[Federal Register Volume 61, Number 7 (Wednesday, January 10, 1996)]
[Notices]
[Pages 735-739]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-348]



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NUCLEAR REGULATORY COMMISSION

Confirmatory Order Suspending Authority for and Limiting Power 
Operation and Containment Pressure; (Effective Immediately); and Demand 
for Information

[Docket No. 50-309; License No. DPR-36 EA-96003]
    In the Matter of Maine Yankee Atomic Power Company; Maine Yankee 
Atomic Power Station

I

    Maine Yankee Atomic Power Company (Licensee) is the holder of 
Facility Operating License No. DPR-36, issued by the Atomic Energy 
Commission, predecessor to the Nuclear Regulatory Commission (NRC or 
Commission), pursuant to 10 CFR Part 50 on September 15, 1972. The 
license authorizes the operation of Maine Yankee Atomic Power Station 
(facility or Maine Yankee) in accordance with conditions specified 
therein. The facility is located on the Licensee's site in Lincoln 
County, Maine. The facility has been shut down for refueling and 
repairs to its steam generators since February 6, 1995.

II

    On December 4, 1995, the NRC received both technical allegations 
and allegations of wrongdoing by Yankee Atomic Electric Company (YAEC) 
and the Licensee. In brief, it is alleged that YAEC, acting as agent 
for the Licensee, knowingly performed inadequate analyses of the 
emergency core cooling systems (ECCS) and the containment to support 
two license amendments to increase the rated thermal power at which 
Maine Yankee may operate. It is further alleged that the Licensee 
deliberately misrepresented the analyses to the NRC in seeking the 
license amendments. Specifically, it is alleged that YAEC management 
knew that the 

[[Page 736]]
ECCS for Maine Yankee, if evaluated in accordance with 10 CFR Section 
50.46 using the RELAP5YA code, did not meet the licensing requirements 
for either the 2630 MWt or 2700 MWt power uprates that had previously 
been granted, and that deliberate misrepresentations were made to the 
NRC in order to obtain the 2700 MWt power uprate. (Operation at the 
initially licensed power level of 2440 MWt was not identified as a 
concern.)
    It is also alleged that the Licensee had applied for power uprates 
on the basis of a fraudulent containment analysis. Specifically, the 
facility containment was designed for a pressure of 55 psig, but 
allegedly, YAEC deliberately excluded an energy source (steam 
generators) from the calculations to conceal the possibility that 
containment pressure could increase beyond the design pressure during a 
loss-of-coolant accident (LOCA).
    In response to technical issues raised by these allegations, the 
NRC initiated a special technical review of the safety analysis 
performed by YAEC relating to the Licensee's license amendment 
applications for power uprate. An assessment team of NRC employees was 
dispatched to YAEC Headquarters in Bolton, Massachusetts, on December 
11, 1995. The NRC team was accompanied by two employees of the State of 
Maine, who observed the activities of the team. The team reviewed 
documents and interviewed YAEC employees for 4 days, concentrating 
their efforts in the areas of small-break loss-of-coolant accident 
(SBLOCA) analyses and peak containment pressure determinations. YAEC 
provided additional documents to the NRC after the inspection team 
completed its inspection and departed, but prior to the close of 
business on December 14, 1995. This additional information is related 
to the SBLOCA analysis supporting the Licensee's 15th operating cycle 
(Cycle 15).
    This Order and Demand address requirements and information related 
to future reactor operation. Allegations related to violations of NRC 
requirements, including wrongdoing, will be addressed separately from 
this Order and Demand.

