the licensee,

the authors of the Part 21 report, and the

Advisory Committee on Reactor

Safeguards, the staff issued a final SE

regarding the issues raised in the Part 21

report for the SSES on June 19, 1995

(SSES SE).2

The NRC staff reviewed and evaluated

the SSES plant design and inspected

operation of SSES plant equipment with

respect to the various event sequences

described in the Part 21 report. The staff

also evaluated the response of SSES

plant equipment to a broader range of

initiating events than was identified in

the Part 21 report. For example, the staff

considered the safety significance of a

loss of spent fuel pool decay heat

removal capability resulting from a loss

of offsite power events, from seismic

events, and from flooding events. The

staff considered the safety significance

of such events potentially leading to

spent fuel pool boiling sequences that

could, in turn, jeopardize safety-related

equipment needed to maintain reactor

core cooling. The NRC staff conducted

both deterministic and probabilistic

evaluations to fully understand the

safety significance of the issues raised.

The staff evaluated the safety

significance of the issues as they

pertained to the plant at the time the

Part 21 report was submitted and as

they pertained to the plant after the

completion of certain voluntary

modifications made at SSES during the

course of the NRC staff’s review.

Finally, the staff examined licensing

issues associated with the design of the

spent fuel pool cooling system to

determine the extent to which SSES’s

design and operation met the applicable

regulatory requirements.

On the basis of the staff’s

deterministic analysis of the plant as it

was configured at the time the SSES SE

was prepared, the NRC staff concluded

that systems used to cool the spent fuel

storage pool are adequate to prevent

unacceptable challenges to safetyrelated systems needed to protect the

health and safety of the public during

design-basis accidents.

On the basis of its probabilistic

evaluation, the NRC staff concluded that

the specific scenario involving a large

radionuclide release from the reactor

vessel, which was described in the Part

21 report, is a sequence of very low

probability. The staff’s evaluation

concluded that even with consideration

of the additional initiating events

previously described, ‘‘loss of spent fuel

pool cooling events’’ represented a

challenge of low safety significance to

the plant at the time the Part 21 report

was submitted. However, the staff also

concluded that the plant modifications

and procedural upgrades made during

the course of the staff’s review, which

included removing the gates that

separate the spent fuel storage pools

from the common cask storage pit,

installation of remote spent fuel pool

temperature and level indication in the

control room, and numerous procedural

upgrades, provided a measurable

improvement in plant safety and that

these conclusions had potential generic

implications. In summary, with regard

to loss of spent fuel pool cooling events,

the SSES SE concluded that the design

of the SSES facility was adequate to

protect public health and safety.

With regard to licensing-basis design

issues, the staff concluded that only a

loss of spent fuel pool cooling initiated

by a seismic event was considered in

the original granting of the SSES license

by the NRC.

The staff issued IN 93–83,

Supplement 1, ‘‘Potential Loss of Spent

Fuel Pool Cooling After a Loss-ofCoolant Accident or a Loss of Offsite

Power,’’ to all power reactor licensees

on August 24, 1995, describing the

conclusions of the June 19, 1995, SSES

SE. The information notice described

the staff’s plans to implement a generic

action plan to evaluate the generic

concerns raised in the SSES SE and to

address certain additional concerns

arising from a special inspection at a

permanently shutdown reactor facility.3

The generic action plan, entitled ‘‘Task

Action Plan for Spent Fuel Storage Pool

Safety’’ (Task Action Plan), was issued

on October 13, 1994, and included the

following actions: (1) A search for and

analysis of information regarding spent

fuel storage pool issues, (2) an

assessment of the operation and design

of spent fuel storage pools at selected

reactor facilities, (3) an evaluation of the

assessment findings for safety concerns,

and (4) selection and execution of an

appropriate course of action based on

the safety significance of the findings.

As part of the Task Action Plan

review, the staff reviewed operating

experience, as documented in licensee

event reports and other information

sources, as well as in previous studies

of spent fuel pool issues. The staff also

gathered detailed design data relating to

the design basis and functional

capability of the fuel storage pool, the

fuel pool cooling system, and other

systems associated with fuel storage for

every operating reactor and analyzed

these data to identify potential safety

issues regarding a loss of spent fuel pool

cooling or a loss of coolant inventory.

