[Federal Register Volume 63, Number 242 (Thursday, December 17, 1998)]
[Notices]
[Pages 69685-69693]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-33467]
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NUCLEAR REGULATORY COMMISSION
Vermont Yankee Nuclear Power Corporation; Vermont Yankee Nuclear
Power Station; Issuance of Director's Decision Under 10 CFR 2.206
[Docket No. 50-271]
Notice is hereby given that the Director, Office of Nuclear Reactor
Regulation, has acted on a Petition for action under 10 CFR 2.206
received from Mr. Jonathan Block on May 27, 1998, and supplemented on
June 9, 1998, concerning the Vermont Yankee Nuclear Power Station.
The Petition requests that the Commission take immediate
enforcement action by suspending the operating license for the Vermont
Yankee Nuclear Power Station, operated by the Vermont Yankee Nuclear
Power Corporation, until the entire facility has been subjected to an
independent safety analysis review similar to the one conducted at the
Maine Yankee Atomic Power Station. As an alternative, Petitioner
requests that the U.S Nuclear Regulatory Commission (NRC) immediately
act to modify the operating license for the facility by requiring that
before restart (1) Vermont Yankee management certify under oath that
all backup safety systems and all security systems are fully operable
and that all safety systems and security systems meet and comply with
NRC requirements; (2) Vermont Yankee be held to compliance with all of
the restart criteria and protocols in the NRC [Inspection] Manual; (3)
Vermont Yankee only be allowed to resume operations after the NRC has
conducted a ``vertical slice'' examination of the degree to which the
new design-basis documents (DBDs) and the Final Safety Analysis Report
(FSAR) accurately describe at least two of the primary safety systems
for the Vermont Yankee reactor; (4) once operation resumes Vermont
Yankee only be allowed to continue operation for as long as it adheres
to its schedule for coming into compliance and completing the DBD and
the FSAR projects; and (5) the NRC hold a public hearing to discuss the
changes to the torus, the Vermont Yankee DBD and FSAR projects, and
Vermont Yankee's scheduled completion of these projects in relation to
operational safety.
As a basis for the request, the Petitioner raised concerns about
the operation of the Vermont Yankee facility, including challenges to
the single-failure criterion, inadequate safety evaluations, potential
over-reliance on Yankee Atomic Electric Company analyses, an inadequate
operational experience review program, high potential for other serious
safety problems, and lack of adequate perimeter security. The
Petitioner also attached four documents prepared by the Union of
Concerned Scientists (UCS). One UCS document, dated May 14, 1998,
provided a review of Vermont Yankee daily event reports (DERs) made
over the previous year as requested by the Citizens Awareness Network,
Inc., (CAN). These DERs are verbal reports made by licensees under 10
CFR 50.72 to the NRC and put in written form by the NRC. Another UCS
document, dated January 29, 1998, was addressed to the NRC Region I
Senior Allegation Coordinator; it discussed a specific concern with NRC
DER 33545 of January 15, 1998, associated with Vermont Yankee water
hammer effects on certain systems. The third document, a UCS letter
dated May 5, 1997, to the NRC Chairman and Commissioners, discussed
mis-located fuel bundle loading errors. The final UCS document attached
was titled ``Potential Nuclear Safety Hazard Reactor Operation With
Failed Fuel Cladding,'' dated April 2, 1998. In the supplement of June
9, 1998, Petitioner asserted that the event on June 9, 1998, at Vermont
Yankee indicated a lack of reasonable assurance that safety-related
systems at Vermont Yankee will perform adequately.
The Director of the Office of Nuclear Reactor Regulation has
determined that the request should be denied for the reasons stated in
the ``Director's Decision Pursuant to 10 CFR 2.206'' (DD-98-13), 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 20555-0001, and at the
Local Public Document Room located at the Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
As provided in 10 CFR 2.206(c) a copy of this Decision will be
filed with the Secretary of the Commission for the Commission's review.
This Decision will constitute 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 7th day of December 1998.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
Director's Decision Pursuant to 10 CFR 2.206
[DD-98-13]
I. Introduction
By a Petition submitted pursuant to 10 CFR 2.206 on May 27, 1998,
Mr. Jonathan M. Block, on behalf of the Citizens Awareness Network,
Inc. (CAN or Petitioner), requested that the U.S. Nuclear Regulatory
Commission (NRC) take immediate action with regard to the Vermont
Yankee Nuclear Power Station operated by the Vermont Yankee Nuclear
Power Corporation (licensee or Vermont Yankee). By letter dated June 9,
1998, Petitioner supplemented the Petition.
In the Petition of May 27, 1998, the Petitioner requested that the
NRC take immediate enforcement action by
[[Page 69686]]
suspending the operating license for the Vermont Yankee facility until
the entire facility has been subjected to an independent safety
analysis review similar to the one conducted at the Maine Yankee Atomic
Power Station. As an alternative, the Petitioner requested that the NRC
immediately act to modify the operating license for the facility by
requiring that before restart in June 1998 (1) Vermont Yankee
management certify under oath that all backup safety systems and all
security systems are fully operable and that all safety systems and
security systems meet and comply with NRC requirements; (2) Vermont
Yankee be held to compliance with all of the restart criteria and
protocols in the NRC [Inspection] Manual; (3) Vermont Yankee only be
allowed to resume operations after the NRC has conducted a ``vertical
slice'' examination of the degree to which the new design-basis
documents (DBDs) and Final Safety Evaluation Report (FSAR) accurately
describe at least two of the primary safety systems for the Vermont
Yankee reactor; (4) once operation resumes, Vermont Yankee only be
allowed to continue operation for as long as it adheres to its schedule
for coming into compliance and completing the DBD and the FSAR
projects; and (5) the NRC hold a public hearing to discuss the changes
to the torus, the Vermont Yankee DBD and FSAR projects, and Vermont
Yankee's scheduled completion of these projects in relation to
operational safety.
By letter dated June 9, 1998, Petitioner renewed its requests for
relief on the basis of an event occurring on June 9, 1998, at Vermont
Yankee and reported by the licensee in Daily Event Report (DER) 34366.
This event involved the automatic shutdown of the reactor because of
problems in the feedwater system. The Petitioner stated that this event
indicated a lack of reasonable assurance that safety-related systems at
Vermont Yankee will perform adequately.
