[Federal Register Volume 60, Number 154 (Thursday, August 10, 1995)]
[Notices]
[Pages 40867-40871]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-19766]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-219]
GPU Nuclear Corporation, Oyster Creek Nuclear Generating Station;
Issuance of Partial Director's Decision Under 10 CFR Sec. 2.206
Notice is hereby given that the Director, Office of Nuclear Reactor
Regulation, U.S. Nuclear Regulatory Commission (NRC) has denied in part
a Petition, dated September 19, 1994, and supplemented December 13,
1994, submitted by Oyster Creek Nuclear Watch, Reactor Watchdog
Project, and Nuclear Information and Resource Service (Petitioners).
The Petition requested that the NRC take action regarding the Oyster
Creek Nuclear Generating Station (OCNGS) pursuant to 10 C.F.R.
Sec. 2.206.
The September 19, 1994, Petition requests 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 units are not in compliance.
The December 13, 1994, supplemental Petition requests that the NRC:
(1) suspend the license of the OCNGS 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 (IGSCC) and completion of
any and all necessary repairs and modifications; (2) explain
discrepancies between the response of the NRC staff dated October 27,
1994, to the Petition of September 19, 1994, and the time-to-boil
calculations for the FitzPatrick plant; (3) require the GPU Nuclear
Corporation 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 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 the
Petitioners' letter of December 13, 1994, as a formal appeal of the
denial of the Petitioners' request of September 19, 1994, to
immediately suspend the OCNGS operating license.
The Director of the Office of Nuclear Reactor Regulation has denied
Requests (1) and (2) of the September 19, 1994, Petition and Request
(1) of the December 13, 1994, supplemental Petition to suspend the
operating license of the OCNGS until the Licensee inspects and repairs,
modified, or replaces all safety-class reactor internal component parts
subject to embrittlement and intergranular stress corrosion cracking.
The reasons for this denial are explained in the ``Partial Director's
Decision Under 10 CFR Sec. 2.206'' (DD-95-18), the complete text of
which follows this notice, and which is available for public inspection
at the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local
[[Page 40868]]
public document room for the Oyster Creek Nuclear Generating Station
located at the Ocean County Library, Reference Department, 101
Washington Street, Toms River, NJ 08753. A decision regarding Requests
(3), and (4) of the September 19, 1994 Petition, and Requests (2), (3),
and (4), of the December 13, 1994, supplemental Petition will be issued
under separate cover upon completion of the NRC staff's review.
A copy of this Partial 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 constitute
the final action of the Commission 25 days after the date of the
issuance of the Decision, unless the Commission, on its own motion,
institutes a review of the Decision within that time.
Dated at Rockville, Maryland this 4th day of August 1995.
For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.
Appendix A--Partial Director's Decision Under 10 CFR Sec. 2.206 (DD95-
18)
I. Introduction
By letter dated September 19, 1994, Reactor Watchdog Project,
Nuclear Information and Resource Service (NIRS), and Oyster Creek
Nuclear Watch (Petitioners), submitted a Petition pursuant to Section
2.206 of Title 10 of the Code of Federal Regulations (10 C.F.R.
Sec. 2.206), requesting that the U.S. Nuclear Regulatory Commission
(NRC) take action with regard to the Oyster Creek Nuclear Generating
Station (OCNGS), operated by the GPU Nuclear Corporation (GPUN or the
Licensee). By letter dated December 13, 1994, Petitioners supplemented
the Petition.
The September 19, 1994, Petition requests 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 December 13, 1994, supplemental Petition requests that the NRC:
(1) suspend the license of the OCNGS 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 (IGSCC) and completion of
any and all necessary repairs and modifications; (2) explain
discrepancies between the response of the NRC staff dated October 27,
1994, to the Petition of September 19, 1994, and the time-to-boil
calculations for the FitzPatrick plant; (3) require GPUN 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 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 the Petitioners'
letter of December 13, 1994, as a formal appeal of the denial of the
Petitioners' request of September 19, 1994, to immediately suspend the
OCNGS operating license.