The NRC staff forwarded the results of

its Task Action Plan review to the

Commission on July 26, 1996.4 The staff

concluded that existing spent fuel

storage pool structures, systems, and

components provided adequate

protection of public health and safety at

all operating reactors. Protection is

provided by several layers of defenses

that perform accident prevention

functions (e.g., quality controls on

design, construction, and operation),

accident mitigation functions (e.g.,

multiple cooling systems and multiple

makeup water paths), radiation

protection functions, and emergency

preparedness functions. Design features

addressing each of these areas for spent

fuel storage for each operating reactor

have been reviewed and approved by

the staff. In addition, the risk analyses

available for spent fuel storage suggest

that current design features and

operational constraints cause issues

related to spent fuel pool storage to be

a small fraction of the overall risk

associated with an operating light-water

reactor.

Notwithstanding these findings, the

NRC staff reviewed the design of every

operating reactor’s spent fuel pool to

identify strengths and weaknesses and

potential areas for safety enhancements.

The NRC staff identified seven

categories of design features that reduce

the reliability of spent fuel pool decay

heat removal, increase the potential for

loss of spent fuel coolant inventory, or

increase the potential for consequential

loss of essential safety functions at an

operating reactor. The NRC staff

determined that these design features

existed at 22 sites; OCNGS was not one
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17255

of the 22 sites. As the staff has

concluded that present facility designs

provided adequate protection of public

health and safety, possible safety

enhancements will be evaluated

pursuant to 10 CFR 50.109(a)(3). The

analyses for possible safety

enhancement backfits will consider

whether modifications to the plant

design to address the plant-specific

design features identified by the NRC

staff could provide a substantial

increase in the overall protection of

public health and safety and whether

such modifications could be justified on

a cost-benefit basis.

The NRC staff also identified three

additional categories of design features

that may have the potential to reduce

the reliability of spent fuel pool decay

heat removal, increase the potential for

loss of spent fuel coolant inventory, or

increase the potential for consequential

loss of essential safety functions at an

operating reactor. The NRC staff

preliminarily determined that these

design features existed at 11 sites.

OCNGS was not one of the 11 sites. The

staff has insufficient information at this

time to determine whether backfits

pursuant to 10 CFR 50.109(a)(3) are

warranted at the 11 sites. For plants

identified as having design features in

these three categories, the NRC staff will

gather and evaluate additional

information prior to determining

whether to require any backfits.

In addition to the plant-specific

analyses described above for 22 sites

which will address certain design

features, the NRC staff informed the

Commission in the July 26, 1996, Task

Action Plan report that it plans to

address issues related to the functional

performance of spent fuel pool decay

heat removal, as well as the operational

aspects related to coolant inventory

control and reactivity control, in a new

proposed performance-based rule for

shutdown operations (10 CFR 50.67) at

all operating reactors. The new rule is

schedule to be issued for public

comment in 1997.

The NRC staff sent the Task Action

Plan report of July 26, 1996, to all

operating power reactor licensees. For

those licensees whose plants have one

or more of the design features that

warrant a plant-specific safety

enhancement backfit analysis, the staff

has provided an opportunity to

comment on: (1) The accuracy of the

NRC staff’s understanding of the plant

design, (2) the safety significance of the

design concern, (3) the cost of potential

modifications to address the design

concern, and (4) the existing protection

from the design concern provided by

administrative controls or other means.

In developing a schedule and plans for

conducting all of the plant-specific

regulatory analyses, the NRC staff will

consider comments received from

licensees.

III. Discussion

A. Issuance of Generic Letter,

Compliance Verification, and Mitigative

Action (September 19, 1994 Petition

Items (3) and (4))

The Petitioners requested (Items (3)

and (4) of the September 19, 1994,

Petition) that the NRC immediately

suspend the OCNGS operating license

until GPU analyzes and mitigates any

areas of noncompliance with regard to

irradiated fuel pool cooling as a singleunit boiling water reactor, and that the

NRC issue a generic letter requiring

other licensees of single unit BWRs to

submit information regarding fuel pool

boiling in order to verify compliance

with NRC requirements and to take

quick mitigative action if the unit is not

in compliance.

