[Federal Register Volume 62, Number 68 (Wednesday, April 9, 1997)]
[Notices]
[Pages 17252-17257]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-8915]


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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 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.
    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

[[Page 17253]]

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 stress-corrosion 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.
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    \1\ Specifically, the NRC staff observed that a loss-of-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.
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    The staff issued Information Notice (IN) 93-83, ``Potential Loss of 
Spent Fuel Pool Cooling After a Loss-of-Coolant 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.

[[Page 17254]]

    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\
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    \2\ Letter 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.
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    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 safety-related 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-of-Coolant 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.
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    \3\ On 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.
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    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.
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    \4\ Memorandum to the Commission, from J. Taylor, ``Resolution 
of Spent Fuel Storage Pool Action Plan Issues,'' dated July 26, 
1996.
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    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

[[Page 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 single-unit 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-of-Coolant 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

[[Page 17256]]

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.
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    \5\ Memorandum to the Commission, from J. Taylor, dated May 21, 
1996.
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    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 dual-unit 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-
of-coolant 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-to-boil 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  deg.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 single-unit 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 time-to-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 
safety-related emergency diesels (MCCs 1A21, 1A23, 1B21 and 1B23). The 
augmented train is seismically qualified. Portions of

[[Page 17257]]

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.
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    \6\ IEEE Std 308-1980.
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    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