The September 19, 1994, Petition sought relief concerning safety-
class reactor internal components based on the following premises: (a)
the core shroud in General Electric BWRs is vulnerable to age-related
deterioration; (b) 12 domestic and foreign BWR owners have found
extensive cracking on welds of the core shroud; (c) only 10 of 36 U.S.
BWR owners have inspected their core shrouds and 9 of the 10 core
shrouds had cracks; (d) 19 of 25 selected BWR internal components are
susceptible to stress corrosion cracking and 6 of 19 are susceptible to
irradiation-assisted stress corrosion cracking; (e) as the oldest
operating General Electric Mark I BWR and the third oldest operating
reactor in the United States, OCNGS has been subjected to the longest
period of operational conditions that cause embrittlement and cracking;
(f) the BWR Owners Group (BWROG) stated that cracking of the core
shroud is a warning signal that additional safety-class reactor
internals are increasingly susceptible to age-related deterioration;
(g) cracking of any single part or multiple components jeopardizes safe
operation of that nuclear station; (h) Oyster Creek did not inspect for
core shroud cracking prior to the current refueling outage and other
safety-class reactor internals have not been adequately inspected for
cracking; and (i) a safety analysis has not been performed on the
potential synergistic effects of multiple-component cracking.
The September 19, 1994, Petition also sought relief concerning fuel
pool cooling design deficiencies, based on the following premises: (a)
various design defects in BWR fuel pool cooling systems pose a
significant increase in risk to the public safety and violate 10 CFR
50.59; 10 CFR Part 50, Appendix A, Criterion 63; 10 CFR Part 50,
Appendix B, Criterion III; and Regulatory Guides 1.13, 1.89, and 1.97;
(b) OCNGS is a single-unit facility with no adjacent units to rely upon
in the event that a design-basis event were to disable the fuel pool
cooling system; and (c) OCNGS has not docketed any material with regard
to BWR design deficiencies identified in the 10 CFR Part 21 Report of
Substantial Safety Hazard (November 27, 1992) of Messrs. Lochbaum and
Prevatte, and thus OCNGS may be in violation of NRC regulatory
requirements.
The Petitioners assert the following bases to support their
requests in the December 13, 1994, supplemental Petition: (a) the
October 27, 1994, letter of the NRC staff, acknowledging receipt of the
Petition and denying the requests for immediate suspension of the
operating license, failed to address concerns central to the Petition,
such as the Licensee's failure to recognize that IGSCC indicates that
cracking could be occurring in additional safety-class reactor internal
components and the Licensee's failure to perform inspections of all
safety-class components to determine whether cracking is occurring; (b)
recently discovered cracking in the top guide and core plates in
foreign BWRs and cracking discovered on December 8, 1994, at the New
York Power Authority's (NYPA's) FitzPatrick reactor underscore the
Petitioners' concern that additional safety-class components at OCNGS
are degrading; (c) the Licensee did not conduct an enhanced inspection
of the core plate and top guide of the OCNGS facility during the
current outage, despite notification by the General Electric Rapid
Information Communication Service Information Letter (GE RICSIL) 071
dated November 22, 1994; (d) the Licensee, the NRC, and the BWR Owners
Group (BWORG) have failed to provide an analysis of the synergistic
effects of multiple-component cracking of additional safety-class
reactor internal
[[Page 40869]]
components; (e) the time-to-boil calculation is dictated by the amount
of decay heat generated and the volume of water in the fuel pool rather
than the number of reactors at a site that store irradiated fuel in a
separate pool; (f) NRC documents state that the time-to-boil
calculation for FitzPatrick following a loss-of-coolant accident is 8
hours, and NYPA documents state that the time-to-boil calculations in
two cases are 11.86 and 5.36 hours. Finally, nothing indicates that the
time-to-boil calculation at OCNGS is longer than the time-to-boil
calculation at the Susquehanna facility; and (g) the NRC and the
licensee have failed to establish whether redundant components and
power supplies to the OCNGS fuel pool cooling system have been
qualified as Class 1E systems.