As stated in the cover letter, the

October 27, 1994, Director’s letter

informed you that he denied your

request for immediate suspension of the

OCNGS operating license.

While the NRC has not issued and

does not plan to issue a generic letter,

the staff has communicated the

importance of conducting relevant spent

fuel pool decay heat removal activities

in accordance with technical

specifications and other plant-specific

applicable regulatory requirements to

licensees through the issuance of other

generic communications, as described

below. The staff also surveyed all

operating reactor licensees, including

GPU Nuclear Corporation, licensee for

OCNGS, to collect information on,

among other things, parameters affecting

boiling of the spent fuel pool. Results of

the survey relevant to this Petition are

discussed below.

The NRC staff issued three

information notices on matters related

to adequate removal of decay heat from

the spent fuel pool. IN 93–83, ‘‘Potential

Loss of Spent Fuel Pool Cooling After a

Loss-of-Coolant Accident or a Loss of

Offsite Power,’’ was issued on October

7, 1993, and described the concerns in

the November 27, 1992, SSES Part 21

report discussed above. IN 93–83,

Supplement 1, ‘‘Potential Loss of Spent

Fuel Pool Cooling After a Loss-ofCoolant Accident or a Loss of Offsite

Power,’’ issued on August 8, 1995,

informed licensees of the results of the

NRC’s review of the concerns at SSES.

IN 95–54, ‘‘Decay Heat Management

Practices During Refueling Outages,’’

was issued on December 1, 1995, and

described recent NRC assessments of

events at certain plants regarding the

licensee’s control of refueling operations

and the methods for removing decay

heat produced by the irradiated fuel

stored in the spent fuel pool during

refueling outages. IN 95–54

communicated to licensees that the

plant-specific events described therein

and in the previous information notices

illustrated the importance of ensuring

that (1) planned core offload evolutions,

including refueling practices and

irradiated fuel decay heat removal, are

consistent with the licensing basis,

including the final safety analysis

report, technical specifications, and

license conditions; (2) changes to these

evolutions are evaluated through the

application of the provisions of 10 CFR

50.59, as appropriate; and (3) all

relevant procedures associated with

core offloads have been appropriately

reviewed.

The staff surveyed operating reactors,

including Oyster Creek, as part of the (a)

Spent Fuel Pool (SFP) Task Action Plan,

and (b) follow-up actions related to

issues identified at Millstone, and

reviewed the degree to which fuel pool

operations compared with each facility’s

design basis and the degree that the fuel

pool design features conformed with

accepted guidance and standards. In the

case of Oyster Creek, the NRC staff

found no deviations in operation or

design as a result of either review. The

staff issued its report on the results of

spent fuel pool survey regarding

Millstone follow-up issues on May 21,

1996. As described in Section II of this

decision, the NRC staff forwarded its

report on the resolution of the SFP Task

Action Plan on July 26, 1996, to all

operating power reactor licensees.

As part of the SFP Task Action Plan,

the staff considered, on a generic basis,

the history of regulatory requirements

related to spent fuel pools as they were

applied in plant licensing actions. The

staff found that SFP-related regulatory

requirements have been evolving since

the first nuclear power plants were

licensed and that specific regulatory

guidance on the design of spent fuel

pool cooling systems was not formalized

until 1975, when the Standard Review

Plan was issued, which was after the

issuance of construction permits for

most currently operating reactors.

Because the regulatory requirements

were evolving during the era in which

the staff was conducting licensing

reviews for the current generations of

operating reactors, staff-approved

designs varied from plant to plant.

However, based on the recent survey

results, the staff concluded that all

operating reactors had design features
17256 Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices

5Memorandum to the Commission, from J.

Taylor, dated May 21, 1996.

for spent fuel storage (e.g., addressing

accident prevention functions, accident

mitigation functions, radiation

protection functions, and emergency

preparedness functions), which had

been reviewed and approved in the past

by the NRC. In addition, based on the

review of the survey results, the staff

found that all licensees were in

compliance with current NRC

requirements.