On July 6, 1998, the Director of the Office of Nuclear Reactor
Regulation informed the Petitioner that he was denying the request for
immediate suspension or modification of the operating license at
Vermont Yankee, that the Petition was being evaluated under 10 CFR
2.206 of the Commission's regulations, and that action would be taken
in a reasonable time. In that letter, the Director also denied
Petitioner's request for a public hearing.
On July 9, 1998, in accordance with established staff guidance for
reviewing 10 CFR 2.206 Petitions, the NRC requested that the licensee
address the concerns raised in the Petition and the need to perform the
actions requested by the Petitioner. The licensee responded by letter
dated September 14, 1998, and the information provided by the licensee
was taken into consideration by the NRC staff.
The NRC staff's review of the Petition and its supplement is now
complete. For the reasons set forth below, the Petitioner's remaining
requests are denied.
II. Background
In support of these requests, the Petitioner raised concerns about
the operation of the Vermont Yankee facility, including challenges to
the single-failure criterion, inadequate safety evaluations, potential
over-reliance on Yankee Atomic Electric Company analyses, an inadequate
operational experience review program, high potential for other serious
safety problems, and lack of adequate perimeter security. The
Petitioner also attached four documents prepared by the Union of
Concerned Scientists (UCS). One UCS document, dated May 14, 1998,
provided a review of Vermont Yankee DERs made over the previous year as
requested by CAN. These DERs are verbal reports made by licensees under
10 CFR 50.72 to the NRC and put in written form by the NRC. Another UCS
document, dated January 29, 1998, was addressed to the NRC Region I
Senior Allegation Coordinator; it discussed a specific concern with NRC
DER 33545 of January 15, 1998, associated with Vermont Yankee water
hammer effects on certain systems. The third document, a UCS letter
dated May 5, 1997, to the NRC Chairman and Commissioners, discussed
mislocated fuel bundle loading errors. The final UCS document attached
was titled ``Potential Nuclear Safety Hazard Reactor Operation With
Failed Fuel Cladding,'' dated April 2, 1998. In the supplement to the
Petition of June 9, 1998, Petitioner asserted that the event on June 9,
1998, at Vermont Yankee indicated a lack of reasonable assurance that
safety-related systems at Vermont Yankee will perform adequately.
Many of the DERs have been generated as a result of the licensee's
ongoing review of Vermont Yankee design-basis information, and the
following is a brief history describing this effort. On October 9,
1996, the NRC issued a request for information to licensees pursuant to
10 CFR 50.54(f) regarding the adequacy and availability of design-basis
information. The purpose of this request was to provide the NRC with
added confidence and assurance that nuclear plants are operated and
maintained within the design bases and any deviations are reconciled in
a timely manner. This request was necessary on the basis of NRC's
findings during inspections and reviews that identified broad
programmatic weaknesses that have resulted in design and configuration
deficiencies at some plants, including Millstone. The licensee
responded by letters dated February 14 and March 11, 1997, stating that
although its overall performance in the areas of design and
configuration control was sound, it would undertake a series of actions
designed to provide improved configuration management. These actions
included developing and implementing a design-basis documentation
program and an FSAR verification program. The DBD program at Vermont
Yankee was initiated in the fall of 1996. The NRC staff evaluated the
licensee's response and determined that subsequent inspection in this
area was necessary. From May 5 through June 13, 1997, the NRC staff
performed an architect/engineer (A/E) inspection, Inspection Report
(IR) 50-271/97-201, to evaluate the capability of selected systems to
perform the safety functions required by their design bases, as well as
the adherence of the systems to their respective design and licensing
bases, and the consistency of the as-built configuration and system
operations with the FSAR. The NRC team concluded that the systems
evaluated were capable of performing their intended safety functions;
however, some concerns (apparent violations of NRC requirements) were
identified. IR 50-271/97-10 documented the NRC follow-up inspection
completed in November 1997 and provided the Notice of Violations (NOVs)
associated with the concerns noted in the A/E report. On March 2, 1998,
an enforcement conference was held with the licensee to discuss the
apparent violations of NRC requirements identified in the A/E
inspection. The licensee responded to the NOVs by letter dated May 14,
1998, and the NRC will continue to evaluate the adequacy of the
licensee's corrective actions during future inspections, currently
expected to be completed by the end of 1998.
The licensee's DBD program has identified numerous design-basis
issues, many of which required reporting under 10 CFR 50.71, 10 CFR
50.72, and/or 10 CFR 50.73. In the NRC's systematic assessment of
licensee performance (SALP) for the period January 19, 1997, through
July 18, 1998, issued on August 28, 1998, the NRC staff found that the
[[Page 69687]]
licensee's program to review and document the plant's design basis has
been rigorous, as evidenced by the number and significance of the
issues identified during the development and validation of the system
DBDs. The NRC staff considers that the number and significance of the
issues, some of which required reporting, demonstrate a desirable
situation in which problems are identified and resolved.
The matters raised in support of Petitioner's requests are
discussed below.
III. Discussion
A. Evaluation of Plant Operation With Deficiencies
Petitioner titled this section ``Single-Failure Criterion
Challenged,'' but the discussion focused on the cumulative effect of
deficiencies at Vermont Yankee. Petitioner states that Vermont Yankee's
volume of longstanding deficiencies in safety-related equipment
strongly suggests that the single-failure criterion may have been
violated. In support of this statement, reference is made by the
Petitioner to an evaluation of Vermont Yankee DERs by the UCS dated May
14, 1998. Petitioner also states that it was not able to find any
evidence that Vermont Yankee considered the impact of the cumulative
effect of concurrent degraded conditions on the safety margin of the
plant.
Appendix A to 10 CFR Part 50 gives a definition of the single-
failure criterion. The capability to withstand a single failure is a
consideration in the design of nuclear power plants. For example,
General Design Criterion 35 for emergency core cooling systems in
Appendix A to 10 CFR Part 50 states that suitable redundancy in
components and features shall be provided to assure that the system
safety function can be accomplished, assuming a single failure.