The Petitioners' requests that the Commission immediately suspend
the OCNGS operating license were denied in my letter of October 27,
1994, to the Petitioners, because (1) OCNGS was in a refueling outage,
had inspected core shroud welds, and was making structural
modifications before restart of the unit to address some weld cracks
found during the inspection, and (2) inspections and corrective actions
recommended by General Electric Company and the American Society of
Mechanical Engineers Boiler and Pressure Vessel Code for various
reactor internals had been and continued to be performed by the
Licensee.
The Petitioners' request for treatment of their letter of December
13, 1994, as a formal appeal of the NRC staff's denial of their request
of September 19, 1994, for immediate suspension of the OCNGS operating
license, was denied in my letter of April 10, 1995, to the Petitioners.
The Petitioners provided no basis for revisiting the denial of their
request of September 19, 1994, for immediate suspension of the license.
As discussed below, the Licensee completed all ASME Code Section XI
reactor vessel internal inspections and BWROG recommended inspections
and took appropriate remedial action before re-start of OCNGS in
December 1994. The NRC staff was also aware of the potential problem
for United States BWRs raised by cracking in top guide and core plates
of foreign BWRs before the restart of OCNGS. The NRC staff determined,
as explained below, that cracks in these components would not adversely
affect safety of the plant because of differences in the OCNGS design
as compared to the affected foreign reactors.
Regarding the OCNGS spent fuel pool cooling system capability, the
staff determined that the time to the onset of spent fuel pool boiling
following a loss of spent fuel pool cooling during periods where the
reactor vessel contains irradiated fuel at single unit BWR sites, such
as OCNGS, is long enough to allow compensatory measures. The
probability of a sustained loss of spent fuel pool cooling creating
adverse environmental conditions that may cause failure of essential
equipment is extremely low. Therefore, the staff has concluded that
immediate action to address the concerns the Petitioners have
identified at OCNGS is not justified. As stated in my letter of October
27, 1994, spent fuel pool safety is being reviewed generically by the
staff and this review has not yet been completed.
The Petitioners' request for a public meeting was denied in my
letter of April 10, 1995.1 The issue of internals cracking has
been discussed at several public meetings, including a public meeting
on November 4, 1994, that a representative of NIRS attended regarding
the OCNGS core shroud. With respect to spent fuel pool cooling, the
staff has held several public meetings and public briefings with the
Advisory Committee on Reactor Safeguards. Summaries of these public
meetings are available in the NRC Public Document Room, the Gelman
Building, 2120 L Street NW., Washington, DC, and at the local public
document rooms for the affected BWR plants. Transcripts of ACRS
meetings are also available.
\1\ In addition, the NRC staff determined, in accordance with
the guidance in NRC Management Directive 8.11, ``Review Process for
10 CFR 2.206 Petitions,'' that an informal public hearing was not
warranted because the Petition did not present new information or a
new approach for evaluating the concerns Petitioners raised.
The NRC staff's review of the issues related to cracking of reactor
internal components, raised by Requests (1) and (2) of the September
19, 1994, Petition, and Request (1) of the December 13, 1994,
supplemental Petition, is now complete. For the reasons set forth
below, the Petition is denied with respect to these requests. A
Director's Decision concerning the issues related to irradiated fuel
pool cooling and fuel pool boiling, raised by Requests (3) and (4) of
the September 19, 1994, Petition and Requests (2), (3), and (4) of the
December 13, 1994, supplemental Petition will be issued upon completion
of the NRC staff's review regarding those matters.