Although the NRC staff concluded

that all plants, including OCNGS, are in

compliance with the NRC spent fuel

pool design requirements, the staff

reviewed certain operating practices at

all operating reactor plants to verify that

the plants were being operated

consistent with the plant design as

described in the licensing basis,5

specifically with respect to refueling

outage practices associated with

offloading irradiated fuel into the spent

fuel pool. The staff concluded, on the

basis of the information collected and

reviewed and the specific licensee

actions taken and commitments made

during the course of this review, that

core offload practices are consistent

with the spent fuel pool decay heat

removal licensing basis for all plants, or

will be before the next refueling outage.

It should be noted, however, that during

the course of its review, the staff

determined that nine sites (involving

fifteen units) needed to modify their

licensing basis or plant practices,

pursuant to 10 CFR 50.59 or 10 CFR

50.90, to ensure that their refueling

practices adhered to their licensing

basis. This is an indication that these

plants may have previously performed

full core offloads inconsistent with their

licensing basis. The staff is reviewing

potential enforcement action for these

facilities. It should be noted that

OCNGS is not one of the nine sites.

The Petitioners requested that the

NRC immediately suspend the OCNGS

operating license until GPU analyzes

and mitigates any areas of

noncompliance with regard to irradiated

fuel pool cooling as a single-unit BWR,

and that the NRC issue a generic letter

requiring other licensees of single unit

BWRs to submit information regarding

fuel pool boiling in order to verify

compliance with NRC requirements and

take quick mitigative action if the unit

is not in compliance. These requests are

granted in part as described above.

Petitioners’ request for immediate

suspension of OCNGS operating license

was previously denied.

B. Time-to-Boil Calculations (December

13, 1994, Supplemental Petition Items

(2) and (3))

Petitioners’ supplementary request of

December 13, 1994, asked the NRC to

explain ‘‘discrepancies’’ between the

response of the NRC staff dated October

27, 1994, to the Petition and the

documented time-to-boil calculations

for the FitzPatrick Plant as they bear on

time-to-boil calculations for other

single-unit General Electric BWRs,

including OCNGS. Petitioners contend

that documents available in the Public

Document Room for FitzPatrick Plant, a

single-unit site, indicated a time-to-boil

following a loss-of-coolant accident of 8

hours, considerably less than the 25

hours SSES, a dual-unit site, committed

to in a letter dated June 1, 1994.

Petitioners also requested that the

Licensee, GPUN, produce time-to-boil

calculations for OCNGS.

The NRC staff letter of October 27,

1994, to Petitioners concluded that

time-to-boil conditions at single-unit

BWR sites, such as OCNGS, are of low

safety significance because, unlike dualunit sites, such as SSES, a large decay

heat rate associated with a short time to

reach boiling conditions is an

unrealistic assumption during periods

when the unit is operating and fuel in

the reactor vessel is subject to a loss-ofcoolant accident.

As explained in the Director’s letter to

Petitioners dated April 10, 1995, the

time-to-boil calculation results for the

FitzPatrick Plant single-unit BWR,

which were presented in a New York

Power Authority document dated May

31, 1990, were based on the maximum

postulated decay heat rates during a

refueling outage fuel discharge and full

core offload that occurred about 7 and

10 days, respectively, after reactor

shutdown. These calculations also

assumed that spent fuel pool cooling

was lost when the pool was at its

maximum calculated temperature. In

contrast, the staff calculated the time-toboil for FitzPatrick to be 25 hours for a

one-third core discharge 30 days after

reactor shutdown, assuming the spent

fuel pool was at its maximum

temperature limit for normal operation,

which is 125 °F. The details of this

calculation were provided in our

Director’s letter to you dated April 10,

1995. Additionally, the staff had

surveyed the factors that would most

significantly affect the time-to-boil (i.e.,

spent fuel pool volumes, rated reactor

thermal power level, total number of

fuel assemblies in the reactor vessel,

and spent fuel pool temperature limits)

for 12 General Electric Company BWR/

3 and BWR/4 reactors. The staff

concluded that its time-to-boil

calculations for FitzPatrick are

representative for United States singleunit BWRs as a whole, and OCNGS in

particular.