Technical specification requirements must be met. A deficiency in a
safety system, including deficiencies in which the capability to
withstand a single failure is lost, is to be evaluated by licensees and
treated as a degraded and nonconforming condition. A prompt
determination of operability is to be made by licensees. For any
deficiency, including those in which the capability to withstand a
single failure is lost, licensees must evaluate the deficiency and, if
the deficiency affects the design-basis requirements for the particular
plant, correct the deficiency in accordance with 10 CFR Part 50,
Appendix B, Criterion XVI, Corrective Action. The NRC has issued
guidance regarding resolution of deficiencies in the form of Generic
Letter (GL) 91-18, Revision 1, ``Information to Licensees Regarding NRC
Inspection Manual Section on Resolution of Degraded or Nonconforming
Conditions.'' The guidance in Vermont Yankee's corrective action
program is consistent with the NRC's guidance in GL 91-18. Identified
deficiencies are evaluated by the licensee in accordance with the
licensee's corrective action program, which meets the requirements of
10 CFR Part 50, Appendix B. If required by 10 CFR 50.71, 50.72, and/or
50.73 the deficiency is reported to the NRC.
NRC regulations do not explicitly require an integrated assessment
of deficiencies. If a deficiency cannot be immediately corrected, the
licensee evaluates the acceptability of continued operation consistent
with the NRC guidance in GL 91-18. A determination of operability is
needed for each deficiency.
The NRC staff requested and the licensee provided an integrated
assessment of items that were scheduled for final resolution after the
spring 1998 outage by letters to the NRC dated May 1 and May 28, 1998.
IR 50-271/98-06 documented the NRC's review of the licensee's letter of
May 1, 1998, and concluded that the licensee's actions to resolve the
outstanding items, as they pertain to restart of the plant following
the spring 1998 refueling outage, have been appropriate. No concerns
were identified by the NRC staff regarding the operability
determinations, compensatory actions, or corrective actions, as
documented in IR 50-271/98-06.
In summary, deficiencies at Vermont Yankee are entered in the
licensee's corrective action program which meets the requirements of 10
CFR Part 50, Appendix B. The acceptability of continued operation with
outstanding deficiencies is evaluated using the NRC guidance in GL 91-
18. The NRC has been aware of the events and deficiencies referred to
by the Petitioner as the basis for its concern. The staff assessed the
DERs and concluded an appropriate response would be to inspect licensee
activities. The results of the NRC review are documented in NRC
inspection reports. For example, NRC IR 50-271/98-06 documented the
NRC's inspection of the licensee's engineering and technical support
for operations as they pertain to the licensee's process for evaluating
deficiencies and determining the acceptability of continued operation
with the deficiency. No concerns were raised with regard to operability
determinations, compensatory actions, or corrective actions. No
additional NRC actions were deemed necessary in this area.
B. Inadequate Safety Evaluations
Petitioner states that there is evidence that the Vermont Yankee
licensee performed inadequate safety evaluations required by 10 CFR
50.59 and listed DERs 31906, 31949, 32106, and 34005 as examples.
The licensee stated in its response of September 14, 1998, to the
Petition that the examples cited are similar in that their cause can be
traced to the difficulty in quickly retrieving the specific design-
basis information in the time period available to determine system
operability. Had the design bases been readily retrievable, it is
unlikely that these issues would have constituted a condition requiring
reporting. The licensee has recognized the need to upgrade the DBDs and
is currently performing this action, as previously discussed.
In Inspection Report 50-271/98-12, the NRC reviewed the four event
reports listed by the Petitioner as examples of inadequate safety
evaluations at Vermont Yankee. DER 34005 was found to not involve an
inadequate safety evaluation. In this case, the licensee was not able
to immediately retrieve a necessary design-basis calculation for the
anticipated transient without scram (ATWS) mitigation system.
Subsequently, the licensee found that the calculation had been
performed by their fuel vendor and was in fact available. The licensee
retracted that event report due to the retrieval of this calculation.
DERs 31906, 31949 and 32106 were each partially a result of inadequate
design-basis information being available. This led to safety
evaluations in support of modifications to plant RHR system operating
procedures and installation of fire protection hardware that were
erroneously found acceptable. The licensee notified the NRC of these
three conditions in March and early April 1997.
At the time of discovery, the licensee was implementing their
Individual Plant Examination of External Events (IPEEE) program. This
special review revealed errors in both the original design of the
plant, as well as weak documentation of certain design bases that led
to the prior acceptance of these plant vulnerabilities to external
event initiated internal flooding events. The licensee appropriately
reported these conditions to the NRC and took necessary corrective
actions to remove the identified vulnerabilities. Since the conditions
had not occurred that were necessary to exploit these plant
vulnerabilities, such as a seismic event, no adverse safety
consequences were
[[Page 69688]]
realized even though the plant had operated outside of the design
bases.
The Licensee Event Reports (LERs) associated with DERs 31906, 31949
and 32106 (LER 50-271/96-012 and 50-271/97-004, respectively) were
reviewed by the NRC in Section E8.3 of IR 50-271/97-10. In that report,
the NRC concluded that the licensee's root cause analyses and
corrective actions were acceptable and that these issues met the
criteria for handling as non-cited violations per Section VII.B.3,
``Old Design Issues,'' of the NRC Enforcement Policy.
Subsequent to the licensee notifying the NRC of these events, the
NRC performed two major engineering/design inspections at the Vermont
Yankee plant. The A/E team inspection in June 1997, concluded that
there were weaknesses in the design control process; but, that the
licensee was to address these deficiencies in their Configuration
Management Improvement Project. In the engineering team follow-up
inspection of November 1997, the NRC concluded that the licensee had
strengthened its design bases documentation validation process as a
result of the lessons learned from the A/E inspection. Further, the NRC
found that the licensee had adjusted the depth and breadth of its
validation inspection using the Safety System Function Inspection
techniques, similar to those used in the A/E team inspection, and
concluded that its validation efforts should produce results similar to
the A/E team review. The inspection results also included a number of
findings, some of which were design bases control violations that
resulted in a Civil Penalty issued in April 1998.
In response to the Civil Penalty, the NRC determined that the
licensee's corrective actions were sufficient to identify and resolve
existing design bases errors. As a result of the licensee's
comprehensive corrective actions, the NRC concluded that no additional
measures were warranted for the design bases concerns at Vermont
Yankee. The NRC will continue to monitor and assess the licensee's
progress in completing their proposed corrective actions as part of the
regular inspection process for follow-up to identified violations.
The NRC has recently assessed the licensee's performance in the
area of safety evaluation as documented in IR 50-271/98-80 issued on
July 16, 1998. The NRC reviewed the licensee's procedural guidance for
the safety evaluation program to assess that program against the latest
guidance contained in NRC Inspection Manual 9900 and the regulatory
requirements of 10 CFR 50.59. In addition, selected safety screenings
and safety evaluations were reviewed. Although some deficiencies were
noted, neither the deficiencies noted in the report, nor the examples
referenced in the Petition constitute a condition warranting further
extensive inspection in this area. The licensee's corrective actions
for the deficiencies noted in IR 50-271/98-80 will be evaluated during
future inspections.