II. Background
Intergranular stress corrosion cracking (IGSCC) of BWR internal
components has been identified as a technical issue of concern by both
the NRC staff and the nuclear industry. The core shroud is among the
internal reactor components susceptible to IGSCC. Identification of
cracking at the circumferential beltline region welds in several plants
during 1993 led to the publication of NRC Information Notice (IN) 93-
79, ``Core Shroud Cracking at Beltline Region Welds in Boiling-Water
Reactors,'' issued on September 30, 1993. Several licensees inspected
their core shrouds during planned outages in the spring of 1994 and
found cracking at the circumferential welds. The NRC has closely
monitored these inspection activities. Additionally, licensees have
inspected other BWR reactor vessel internal components as discussed
below. NRC issued IN 94-42, ``Cracking in the Lower Region of the Core
Shroud in Boiling-Water Reactors,'' on June 7, 1994, and Supplement 1
to IN 94-42, on July 19, 1994, concerning cracking in the core shroud
found at Dresden Unit 3 and Quad Cities Unit 1. IN 95-17, ``Reactor
Vessel Top Guide and Core Plate Cracking,'' issued on March 10, 1995,
concerned reactor vessel top guide and core plate cracking. The NRC has
monitored Licensee inspection activities of these components at the
OCNGS as discussed below.
III. Discussion
A. Petitioners request that the NRC suspend the OCNGS license until
the Licensee inspects and repairs or replaces all safety-class reactor
internal component parts subject to embrittlement and cracking. Nuclear
power reactor licensees, including GPUN, are required by 10 C.F.R.
Sec. 50.55a to implement inservice inspection programs in accordance
with the guidelines of the American Society of Mechanical Engineers
Boiler and Pressure Vessel Code (ASME Code). The scope of the inservice
inspection programs for reactor pressure vessels and their internal
components is prescribed by ASME Code, Section XI, Division 1,
Subsections IWA and IWB. The Licensee is also required by ASME Code,
Section XI, Article IWA-6000, to submit the results of these
inspections to the NRC within 90 days of completion. The NRC staff
performs periodic audits of licensee-implemented inservice inspection
programs to determine compliance with applicable codes and regulations.
These audits are documented in NRC inspection reports, which are
publicly available at the NRC Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and at the local public
document room for the OCNGS located at the Ocean County Library,
Reference
[[Page 40870]]
Department, 101 Washington Street, Toms River, NJ 08753.
The Licensee performed inspections of the OCNGS reactor vessel and
its internal safety-related components in accordance with the
requirements of ASME Code, Section XI, and the NRC staff has reviewed
the Licensee's inservice inspection programs, as discussed below.
Cracking of the core spray piping was first detected during
Licensee inspections at OCNGS in 1978, and its extent has been
evaluated by the Licensee during each subsequent outage. The core spray
piping was repaired in 1978 and 1980. Since that time, additional
visual inspections by the Licensee have not identified any significant
degradation of the piping or of the repairs made to the piping. The
NRC's review of the Licensee's inspection results and disposition
during the 14R outage, documented in NRC Inspection Report 50-219/92-
22, dated March 19, 1993, and a letter to GPUN dated November 18, 1994,
regarding the 15R inspection concluded that the Licensee inspections
and dispositions of core spray system findings were appropriate.
The Licensee first detected cracking of the top guide in 1991 and
has closely monitored it in successive outages. The NRC staff conducted
an inspection in June 1991, and concluded that the Licensee's
disposition of the top guide crack as ``acceptable as is'' was
adequate. The results of the inspection were reported in NRC Inspection
Report 50-219/91-21, dated August 9, 1991. During an NRC inspection
conducted in December 1992 and January 1993, the NRC staff evaluated
the results of a remote visual inspection of the top guide conducted by
General Electric Corporation for GPUN. The staff evaluated the quality
of the Licensee's visual inspection of the top guide and agreed with
the Licensee's determination that the top guide was acceptable to ``use
as is''. The results of the inspection were reported in NRC Inspection
Report 50-219/92-22, dated March 19, 1993.
