[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