C. Potential Over-Reliance on Yankee Atomic Electric Company Analyses
Petitioner states that there is evidence that the Vermont Yankee
licensee has been relying upon Yankee Atomic Electric Company (YAEC) to
conduct engineering analyses, and there is a potential that Vermont
Yankee may have the same kind of serious compromises in safety systems
that existed at other facilities that relied upon YAEC's engineering
analyses. Petitioner refers to an NRC demand for information (DFI) to
YAEC regarding information needed by the NRC to determine whether
enforcement action should be taken against YAEC to ensure future
compliance, on the part of NRC licensees, with NRC requirements. DERs
31915, 32106, 33259, 33502, and 34145 were listed by the Petitioner as
those that may have involved analyses by YAEC. Petitioner requested
that the NRC suspend Vermont Yankee's license to operate until
assurance can be obtained that all analyses that YAEC prepared for
Vermont Yankee have been reviewed by the NRC staff to ensure that they
have been performed properly.
The NRC staff acknowledges that YAEC performed many engineering
analyses for Vermont Yankee.
The serious compromises (according to the Petitioner) in safety
systems that existed at other facilities that relied upon YAEC's
engineering analysis to which the Petitioner refers originated with an
allegation involving YAEC's analyses performed for Maine Yankee Atomic
Power Company (MYAPCo). A letter dated December 1, 1995, from the UCS
contained an anonymous allegation that certain analyses performed by
YAEC for MYAPCo were flawed. A number of investigations and technical
reviews were initiated, and the NRC issued a DFI to YAEC and Duke
Engineering & Services, Inc. (DE&S),1 in December 1997. The
DFI required an explanation why the NRC should permit any NRC licensee
to use the services of YAEC and/or DE&S to perform loss-of-coolant
accident (LOCA) analyses or any safety-related analyses to meet NRC
requirements. The DFI was issued on the basis of NRC's concerns
regarding specific inadequacies in small-break LOCA analyses provided
by the YAEC LOCA Group to MYAPCo that caused MYAPCo to be in violation
of NRC requirements. DE&S responded on February 27, 1998, to the NRC's
DFI regarding continued engineering services to nuclear utilities. The
response provided a detailed description of the reviews that had been
conducted and the associated findings. NRC subsequently issued
violations to MYAPCo on October 8, 1998.
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\1\ DE&S acquired portions of YAEC, including the YAEC LOCA
Group, in December 1997.
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After review of the complete record in this matter, the NRC staff
concluded that the actions taken by the YAEC LOCA Group caused MYAPCo
to be in violation of Commission requirements in a number of areas, but
that these actions did not result from willfulness on the part of DE&S
and/or YAEC employees.2 The staff further concluded that the
corrective actions accomplished and planned, as discussed in the DE&S
response to the DFI, provide a basis for reasonable assurance that in
the future, the NRC and licensees can rely upon DE&S to provide
complete and accurate information and that DE&S is willing and able to
otherwise conduct its activities in accordance with the Commission's
requirements. Therefore, the NRC staff determined that no further
enforcement action shall be taken against YAEC or DE&S regarding the
actions of the LOCA Group of concern in the DFI.
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\2\ The NRC staff addressed its final conclusions regarding the
SBLOCA analysis violations at Maine Yankee in the NOV issued to
MYAPCo on October 8, 1998. The NRC staff's conclusions regarding the
provision of LOCA analyses or other safety-related analyses to NRC
licensees by YAEC and/or DE&S are discussed in letters to YAEC and
DE&S dated October 8, 1998.
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In reaching these conclusions, the NRC staff considered the entire
record of investigations and technical reviews that resulted in part or
in whole from the allegation of December 1995. The broader implication
of the allegation, beyond the specific analysis performed for Maine
Yankee, suggested cause for concern in two areas. First, there was a
concern regarding the adequacy of LOCA analyses provided to other NRC
licensees, including Vermont Yankee, by the YAEC LOCA Group. Secondly,
it also suggested cause for concern regarding the adequacy of other
safety-related analyses performed by the Yankee Nuclear Services
Division of YAEC on behalf of NRC licensees to demonstrate compliance
with Commission requirements.
[[Page 69689]]
Regarding the first concern, in May 1996 the NRC staff audited the
LOCA analyses provided to Vermont Yankee by the YAEC LOCA Group. This
review also incorporated a concern regarding the conditions and events
leading to Vermont Yankee's LER No. 96-010 dated May 9,
1996.3 The review concluded that the analyses performed by
the YAEC LOCA Group for Vermont Yankee were consistent with the
conditions on the use of the RELAP5YA code for Vermont Yankee as
specified in the staff's safety evaluations for the code dated August
25, 1987, and October 21, 1992. Note that the RELAP5YA code was a BWR
version and was different than the Maine Yankee version, a pressurized
water reactor version. Since the staff's approval of the use of the
code, the staff found that the code had been transferred to a different
computer operating system and that the fuel behavior package had been
modified. The staff reviewed these changes and concluded that approved
quality assurance procedures were followed throughout the code
modifications.
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\3\ LER 96-010 was associated with an inadequate design/single
failure evaluation during a design change. The NRC staff found that
the plant-specific analysis had failed to consider the limiting
single-failure scenario. This issue was addressed by the staff in an
NOV and Proposed Imposition of Civil Penalty--$50,000, dated August
23, 1996. The staff concluded that this violation resulted from
ineffective communications between the plant operations staff and
the YAEC safety analysts, resulting in failure to identify the fact
that the safety analysis assumptions were not consistent with the
plant configuration. In its response to the DFI, DE&S noted that
ineffective communication between YAEC, MYAPCo, and the NRC also
played an important role in the assumptions of all parties regarding
the demonstration of compliance with the technical requirements of
10 CFR 50.46. DE&S identified corrective actions to clearly define
and formally document regulatory and organizational interface
requirements with its nuclear clients to prevent recurrence of the
communication and organizational responsibility uncertainties that
contributed to the events described in the DFI.