As part of the NRC staff’s Task Action

Plan activities, the staff collected

information from licensee documents to

calculate the time-to-boil for all

operating reactors on a consistent basis.

While the staff did not specifically

require licensees (including GPU) to

provide documentation to support timeto-boil calculations, the staff did

independently calculate the time-to-boil

for each plant from licensee-supplied

information in Final Safety Analysis

Reports and other design documents.

On this basis, the staff determined that

the time-to-boil at Oyster Creek is

average among single-unit BWRs, thus

confirming the same conclusion reached

earlier in the Director’s letter of April

10, 1995.

Accordingly, the Petitioners’ requests

to explain the ‘‘discrepancies’’ between

the response of the NRC staff dated

October 27, 1994, to the Petition and the

documented time-to-boil calculations

for the FitzPatrick Plant as they bear on

time-to-boil calculations for other

single-unit General Electric BWRs,

including OCNGS, and that GPU

produce documents for evaluation of

time-to-boil calculations are granted as

described above.

C. Redundant Class 1E Components and

Power Supplies (December 13, 1994,

Supplemental Petition Item (4))

In the supplemental Petition

submittal of December 13, 1994, the

Petitioners requested that the NRC

identify redundant components that

may be powered from on-site power

supplies to be used for spent fuel pool

cooling as qualified Class 1E systems at

Oyster Creek.

The Petitioners noted that while

Oyster Creek may have redundant

components, in their view it is

meaningless to have redundant

components and power supplies if they

have not been qualified to operate under

emergency conditions.

At Oyster Creek, spent fuel decay heat

removal consists of a two-train spent

fuel pool cooling system. The first train

(‘‘Spent Fuel Pool Cooling System’’) has

two pumps and two heat exchangers.

The second or augmented train,

installed in parallel with the first train,

contains two full capacity pumps and a

single heat exchanger. The four pumps

in both trains are powered from

electrical busses supported by safetyrelated emergency diesels (MCCs 1A21,

1A23, 1B21 and 1B23). The augmented

train is seismically qualified. Portions of
Federal Register / Vol. 62, No. 68 / Wednesday, April 9, 1997 / Notices 17257

6 IEEE Std 308–1980.

the spent fuel pool cooling system,

initially designed to be a non-seismic

system, has been upgraded to Seismic

Category I requirements. Those portions

of the system that do not meet seismic

requirements can be isolated from the

spent fuel pool cooling system if a

seismic event renders them inoperable.

It should be made clear that the NRC

staff does not require Class 1E

qualification for spent fuel pool cooling

equipment and instrumentation. Class

1E is the safety classification of electric

equipment and systems that are

essential to emergency reactor

shutdown, containment isolation,

reactor core cooling, and containment

and reactor heat removal, or are

otherwise essential in preventing

significant release of radioactive

material to the environment.6 The spent

fuel pool cooling system and monitoring

instrumentation are not required for

such functions.

In his letter of April 10, 1995, the

Director informed Petitioners that they

have not presented, nor was the staff

aware of, any evidence that the spent

fuel pool cooling system fails to comply

with its design basis, or that the licensee

failed to qualify these components to

the degree Petitioners describe such that

it would alter his decision as it pertains

to the safety significance of these issues.

Therefore, further review of the

qualification of spent fuel cooling

system components at OCNGS is not

warranted. Additionally, Petitioners

were informed that the staff would

continue its generic review of spent fuel

storage pool safety and would take

appropriate action based on the

conclusions of that review. Based on the

results of the generic review of spent

fuel storage pool safety thus far, the staff

has concluded that no additional

actions are warranted for the spent fuel

pool cooling system components at

OCNGS.

The Petitioners’ request to identify

redundant qualified Class 1E systems

was granted as described above.

IV. Conclusion

Although the staff has not initiated

formal enforcement proceedings in

response to the Petition, the staff has

taken a number of actions that address

the concerns raised in the Petition. For

example, during the course of its

review, the NRC staff has issued generic

communications responsive to

Petitioners’ request (4) of September 19,

1994. In addition, the NRC staff

reviewed the compliance of NRC

licensed facilities in the area of spent

fuel pool design responsive to

Petitioners’ request (3) of September 19,

1994. To this extent, the Petition is

granted in part. Finally, Petitioners’

supplemental petition requests (2), (3),

and (4) are granted as explained above.

