[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