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Regarding the second concern, the Independent Safety Assessment
(ISA) of MYAPCo conducted in the summer of 1996 evaluated non-LOCA
safety-related analyses performed by YAEC on behalf of MYAPCo. As
stated in the ISA report dated October 7, 1996, the ISA concluded that
conditions of approval in NRC safety evaluations were met in the use of
selected analytic codes for performing non-LOCA safety-related
analyses, but that weaknesses in documentation and validation
represented vulnerabilities that warranted licensee attention. The ISA
also concluded that cycle-specific core performance analyses were
excellent, but that weaknesses were found in more complicated, less
frequently performed analyses. These weaknesses did not cause the
analyses results to exceed the facility design and licensing bases. In
its response to the DFI, DE&S described corrective actions, including
strengthened personnel training; formal documentation of organizational
roles, responsibilities, and communication requirements; and
independent assessment to provide management with direct feedback on
the compliance of work process, practices, and products. These
corrective actions address the weaknesses identified by the ISA in
documentation, validation, and the conduct of complicated, infrequently
performed analyses.
In its letter of September 14, 1998, the Vermont Yankee licensee
indicated that the conclusions reached on the basis of the reviews
conducted give confidence that the analyses performed by YAEC on
Vermont Yankee's behalf are of high quality. The Vermont Yankee
licensee reviewed the concerns raised by the DFI for potential impact
on Vermont Yankee. The licensee indicated that an independent technical
assessment of specific analyses performed for Vermont Yankee was
conducted and stated that the assessment identified no significant
technical errors. The licensee did not uncover any reason to suspect
the quality or the accuracy of engineering analyses performed by YAEC
for Vermont Yankee.
On the basis of the results of several NRC staff investigations and
technical reviews, the NRC staff has concluded that the violations
associated with the SB LOCA analyses provided to MYAPCo by the YAEC
LOCA Group were isolated. LOCA analyses and other safety-related
analyses provided to NRC licensees by YAEC and/or DE&S, including the
LOCA Group, have generally been found to be in compliance with NRC
requirements. Therefore, the actions requested by the Petitioner are
not necessary.
With respect to future work by DE&S, weaknesses or vulnerabilities
identified during these reviews are being addressed by DE&S. Therefore,
the NRC staff has concluded that there is no basis for taking action
against DE&S and/or YAEC to prevent them from providing safety-related
analysis services to NRC licensees, nor to take action against NRC
licensees, including Vermont Yankee, to prevent them from using the
engineering services provided by YAEC and/or DE&S.
D. Inadequate Operational Experience Review Program
Petitioner states that there is evidence that strongly suggests
that the Vermont Yankee licensee does not have an adequate operational
experience review program and listed DERs 31923, 32016, and 33789 as
examples of inadequacy and violation of NRC regulations. Petitioner
states that an inadequate operational experience review program leads
to ``compromised engineering conservation in safety systems, and the
eventual failure of such systems during a serious emergency event.''
The licensee acknowledges that weaknesses have been identified in
the reviews of industry operation experiences in that reviews were not
always timely and some opportunities to learn from industry operating
experiences were sometimes missed. A task force was developed to
address the weaknesses.
The NRC assessed licensee performance in this area on September 6,
1997, and documented the findings in IR 50-271/97-06. The NRC concluded
that the previous weaknesses identified in the licensee's operating
experience review process had been appropriately addressed through
implementation of a new administrative procedure. This report also
stated that a selected sample of recently dispositioned items
identified that a proper review of the individual concerns had been
made and that closure of the individual concerns had been achieved.
In IR 50-271/98-12, the NRC reviewed the three event reports listed
by the Petitioner as examples that the licensee does not have an
adequate operational experience feedback (OEF) review program. On March
10, 1997, DER 31923 was identified as a result of the licensee's IPEEE
program. The licensee determined that the root cause of this event was
an inadequate initial design. Related to this cause was an inadequate
flood design bases. This contributed to the licensee's failure to
identify this concern during earlier design studies, including those in
response to NRC Information Notices on similar events in the industry.
The licensee's IPEEE program was a very detailed and intrusive review
that questioned design basis assumptions. Due to the scope of that
review, this concern as well as several other flooding design concerns
were discovered by the licensee. The root cause and corrective actions
for this event were described in LER 50-271/97-002. This LER was
previously reviewed in Section E8.3 of IR 50-271/97-10. In that report,
the NRC concluded that the licensee's root cause analyses and
corrective actions were acceptable and that this issue met the NRC
Enforcement Policy for handling as a non-cited violation per Section
VII.B.3, ``Old Design Issues.''
DERs 32016 and 33789 were found to be related. The earlier of these
two
[[Page 69690]]
events was discovered on March 25, 1997, as a result of the licensee's
operational experience feedback review of an event report by Lasalle on
February 21, 1997. After this initial discovery, the licensee took
appropriate corrective measures to ensure that the standby gas
treatment system would not be operated in a configuration that could
lead to failure of the system during a design basis accident. The
licensee prematurely removed the corrective actions, which resulted in
a second event with the standby gas treatment system operated in a
configuration that could lead to failure. The NRC issued a violation in
IR 50-271/97-06 for this second event. The licensee attributed the
cause of this second event to a weakness in the license and design-
basis information for this system. The licensee appropriately reported
this event to the NRC in LER 50-271/97-014.
As a result of additional engineering review committed to as a
corrective action listed in LER 50-271/97-014, the licensee discovered
an additional vulnerability for the standby gas treatment system that
was subsequently reported to the NRC on February 25, 1998, in DER
33789. The NRC concluded that this latter event was not a result of
ineffective operational experience review, but rather a result of the
corrective actions for an identified problem.
The NRC concluded that these event reports were a result of
original design deficiencies, and related weaknesses in the design and
licensing basis information for the plant systems in question. The root
causes of these events did not raise concern with the adequacy of the
licensee's current OEF review program, as discussed in IR 50-271/97-06.
Except for DER 33789, which was a result of the licensee's corrective
actions program, these events predated the licensee's revised OEF
program as discussed in IR 50-271/97-06. Also, one of the events was
licensee identified by use of the OEF process.
The DERs referenced by the Petitioner do not constitute a failure
of the operational experience review program. On the basis of NRC's
previous inspection in this area, the licensee has an adequate industry
operational experience review program. Follow-up on the effectiveness
of the licensee's operational experience program remains an item of
routine review for the NRC inspection staff.
E. High Potential for Other Serious Safety Problems
Petitioner states that since Vermont Yankee's safety evaluation and
operational experience review program do not seem adequate, and since
it has relied on YAEC engineering analyses, it is reasonable to expect
that there are many more design and licensing-bases problems yet to be
dealt with at Vermont Yankee. Petitioner states that the NRC required
Salem and Millstone reactor licensees to certify that the safety-
related systems at these facilities were within their design and
licensing basis before permitting them to be restarted when pervasive
and systemic problems very similar to those at Vermont Yankee were
identified at these facilities.
