[Federal Register Volume 61, Number 246 (Friday, December 20, 1996)]
[Notices]
[Pages 67352-67354]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-32349]


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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-219]


GPU Nuclear Corporation Oyster Creek Nuclear Generating Station 
Issuance of Director's Decision Under 10 CFR 2.206

    Notice is hereby given that by letters dated May 11 and June 14, 
1996, Mr. deCamp, on behalf of Oyster Creek Nuclear Watch (Petitioner), 
requested NRC, pursuant to 10 CFR 2.206, to investigate and correct a 
highly inaccurate public statement in the ``Neighborhood Update'' (the 
licensee's news magazine) and apparently false public testimony given 
by GPU management at a local zoning board hearing and to take 
appropriate disciplinary action in the matter. Specifically, 
Petitioner's concerns relate to (1) the statement that GPU and the 
Commission agree that a license amendment request that involves the 
movement of spent fuel from the Oyster Creek Nuclear Generating Station 
spent fuel pool to the storage facility while the plant is at power 
``is not a safety issue but a procedural one'' and (2) whether there is 
some special factor at Oyster Creek that would indeed justify Mr. 
Barton's sworn statement that it is unsafe to operate the Oyster Creek 
reactor without full core offload capacity. If no special situation is 
found that prevents Oyster Creek from operating without full offload 
capacity, Petitioner requests that the Commission take appropriate 
disciplinary action against GPU Nuclear management for making a false 
statement under oath.
    As a basis for the request regarding the first concern that the 
statement in the ``Neighborhood Update'' is untrue, Petitioner 
referenced the following excerpts from NRC Bulletin 96-02 (NRCB 96-02) 
of April 11, 1996:
    The NRC staff audited both the initial and updated 10 CFR 50.59 
evaluations performed by the licensee [GPU Nuclear] and determined 
that the proposed cask movement activities represent an unreviewed 
safety question that should be submitted to the NRC for review and 
approval pursuant to the requirements of 10 CFR 50.59 and 50.90. * * 
* Accordingly, as defined in 10 CFR 50.59(c), if an activity is 
found to involve an unreviewed safety question, an application for a 
license amendment must be filed with the Commission pursuant to 10 
CFR 50.90.

    As a basis for the Petitioner's other concerns, the Petitioner sets 
forth the relevant excerpts from Mr. Barton's testimony of March 7, 
1994, and states that ``the NRC ruled in February 1985 in 10 CFR Part 
53 that reactors may safely be run without full core offload 
capacity.''
    Notice is hereby given that by a Director's Decision (DD 96-22) 
dated December 11, 1996, the Acting Director, Office of Nuclear Reactor 
Regulation, has denied the Petitions. The staff concluded that the 
issues raised by the Petitioner are without merit and that there is no 
basis to take disciplinary action against GPU, as explained in the 
``Director's Decision Pursuant to 10 CFR 2.206'' (DD 96-22), the 
complete text of which follows this notice and is available for 
inspection at the Commission's Public Document Room at 2120 L Street, 
NW, Washington DC, and at the local public document room located at 
Ocean County Library, Reference Department, 101 Washington Street, 
Tom's River, NJ.

    Dated at Rockville, Maryland, this 11th day of December 1996.

    For The Nuclear Regulatory Commission
Frank J. Miraglia,
Acting Director, Office of Nuclear Reactor Regulation.

Director's Decision Under 10 CFR 2.206

I. Introduction

    By letters dated May 11 and June 14, 1996, Mr. William deCamp, Jr., 
requested on behalf of Oyster Creek Nuclear Watch (the Petitioner) that 
the U.S. Nuclear Regulatory Commission (NRC or Commission) take action 
to investigate statements made by GPU Nuclear Corporation (GPU) in the 
April 1996 publication ``Neighborhood Update'' (the licensee's news 
magazine) and during sworn testimony on March 7, 1996, before the Lacey 
Township Zoning Board of Adjustment (the Zoning Board). The Petitioner 
asserts that the statements are false. The Petitioner further requests 
that NRC take appropriate disciplinary action against GPU management. 
The Petitioner's requests are being treated as Petitions pursuant to 
Section 2.206 of Title 10 of the Code of Federal Regulations (10 CFR 
2.206).
    The specific statements of concerns are (1) the statement in the 
``Neighborhood Update'' that GPU and the Commission agree that a 
license amendment request that involves the movement of spent fuel from 
the Oyster

[[Page 67353]]

Creek Nuclear Generating Station spent fuel pool to the storage 
facility while the plant is at power ``is not a safety issue but a 
procedural one'' and (2) a sworn statement by Mr. Barton, who was the 
Director of the Oyster Creek Nuclear Generating Station, before the 
Zoning Board that it is unsafe to operate the Oyster Creek reactor 
without full core offload capacity. The Petitioner, furthermore, 
requests that if no special situation is found that prevents Oyster 
Creek from operating without full offload capacity, the Commission take 
appropriate disciplinary action against GPU management for making a 
false statement under oath.1
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    \1\ The petitioner is not asserting that the licensee has 
provided false information to the Nuclear Regulatory Commission. A 
licensee's obligation to ensure the completeness and accuracy of its 
communications with the Commission is set forth in 10 CFR 50.9(a). 
This regulation requires, in part, that ``[i]nformation provided to 
the Commission by an applicant for a license or by a licensee or 
information required by statute or by the Commission's regulations, 
orders, or license conditions to be maintained by the applicant or 
the licensee shall be complete and accurate in all material 
respects.''
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    For the reasons stated below, I am denying the relief requested by 
the Petitioner.