The Licensee notified the NRC staff during an October 11, 1994,
telephone call that additional cracking in the top guide had been
found. The Licensee also reported that cracks found in earlier
inspections of the top guide had not shown any measurable growth. In
addition, during the refueling outage for Cycle 15 of operation (15R
refueling outage), which began in September 1994, the Licensee assessed
all the cracks that had been identified to ensure they would not
jeopardize the structural integrity or function of the top guide.
It should be noted that the location of the cracks that have been
detected in the OCNGS top guide is different from that in the foreign
reactor cited in the NIRS letter of December 13, 1994, and the subject
of GE RICSIL-071. Moreover, both the top guide and the core plate at
OCNGS are components of a GE BWR while the foreign plant is a non-GE
BWR. Furthermore, the OCNGS core plate is bolted in place, and the top
guide is restrained vertically by hold-down devices and horizontally by
lateral supports. These configurations result in a highly redundant
structure, and even if cracking similar to that observed in the foreign
plant were to occur, it would not adversely affect the safety of the
plant, and these components could still perform their safety-related
functions.
The BWROG has addressed the issue of cracking in the internal
components of reactor pressure vessels by recommending that BWR
licensees perform inspections of various components pursuant to vendor
recommendations of the General Electric Company. Among inspections
recommended by the BWROG are examination of core spray spargers, core
shrouds, top guides, return line nozzles, and in-core instrumentation,
which in the case of OCNGS are the intermediate power range monitors.
The BWROG has also formed the Boiling Water Reactor Vessels & Internals
Project (BWRVIP), chaired by five nuclear industry vice presidents, to
develop a proactive program to address and mitigate cracking in reactor
pressure vessel internal components. NRC staff correspondence with the
BWRVIP, staff evaluation of the BWRVIP generic submittals, summaries of
meetings with the BWRVIP, and staff assessments of plant-specific
submittals in regard to these subjects are also available to the public
for review at the local public document room of each BWR plant.
The Licensee inspected the following safety-related components
during the 15R refueling outage, which began in September 1994: core
spray sparger and annular piping, steam dryer and separator assembly,
core shroud head bolts, core support plate holddown bolts, guide rod
and steam dryer support brackets, feedwater spargers, top guide
assembly, four intermediate-power range monitors, one low-power range
monitor, core shroud brackets, conical support to shell weld, and the
core shroud. Cracking was observed on the core shroud and a steam dryer
bracket, and required repairs to these components were made. Minor
cracking was observed on the core spray piping, a tack weld on the
keeper bolt of the feedwater spargers, and the top guide cross beams.
None of these cracks would have prevented the components from
performing their normal operating and postulated accident functions.
These indications were dispositioned as is. The Licensee submitted
results of its core shroud inspection and its core spray sparger
inspection to the NRC in separate letters, both dated November 3, 1994.
As a result of a conference call on January 19, 1995, the Licensee
submitted a summary of the results of its inspections of reactor vessel
internal components performed during the 15R refueling outage. By a
letter dated March 16, 1995, in accordance with 10 CFR Sec. 50.55a(g)
and ASME Section XI, IWA 6220, (1986 Edition with no addenda), GPUN
forwarded the reports of its inservice inspection activities conducted
during the 15R refueling outage. In the report GPUN lists the
inspections performed and discusses unacceptable indications of certain
components and their disposition. Inservice inspection of reactor
vessel internal components is required by the ASME Code and the
licensee's inservice inspection program for future outages provides
assurance that degradation of components will be detected and
appropriate action will be taken. The documents discussed above are
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street NW., Washington, DC, and at the local public
document room located at the Ocean County Library, Reference
Department, 101 Washington Street, Toms River, NJ 08753.