A copy of this Final Director’s

Decision will be filled with the

Secretary of the Commission for review

in accordance with 10 CFR 2.206(c).

This Decision will become the final

action of the Commission 25 days after

its issuance unless the Commission, on

its own motion, institutes review of the

Decision within that time.

For the Nuclear Regulatory Commission.

Dated at Rockville, Maryland, this 2nd day

of April 1997.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 97–8915 Filed 4–8–97; 8:45 am]

BILLING CODE 7590–01–M

OFFICE OF PERSONNEL

MANAGEMENT

Federal Prevailing Rate Advisory

Committee; Open Committee Meeting

According to the provisions of section

10 of the Federal Advisory Committee

Act (Pub. L. 92–463), notice is hereby

given that meetings of the Federal

Prevailing Rate Advisory Committee

will be held on—

Thursday, April 10, 1997

Thursday, April 24, 1997

Thursday, May 15, 1997

Thursday, May 22, 1997

The meeting will start at 10 a.m. and

will be held in Room 5A06A, Office of

Personnel Management Building, 1900 E

Street, N.W., Washington, D.C.

The Federal Prevailing Rate Advisory

Committee is composed of a Chair, five

representatives from labor unions

holding exclusive bargaining rights for

Federal blue-collar employees, and five

representatives from Federal agencies.

Entitlement to membership on the

Committee is provided for in 5 U.S.C.

5347.

The Committee’s primary

responsibility is to review the Prevailing

Rate System and other matters pertinent

to establishing prevailing rates under

subchapter IV, chapter 53, 5 U.S.C., as

amended, and from time to time advise

the Office of Personnel Management.

These scheduled meetings will start

in open session with both labor and

management representatives attending.

During the meeting either the labor

members or the management members

may caucus separately with the Chair to

devise strategy and formulate positions.

Premature disclosure of the matters

discussed in these caucuses would

unacceptably impair the ability of the

Committee to reach a consensus on the

matters being considered and would

disrupt substantially the disposition of

its business. Therefore, these caucuses

will be closed to the public because of

a determination made by the Director of

the Office of Personnel Management

under the provisions of section 10(d) of

the Federal Advisory Committee Act

(Pub. L. 92–463) and 5 U.S.C.

552b(c)(9)(B). These caucuses may,

depending on the issues involved,

constitute a substantial portion of the

meeting.

The meeting on April 10 may contain

discussion of confidential private sector

survey data for the Newburgh, New

York, appropriated fund wage area. If

so, that portion of the meeting will be

closed to the public under the

provisions of section 10(d) of the

Federal Advisory Committee Act (Pub.

L. 92–463) and 5 U.S.C. 552b(c)(4).

Annually, the Chair compiles a report

of pay issues discussed and concluded

recommendations. These reports are

available to the public, upon written

request to the Committee’s Secretary.

The public is invited to submit

material in writing to the Chair on

Federal Wage System pay matters felt to

be deserving of the Committee’s

attention. Additional information on

these meetings may be obtained by

contacting the Committee’s Secretary,

Office of Personnel Management,

Federal Prevailing Rate Advisory

Committee, Room 5559, 1900 E Street,

NW., Washington, DC 20415 (202) 606–

1500.

Dated: April 3, 1997.

Phyllis G. Foley,

Chair, Federal Prevailing Rate Advisory

Committee.

[FR Doc. 97–9050 Filed 4–8–97; 8:45 am]

BILLING CODE 6325–01–M

SECURITIES AND EXCHANGE

COMMISSION

[Release No. 34–38471; File No. SR–DCC–

96–12]

Self-Regulatory Organizations; Delta

Clearing Corp.; Order Approving on a

Temporary Basis a Proposed Rule

Change Relating to Monitoring and

Limiting Exposure from Repurchase

Agreements

April 2, 1997.

On November 26, 1996, the Delta

Clearing Corp. (‘‘DCC’’) filed with the

Securities and Exchange Commission

(‘‘Commission’’) a proposed rule change

(File No. SR–DCC–96–12) pursuant to