II. Discussion

A. GPU Statement That the Movement of the Fuel Raises a Procedural 
Issue, Not a Safety Issue

    As a basis for the request regarding the first concern that the 
statement in the ``Neighborhood Update'' is untrue, Petitioner 
referenced the following excerpts from NRC Bulletin 96-02 (NRCB 96-02), 
``Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor 
Core, or Over Safety-Related Equipment,'' of April 11, 1996:

    The NRC staff audited both the initial and updated 10 CFR 50.59 
evaluations performed by the licensee [GPU Nuclear] and determined 
that the proposed cask movement activities represent an unreviewed 
safety question that should be submitted to the NRC for review and 
approval pursuant to the requirements of 10 CFR 50.59 and 50.90 * * 
*. Accordingly, as defined in 10 C.F.R. 50.59(c), if an activity is 
found to involve an unreviewed safety question, an application for a 
license amendment must be filed with the Commission pursuant to 10 
CFR 50.90.

    GPU met with the NRC staff on November 19, 1993, to discuss plans 
for using the reactor building crane to move spent fuel out of the 
spent fuel pool in a transfer cask for transportation to the dry cask 
storage facility during power operations at Oyster Creek. During the 
discussions, the NRC staff raised concerns regarding the use of the 
crane and its ability to meet the heavy load criteria of NUREG-0612, 
``Control of Heavy Loads at Nuclear Power Plants.'' GPU indicated that 
this special application of the crane would be evaluated pursuant to 10 
CFR 50.59.2 NRC stated that it would conduct an audit of the 50.59 
evaluation.
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    \2\ 10 CFR 50.59 provides, in part, that a licensee may make 
changes in the facility or procedures as defined in the safety 
analysis report without prior Commission approval unless the 
proposed change involves a change in the technical specifications or 
an unreviewed safety question. The regulation, furthermore, requires 
the licensee to prepare and maintain a written safety evaluation 
addressing the issue of whether the proposal involves an unreviewed 
safety question. A proposal is deemed to involve an unreviewed 
safety question if (1) it involves an increase in the probability or 
consequences of an accident previously evaluated; or (2) creates the 
possibility of a new or different kind of accident from any accident 
previously evaluated; or (3) involves a reduction in a margin of 
safety as defined in the basis for any technical specification.
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    In April 1995, GPU informed NRC that the 50.59 evaluation for use 
of the crane to move the transfer cask was complete. On May 2 and 3, 
June 12, and October 12 and 13, 1995, the NRC staff conducted onsite 
audits and met with GPU at Oyster Creek regarding the use of the crane. 
On November 2, 1995, in a telephone call between the NRC staff and Mr. 
Keaten, Vice President and Director, Technical Functions, GPU, the NRC 
staff advised GPU that the staff's concerns regarding the use of the 
non-single-failure-proof crane to move the 100-ton transfer cask while 
the plant was at power had not been resolved by its 50.59 evaluation. 
Specifically, the staff was concerned that the activity involved the 
movement of loads heavier than previously considered in the final 
safety analysis report (FSAR) and, therefore, might reduce the margin 
of safety, and that a load drop in the reactor building might result in 
consequences greater than previously evaluated in the FSAR and, 
therefore, may pose an unreviewed safety question.
    Consequently, Mr. Keaten advised the staff that GPU was considering 
a plant modification, including reactor building crane upgrades, that 
would address the staff's concerns.
    The NRC staff inspected the licensee's updated 10 CFR 50.59 
evaluation which considered the reactor building crane upgrades. The 
NRC staff's inspections included sending a team to Oyster Creek. The 
staff concluded that its safety concerns had been addressed and 
resolved. The NRC staff also determined that the licensee's planned 
movement of spent fuel to the dry storage facility during plant 
operation was safe and in accordance with all license requirements. 
Notwithstanding the technical acceptability of the licensee's 
methodology and analysis in the updated 10 CFR 50.59 evaluation, NRC 
staff determined that since the possibility of an unreviewed safety 
question (USQ) had been involved before the licensee made modifications 
to upgrade the reactor building crane, GPU must submit a license 
amendment application for the proposed cask movement activities. At the 
public meeting on February 29, 1996, GPU was informed by the NRC staff 
that an amendment was required. When the NRC receives an amendment 
application, it is required to follow specific procedures set forth in 
10 CFR 50.91.3
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    \3\  10 CFR 50.91 requires the Commission to use specified 
procedures when it receives an application requesting an amendment 
to an operating license including procedures that concern consulting 
the State in which the facility is located and procedures concerning 
providing notification to the public of the licensee's amendment, 
the Commission's findings or determinations regarding the amendment, 
and opportunity for a hearing.
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    Accordingly, the staff finds, after its review and evaluation of 
the licensee's proposed action, that there are no safety issues 
preventing the adoption of the proposal, but procedures require 
amendment approval before the proposal can be implemented.