The Licensee's inspection of the OCNGS core shroud found that one
of the ten circumferential welds (the H4 weld) had indications of
substantial cracking. To ensure shroud integrity under all postulated
accidents, the Licensee elected to install a modification, consisting
of ten stabilizing tie-rods, designed to ensure that the core shroud
would perform its design functions under normal operation and
postulated accidents even if it were to develop 360 deg. through-wall
cracks. The NRC staff reviewed this modification and issued a safety
evaluation on November 25, 1994, which concluded that the core shroud
modification proposed by the Licensee is acceptable and, therefore, is
approved. The safety evaluation is also available at the public
document rooms previously listed.
On the basis of the NRC staff's review of various plant-specific
and industry programs implemented by the Licensee, the NRC staff
concluded that the Licensee took appropriate actions to address
embrittlement and cracking in,
[[Page 40871]]
and thus to ensure the reliability of, the OCNGS reactor vessel
internal components.
Based on the above, the staff has concluded that suspension of the
Oyster Creek Nuclear Generating Station operating license due to
embrittlement and cracking of the reactor vessel internal components is
not warranted. As stated previously, continued monitoring of reactor
vessel internals as required by the ASME Code and the licensee's
inservice inspection program will provide assurance that degradation of
components will be detected and appropriate action will be taken.
B. Petitioners request that the NRC suspend the OCNGS operating
license until the Licensee provides an analysis regarding the
synergistic effects of through-wall cracking of multiple safety-class
components. The majority of reactor internals are fabricated from high-
toughness materials such as stainless steel and were designed with
significant margins on allowable stresses. As such, cracking must be
severe to adversely impact plant safety. It is unlikely that licensee
inspections would not find such severe degradation. In fact,
identification and sizing of the cracks in the H4 location on the OCNGS
core shroud are good examples of the effectiveness of the inspections.
In addition, NRC staff evaluation of the results from internals
inspections performed to date at OCNGS resulted in the conclusion that
ASME Code safety margins have been maintained.
The Licensee has not provided an analysis to NRC that addresses the
synergistic effects of cracking in multiple safety-class components.
The NRC staff does not consider the lack of such an analysis to be a
safety concern because of the inspection requirements that pertain to
reactor internals and the results of inspections performed to date. See
Section III.A, supra.
Continued monitoring of reactor vessel internals as required by the
ASME Code and the licensee's inservice inspection program will provide
information about the structural integrity of reactor vessel internals
in the long term. The NRC has asked the BWR Vessel Internals Project
(BWRVIP), an industry group, to develop an assessment to address
cracking in BWR reactor vessel internals. A report from the BWRVIP is
expected on the long term effects of reactor vessel internals cracking
in late 1995. In addition, the NRC has undertaken a longer term
evaluation of the effects of cracking in multiple reactor vessel
internal components that will be approached with appropriate treatment
of the key variables (safety function, material susceptibility,
loading, environment, etc.).
Based on the above, the staff has concluded that suspension of the
Oyster Creek Nuclear Generating Station license, due to the lack of an
analysis of the synergistic effects of through-wall cracking of safety-
class reactor internal components, is not warranted.
IV. Conclusion
The Petitioners requested that the NRC suspend the operating
license of Oyster Creek Nuclear Generating Station until: (1) the
Licensee inspects, repairs, or replaces, all safety-class reactor
internal components subject to embrittlement and cracking, and (2) the
Licensee provides an analysis regarding the synergistic effects of
through-wall cracking of multiple safety-class components. For the
reasons discussed above, I conclude that the issues raised by the
Petitioners are being adequately addressed and that there is no basis
for suspending the OCNGS operating license or taking the other
requested action. Accordingly, the Petitioners' above-referenced
requests are denied.
A copy of this Partial Director's Decision will be filed with the
Secretary of the Commission for review as stated in 10 CFR 2.206(c).
This Decision will become the final action of the Commission 25 days
after issuance unless the Commission, on its own motion, institutes
review of the Decision within that time.
Dated at Rockville, Maryland, this 4th day of August 1995.
For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 95-19766 Filed 8-9-95; 8:45 am]
BILLING CODE 7590-01-P