As stated in the ``Background'' section of this Director's
Decision, the A/E inspection conducted at Vermont Yankee was performed
as a follow-up on the design-basis problems noted at facilities,
including Millstone. As previously stated, the NRC team concluded that
the systems evaluated were capable of performing their intended safety
functions. The concerns identified were not of the significance of
those observed at Millstone.
Salem Units 1 and 2 were shut down in May and June 1995
respectively because of inadequate control room ventilation, and
because of problems with a minimum flow valve that made the residual
heat removal system inoperable. Before the shutdown, both Salem units
were the subject of significant regulatory attention because of a
series of performance problems dating back to 1990. Additionally, NRC
Augmented Inspection Teams were dispatched to the Salem units every
year between 1991 and 1994 to evaluate significant operational events,
including a catastrophic turbine-generator failure and control rod
system failures. The NRC was concerned about Salem operation because of
frequent equipment failures and personnel errors and failure of
previous initiatives to achieve long-term performance improvement. In
June 1995, the Region I Regional Administrator issued a confirmatory
action letter confirming the licensee's commitment to develop a long-
term plan to identify and correct the longstanding equipment
deficiencies and address the poor condition of materials, weak
management oversight, and ineffective corrective actions.
The magnitude of problems that existed at Salem have not been
observed at Vermont Yankee. As previously stated, the NRC considers
that the licensee's safety evaluation and operational experience review
program are adequate on the basis of NRC's inspections. In addition,
the NRC has not identified any significant concerns with the YAEC/DE&S
analysis for Vermont Yankee that warrant the actions requested by the
Petitioner.
The Vermont Yankee licensee is conducting a DBD and FSAR review
that examines safety-related systems to identify and correct design and
licensing-basis problems. Plant operation may continue during these
assessments, provided the plant is operated in accordance with its
license and NRC's regulations. Deficiencies identified are entered into
the corrective action process and operability is determined using
guidance similar to that contained in NRC GL 91-18 as discussed
previously.
In our recent SALP IR 50-271/98-99, dated August 28, 1998, the NRC
concluded that licensee management established a lower threshold for
problem reporting, thereby improving problem identification.
Particularly noteworthy was management's implementation of the
Configuration Management Improvement Project, which improved
identification of design and licensing issues. The activities have been
rigorous, as evidenced by the number and significance of the issues
identified during the development and validation of the system DBDs.
The NRC considered the licensee's performance in engineering to be
good. The SALP was based on the results of numerous NRC inspections at
Vermont Yankee, including a major design (A/E) inspection of certain
systems. On the basis of our recent assessment of engineering at
Vermont Yankee, the staff concluded that the actions requested by the
Petitioner are not warranted.
F. Lack of Adequate Perimeter Security
Petitioner states that Vermont Yankee's lax perimeter security
demonstrates that management did not adequately respond to all of the
implications of the recent incident involving a former Vermont Yankee
contractor. On August 19, 1997, this former contractor was involved in
shootings in New Hampshire and Vermont that left four people dead. The
individual was subsequently killed in a confrontation with Vermont law
enforcement authorities. Law enforcement authorities later found bomb-
making materials stored at the individual's residence. Petitioner
states that NRC inspectors recently discovered a major weakness in the
security system by having five out of eight inspectors successfully
invade the security perimeter, including one inspector who passed
through the metal detector with a gun.
The NRC conducted a special inspection at Vermont Yankee on August
27 and 28, 1997, to determine if
[[Page 69691]]
the access authorization program, access controls, and fitness for duty
program, as implemented, revealed information that should have
prevented the individual involved in the shootings of August 19, 1997,
from being granted unescorted access. The NRC determined that the
licensee's program met regulatory requirements. The NRC did not
identify any information used by the licensee in processing the
individual for access authorization that should have prevented the
licensee from granting the individual unescorted access to the secured
portions of the plant. The results of the inspection are documented in
IR 50-271/97-07. No changes or corrective actions to the licensee's
program were found to be necessary.
The NRC conducted a physical security inspection at Vermont Yankee
on March 16-19, 1998, as documented in IR 50-271/98-05. This inspection
concluded that within the scope of the inspection, the Vermont Yankee
licensee had in place a satisfactory program for the protection of
public health and safety. However, two violations of regulatory
requirements associated with access control of packages and the
intrusion detection (perimeter security) system were identified. The
violations were categorized as Severity Level IV violations in
accordance with the NRC enforcement policy and are discussed below.
Performance testing of the intrusion detection system by the NRC
regional assistance team resulted in the assistance team's successfully
gaining undetected access into the protected area by climbing over the
protected area barrier without generating an alarm in 6 of 10 zones.
This weakness constituted a violation of NRC requirements. The licensee
took adequate corrective actions for the violation by immediately
implementing compensatory measures and adjusting all fence zone
sensors. All zones subsequently successfully detected deliberative,
non-aggressive climbing attempts by a specially selected security force
member. A specifically defined non-aggressive climb test was
incorporated into regularly scheduled operability testing of the
system. Despite this violation, the NRC concluded that the licensee's
security facilities and equipment were well maintained and reliable on
the basis of inspection, testing, maintenance, compensatory measures,
protected area detection aids, and assessment aids.
During the performance testing of the personnel and package search
equipment, a test device was placed in a backpack with other items and
placed on the x-ray machine. The x-ray machine detected an object in
the backpack that could not be identified and the backpack was
physically searched by a security force member. However, the test
device was not discovered during the physical search, constituting a
violation of NRC requirements. The licensee took adequate corrective
actions, including counseling and retraining the search officer
involved, as well as assessing the hand search practices utilized by
other security officers. Lessons learned and performance expectations
were also communicated to each individual member of the security force.
The NRC concluded that the licensee was conducting its security and
safeguards activities in a manner that protected public health and
safety on the basis of the inspection of the access authorization
program, alarm stations, and access control of personnel and packages
in the protected area despite the violation in this area.
The licensee had adequately addressed the issues raised by IR 50-
271/98-05 violations. The NRC performed a follow-up inspection
described in IR 50-271/98-12 during the week of August 31, 1998, which
included an evaluation of the licensee's corrective actions for the
violations and found the corrective actions acceptable. NRC's SALP
report dated August 28, 1998, considered these issues and concluded
that site management continued to provide appropriate oversight of the
security program. These violations were not related to the situation
involving the former Vermont Yankee contractor previously discussed.