B. GPU Statement Concerning Safe Operation and Full Core Discharge 
Capability

    As basis for the Petitioner's request concerning GPU statements 
about safety and full core discharge capability, the Petitioner sets 
forth excerpts from Mr. Barton's testimony of March 7, 1994, before the 
Zoning Board, and states that ``the NRC ruled in February 1985 in 10 
CFR Part 53 that reactors may safely be run without full core offload 
capacity.'' 4
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    \4\ The Commission has stated that a full core reserve 
capability is not an NRC safety requirement. 50 FR 5548, 5549 (1985)
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    The Petitioner quoted in a letter and enclosed, underlined in red, 
copied portions of Mr. Barton's testimony as follows:

    If we do not install the dry spent fuel storage modules by 1996, 
the plant would not have the capacity of totally off-loading fuel 
from the reactor to the in-plant spent fuel pools. (transcript pp. 
94-95)
    In order to operate safely we should be able to remove this fuel 
from the reactor and store it in the spent storage pool * * * 
(transcript p. 95)
    Without dry storage and without the ability to remove this fuel 
from the reactor, the plant would not be able to operate. 
(transcript p. 95)

    Mr. Barton's full testimony in context with the Petitioner's 
extracted quotes is as follows:


[[Page 67354]]


    The fall of 1996 is a critical time for plant operations. If we 
do not install the dry spent fuel storage modules by 1996, the plant 
would not have the capability of totally off-loading fuel from the 
reactor to the in-plant spent fuel pool. This is not a desirable 
operating configuration, should the plant need to conduct internal 
inspections of the reactor vessel that would require fuel to be 
removed from the reactor. In order to operate safely we should be 
able to remove this fuel from the reactor and store it in the spent 
fuel storage pool inside the plant, and after 1996 we will not have 
the flexibility to do that. Without dry storage and without the 
ability to remove all the fuel from the reactor, the plant would not 
be able to operate. (transcript p. 95)

    Taken in context, it appears that what Mr. Barton is stating is 
that he is concerned with operations management due to the inability to 
have full core off-load capability and that having full core off-load 
capability can in certain situations enhance safety. The plant has the 
capacity to complete one more refueling operation before they will not 
be able to operate without dry storage capability as Mr. Barton stated. 
The Commission has stated a similar view with regard to the issue of 
maintaining full core reserve storage capability:

    While a full core reserve capability is not an NRC licensing or 
safety requirement, maintenance of full core reserve would enhance 
safety to some extent, and would also be needed to prevent extended 
reactor outages in the event a core must be discharged in order to 
inspect the reactor pressure vessel and perform other routine and 
unscheduled maintenance operations.5
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    \5\  The NRC's Statements of Consideration concerning the 
amendment of 10 CFR Parts 1 and 53 entitled, ``Criteria and 
Procedures for Determining the Adequacy of Available Spent Nuclear 
Fuel Storage Capacity,'' 50 FR 5548, 5549 (1985)

    The December 6, 1993, Zoning Board hearing testimony of Mr. Gordon 
Bond, Director of Nuclear Analysis and Fuel for GPU Nuclear, also 
supports the view that the concern is with operations management. When 
asked whether it is important to maintain full core discharge 
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capability, Mr. Bond responded as follows:

    We believe it is. It's not required by Federal Regulations, but 
we believe it's prudent to allow sufficient reserve capacity in our 
pool to be able to offload the core any time that we may have to. 
For example, you may want to do some inspections inside the vessel, 
and to do that you'll need to remove all of the fuel. (transcript p. 
32)

Accordingly, the staff finds that the statements and remarks of Mr. 
Barton in their context are not false or misleading.

V. Conclusion

    The NRC staff has reviewed the statements made by GPU in the April 
1996 ``Neighborhood Update'' (the licensee's news magazine) and the 
testimony of GPU managers before a local Zoning Board and concluded 
that the assertions raised by the Petitioner are without merit and that 
there is no basis to take any action against GPU. Accordingly, the 
Petitioner's requests are denied.
    A copy of this Director's Decision will be filed with the Secretary 
of the Commission for the Commission to review as stated in 10 CFR 
2.206(c). This Decision will become the final action of the Commission 
25 days after issuance unless the Commission, on its own motion, 
institutes a review of the Decision within that time.

    Dated at Rockville, Maryland, this 11th day of December 1996.

    For the Nuclear Regulatory Commission
Frank J. Miraglia,
Acting Director, Office of Nuclear Reactor Regulation.
[FR Doc. 96-32349 Filed 12-19-96; 8:45 am]
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