Therefore, since these situations are not related and no changes or
corrective actions to the licensee program were necessary following the
former contractor issue, the NRC considers that Petitioner's statement
that lax perimeter security demonstrates that management did not
adequately respond to all of the implications of the recent incident
involving a former Vermont Yankee contractor is not valid.
G. Operation Conditional Upon the DBD and the FSAR Schedule
Petitioner stated that Vermont Yankee should be allowed to operate
only if it meets the scheduling obligations it set up for completing
DBDs and updating the FSAR (by imposition of a license condition or
Order). The Petition stated that Vermont Yankee's lagging efforts at
regulatory compliance easily justify this action.
As previously stated, on October 9, 1996, the NRC issued a request
for information pursuant to 10 CFR 50.54(f) regarding the adequacy and
availability of design-basis information. By letters dated February 14
and March 11, 1997, the licensee responded to the request for
information. The licensee committed to a series of actions designed to
provide improved configuration management (adequacy and availability of
design-basis information). These actions included a DBD program and an
FSAR verification program. The A/E inspection previously discussed, IR
50-271/97-201, was conducted to review particular aspects of the
licensee's design control programs and processes. The DBD and the FSAR
verification programs were originally scheduled to be completed by
October 1998 and December 1998, respectively. The NRC understands that
these programs require extensive use of engineering resources and that
the scheduled date for completion of these programs may be delayed. The
NRC staff has concluded that licensee management has placed an
appropriately high emphasis on the configuration management improvement
project, which includes the DBD and the FSAR verification programs. A
delay in the licensee's implementation would not necessarily constitute
a condition warranting a license condition or imposition of an Order.
The NRC staff currently believes that an adequate time frame for
completion of the FSAR verification programs is March 30, 2000, for
structures, systems, and components of high safety significance as
defined in the licensee's maintenance rule, and March 30, 2001, for all
other information. Delayed completion of these programs may be subject
to enforcement.
With respect to Vermont Yankee's regulatory compliance, compliance
issues have been appropriately addressed by the NRC and the licensee as
previously discussed. In the SALP report issued on August 28, 1998, the
NRC concluded that licensee performance has been good in all functional
areas, which reflects NRC's assessment of regulatory compliance during
the period of January 19, 1997, to July 18, 1998. On the basis of this
information, the NRC has determined that the requested action is not
necessary.
H. Necessity for a ``Vertical Slice'' Safety Assessment
Petitioner states that a ``vertical slice'' safety assessment on at
least two systems for which the licensee has completed review is
necessary to be certain that Vermont Yankee's DBD and FSAR projects
have accurately captured
[[Page 69692]]
the actual operating condition of the facility's safety systems. By
``vertical slice,'' the Petitioner appears to be referring to an
inspection similar to the A/E inspection previously performed and
documented in IR 50-271/97-201. Petitioner references statements made
during the enforcement conference on March 2, 1998, between the NRC and
the licensee following the NRC A/E inspection, which discussed the
process that the licensee was using in the DBD validation process.
This area was evaluated by the NRC and documented in IR 50-271/97-
10. The NRC had been concerned that at the time of the A/E inspection,
it did not appear that the DBD reviews would have identified the design
issues found by the NRC team based on an initial review of the
licensee's design-basis efforts. At the enforcement conference meeting
on March 2, 1998, the licensee stated that it had committed to perform
the DBD reviews and recognized the need for DBD validation prior to
issuance of the NRC's 10 CFR 50.54(f) letter regarding the adequacy and
availability of design-basis information. However, the validation
effort had not been fully defined at the time of the A/E inspection.
The licensee stated that the validation effort would have been designed
to identify the type of problems found by the A/E team. On the basis of
the findings of the follow-up inspection completed in November 1997 (IR
50-271/97-10) and the information provided at the March 1998 meeting,
the NRC was no longer concerned with DBD validation effort. The NRC
staff documented this conclusion by letter dated April 14, 1998, which
issued the NOV and civil penalty related to the A/E inspection and IR
50-271/97-10. The SALP report issued August 28, 1998, concluded that
overall the activities in this area have been rigorous, as evidenced by
the number and significance of the issues identified during the
development and validation of the system DBDs.
The NRC considers that the licensee's efforts in this area are
adequate, and allocation of additional NRC resources to perform an
additional ``vertical slice'' safety assessment is unnecessary at this
time. The NRC will continue to evaluate the adequacy of the licensee's
corrective actions for the violations identified during the A/E
inspection in future inspections.
I. Conduct of a Public Hearing in Brattleboro, Vermont To Inform the
Public
Petitioner requested that the NRC conduct a public hearing in
Brattleboro, Vermont, to inform the public about changes to the torus,
compliance with the DBD and the FSAR process, results of the A/E
inspection, results of an NRC ``vertical slice'' analysis of Vermont
Yankee's first sets of DBDs, and the implications for public health and
safety of Vermont Yankee's schedule for complying with the requirements
that it verify and update all DBDs and the FSAR.
The NRC has conducted several public meetings on many of these
issues. In addition, the NRC conducted a public meeting in Brattleboro,
Vermont, on September 16, 1998, to discuss the results of the latest
SALP for Vermont Yankee. Following the meeting with the licensee, the
NRC met with members of the public, including members of the
Petitioner's organization, to discuss any issues that members of the
public wished to discuss. Both the July 6, 1998, NRC letter to the
Petitioner and the SALP public meeting notice indicated that NRC
officials would be available following the SALP meeting. Issues
discussed with members of the public included those described by the
Petitioner. Further commitment of NRC staff resources to conduct the
requested hearing is not warranted.
J. Review of Vermont Yankee Daily Event Reports
Petitioner attached to the Petition a letter dated May 14, 1998,
from the UCS to the Petitioner that contained a review of DER
information at Vermont Yankee and provided general observations and
conclusions. Concerns raised included the single-failure criterion,
inadequate safety evaluations, potential over-reliance on YAEC, and the
program to review inadequate operational experience. These issues were
addressed earlier in this Director's Decision. The conditions
documented in the DERs have been addressed by NRC inspection follow-up
when appropriate and no additional action is necessary.
K. Concern About Water Hammer Effects on Certain Systems
Petitioner attached a document titled ``Vermont Yankee HPCI/RCIC
[High Pressure Coolant Injection Reactor Core Isolation Cooling]
Waterhammer, DER 33545,'' dated January 29, 1998, to David J. Vito,
Senior Allegation Coordinator for the NRC, from the UCS.
The document questioned (1) whether the Vermont Yankee FSAR
analyses assume that HPCI and RCIC start and stop, and, if so, is
suppression pool temperature such that conditions for water hammer
exist; (2) whether the FSAR appropriately documents the existence (and
related design and licensing basis) of the vacuum breakers in the HPCI
and RCIC exhaust lines; and (3) whether the related Vermont Yankee LER
should discuss the risk to the public from two fission product barriers
being degraded (the fuel cladding due to known leaking fuel at Vermont
Yankee, and the primary containment boundary due to potential water
hammer).
In response, the NRC reviewed Vermont Yankee's subsequent LER 98-
05, issued on April 9, 1998, and performed inspection activities at
Vermont Yankee in June 1998, as described in IR 50-271/98-80. The NRC
review found that the effect of the suppression pool air space pressure
was not adequately considered in the original HPCI and RCIC vacuum
breaker design. However, the NRC also concluded that the forces
associated with the potential water hammer transients caused by this
design issue would not have challenged the structural integrity of the
piping.
Although the previous vacuum breaker design was not adequately
described in the FSAR, earlier versions of HPCI and RCIC piping and
instrument diagrams did accurately reflect the installed configuration.
A subsequent modification to correct the design deficiency shows that
controlled drawings, the DBDs for HPCI and RCIC, and the FSAR have been
or will be updated to reflect the newly installed vacuum breaker
configurations. The NRC also sampled design changes since 1974 related
to HPCI and RCIC and found none that would have influenced the piping
configuration in question. Further, the DBD prepared for each system
represents a comprehensive evaluation of past modifications and design
information. In January 1998, during the preparation of the HPCI and
RCIC DBDs, the licensee identified the vacuum breaker deficiency.
Therefore, on the basis of the NRC's and the licensee's reviews, there
is reasonable assurance that no past evaluations would have been flawed
as a result of the lack of discussion in the FSAR.
Regarding the content of LER 98-05, the NRC concluded that the
potential water hammer forces would not have been high enough to
challenge pipe structural limits and, therefore, containment integrity.
Regarding the fuel cladding, the leakage experienced in the last cycle
of operation was limited to a single fuel rod bundle, and was within
the operational limits of the Vermont Yankee technical specifications
(TSs) and well below that assumed in the FSAR accident analysis. As
such, no significant increase in risk was presented in this
circumstance.
[[Page 69693]]
L. Mislocated Fuel Bundle Loading Errors
Petitioner also attached a letter dated May 5, 1997, from the UCS
to the NRC regarding ``Mislocated Fuel Bundle Loading Error.'' The
letter urges NRC to revisit the misoriented and mislocated fuel bundle
loading issues for boiling-water reactors (BWRs). It also questioned
the validity of General Electric's (GE's) estimated probability of
these events as submitted to NRC.
GE proposed that these events be reclassified as accidents because
they are potentially limiting events for critical power ratio (CPR)
margin to the CPR safety limit, particularly for the BWR6 design. GE's
estimated probability of these events was not accepted by the staff,
and they continue to be treated as anticipated operational occurrences
for licensing purposes.
The UCS letter implies that GE may have purposely submitted an
unrealistically low probability value for these events. GE's estimated
probability was based on the fact that since 1981, when SIL-347 (which
gives guidelines for core verification procedures for detection of
misoriented fuel bundles) was first implemented, there had been no
reported cases of plant operation with a misoriented bundle. GE's
assessment was made before the Hope Creek misoriented fuel bundle
event. GE's estimated probability in this specific case (Hope Creek)
was not unreasonable considering reactor performance after SIL-347
implementation and before this event.
M. Potential Safety Hazard Reactor Operation With Failed Fuel Cladding
Petitioner also attached a document from the UCS titled ``Potential
Nuclear Safety Hazard Reactor Operation With Failed Fuel Cladding,''
which concludes that existing design and licensing requirements do not
allow plants to operate with known fuel cladding failures. This
document was also provided to the NRC from the UCS to support a
Petition submitted pursuant to 10 CFR 2.206. A Director's Decision is
being prepared. A copy of that Decision will be forwarded to the
Petitioner when it becomes available.
With regard to plant safety, the Vermont Yankee plant is not
prohibited from operation with a minimal amount of fuel cladding
damage, as stated in the letter of July 6, 1998. The Vermont TS Section
1.1 addresses limits to be observed to prevent significant fuel
cladding damage. Operation is allowed to continue with a minimal amount
of fuel damage, provided that the coolant chemistry requirements of TS
3.6.B are met. These limits are set to values of coolant activity that
ensure that the radiological consequences of postulated design-basis
accidents are within the appropriate dose acceptance criteria.
Petitioner did not submit any information indicating that Vermont
Yankee has operated outside these limits.
N. Event of June 9, 1998
In response to the June 9 event, the NRC performed a special team
inspection to review the causes, safety implications, and licensee
actions associated with the event. The event involved a reactor vessel
high water level turbine trip (due to foreign material in a reactor
feedwater valve) and reactor scram followed by an electrical transient.
The NRC staff concluded that continued operation of Vermont Yankee does
not constitute an undue risk to public health and safety and immediate
action to suspend or modify the operating license is not warranted at
this time. IR 50-271/98-09, dated July 10, 1998, documented the team's
findings.
IV. Conclusion
The NRC staff has evaluated the information provided by the
Petitioner as its basis for the actions requested. As previously
discussed, the information provided by the Petitioner does not warrant
any further action.
The NRC staff has been closely monitoring events at Vermont Yankee
and has taken numerous actions to ensure that there is no undue risk to
public health and safety. The Petitioner did not submit any significant
new information about safety issues. The NRC already knew of the
events, inspection reports, and concerns presented in support of the
Petition. Neither the information presented in the Petition nor any
other information of which the NRC is aware warrants the actions
requested by the Petitioner. Accordingly, the Petitioner's requests for
action are denied.
As provided in 10 CFR 2.206(c) a copy of this Decision will be
filed with the Secretary of the Commission for the Commission's review.
This Decision will constitute 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 7th day of December 1998.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 98-33467 Filed 12-16-98; 8:45 am]
BILLING CODE 7590-01-P