III

    Maine Yankee Atomic Power Company was granted a license to operate 
Maine Yankee on September 15, 1972, at a power level of 2440 MWt, based 
in-part on a Combustion Engineering (CE) analysis of ECCS. By 
application dated August 1, 1977, the Licensee requested a single step 
increase in the maximum thermal power rating to 2630 MWt, again based 
on a CE ECCS analysis. On May 10, 1978, the NRC issued Amendment No. 38 
to the License, which increased the licensed power level to 2630 MWt, 
but restricted operation to 2560 MWt until the Advisory Committee on 
Reactor Safeguards reviewed and recommended approval of the power 
increase from 2560 to 2630 MWt. On June 20, 1978, the Commission issued 
Amendment No. 39, which authorized the Licensee to operate its facility 
at 2630 MWt. On December 28, 1988, the Licensee submitted a request to 
amend its license to increase the plant's maximum thermal power rating 
to 2700 MWt. The Commission granted this amendment request on July 10, 
1989.
    Licensees are required, in accordance with Appendix K to 10 CFR 
Part 50 and 10 CFR Section 50.46, to perform specific accident 
analyses, including SBLOCA analysis, for operation at their licensed 
maximum power level. NUREG-0737, ``Clarification of TMI Action Plan 
Requirements,'' (NUREG-0737) issued following the accident at Three 
Mile Island provides guidance for performing SBLOCA analysis. In 
particular, Item II.K.3.30, ``Revised SBLOCA Methods to Show Compliance 
With 10 CFR Part 50, Appendix K,'' and Item II.K.3.31, ``Plant-Specific 
Calculations to Show Compliance with 10 CFR Section 50.46,'' requested 
licensees submit to the NRC for approval both the revised methods and 
SBLOCA analysis. In response to Item II.K.3.30, the Licensee submitted 
licensing topical report YAEC-1300P, ``RELAP5YA: A Computer Program for 
Light Water Reactor System Thermal-Hydraulic Analysis.''
    By letter dated January 30, 1989, the NRC found that RELAP5YA was 
acceptable, under certain conditions, as a licensing method for use in 
meeting 10 CFR Part 50 Appendix K and NUREG-0737 Item II.K.3.30 for 
SBLOCA analysis for Maine Yankee. Specifically, the NRC's Safety 
Evaluation for RELAP5YA listed twelve conditions, including 
specifications for future plant specific licensing submittals, 
justifying options taken and sensitivity studies performed. Of specific 
interest are conditions 4, 7, 8, 9, and 12, which identified 
justification for model nodalization used when a two-phase mixture 
level dropped below the top of the core, justification of all selected 
options and input data used in plant specific licensing submittals, 
documentation of plant specific sensitivity studies including, but not 
limited to, time step and break sizes, justification of steam generator 
nodalization, and the need to perform a break size study to include the 
worst SBLOCA case for the plant specific licensing application. This 
licensee has not provided the justifications or submittals specified by 
the safety evaluation to support Maine Yankee compliance with II.K.3.31 
and 10 CFR Section 50.46. The NRC review team found that the RELAP5YA 
code as applied for the Maine Yankee Cycle 15 reload included noding 
changes and time step selection which differed from those reviewed by 
NRC in its January 30, 1989 SER for RELAP5YA.
    NUREG-0737 Item II.K.3.5, ``Automatic Trip of Reactor Coolant Pumps 
During Loss-of-Coolant Accident,'' also identified issues related to 10 
CFR Section 50.46. Generic Letter 83-10, ``Resolution of TMI Action 
Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps'' requested 
licensees to justify use of manual action to trip the RCPs for SBLOCA 
events.
    In its reply of June 28, 1985, the licensee concluded that use of a 
sub-cooled margin of 25 deg.F for manually tripping the RCPs satisfied 
the generic letter and 10 CFR Section 50.46. By letter dated April 15, 
1986, the NRC accepted the licensee's position which was based upon 
analyses performed with the RELAP5YA code.
    The containment surrounding the facility's nuclear reactor is 
designed to an internal pressure of 55 psig. The containment was tested 
at 115% (63 psig) of its design pressure for structural acceptance. The 
original licensing basis analysis to predict the peak containment 
pressure, following a postulated loss-of-coolant accident, yielded a 
peak containment pressure of 49.5 psig when an initial containment 
pressure of 0.8 psig was assumed. Because the containment is designed 
to 55 psig, approximately 5 psig margin was available at the time of 
initial licensing. As a result of plant changes (e.g., increase in 
licensed power, and reactor coolant temperature increase) and 
calculational assumptions (e.g., containment volume) the calculated 
peak design-basis accident (DBA) pressure has increased. In the 
December 18, 1995, meeting, the licensee discussed containment 
calculations performed. The licensee stated that, when plant changes 
and calculation assumptions consistent with the as built plant are 
included and the initial containment pressure is limited to 2.0 psig, 
the calculated peak DBA pressure is less than 55 psig, the containment 
design pressure. It is noted that plant Technical Specifications limit 
the maximum operating pressure in containment to 3.0 psig. Assuming an 
initial containment pressure is 3.0 psig, the Technical Specification 
limit, the 

[[Page 737]]
calculated peak design pressure would exceed the containment design 
pressure.
    As required by 10 CFR Part 50, Appendix J, ``Primary Reactor 
Containment Leakage Testing for Water-Cooled Power Reactors,'' the 
Licensee has tested its containment based upon peak DBA pressure, Pa, 
of 50 psig as specified in plant Technical Specifications. The last 
containment leakage test conducted at this pressure was in October 
1988. This value of Pa (i.e., 50 psig) is not consistent with plant 
changes and calculational assumptions reflective of the as built plant 
as discussed above.

IV

    As a result of technical concerns discussed above, questions remain 
as to whether operation of Maine Yankee at a power level of 2700 MWt 
and 3 psig containment pressure meets NRC requirements for ECCS and 
containment design. Thus, this Order and Demand for Information address 
actions necessary to ensure safe operation of the Maine Yankee Nuclear 
Power Plant pending completion of the NRC staff's evaluation of the 
allegations, including the allegations of wrongdoing, and information 
necessary to complete the staff's evaluation.
    Based upon a meeting held with the Licensee on December 18, 1995, 
and the NRC staff's assessment team review, the NRC has determined that 
computer code RELAP5YA, which was proposed for use by Maine Yankee for 
Cycle 15 SBLOCA analyses to demonstrate, in part, compliance with the 
ECCS requirements specified at 10 CFR Section 50.46, has not been 
applied in a manner conforming to the requirements of 10 CFR Part 50, 
Appendix K, ``ECCS Evaluation Model,'' nor has it been applied in a 
manner conforming to the conditions specified in the staff's Safety 
Evaluation dated January 30, 1989 (SE), as necessary for NRC acceptance 
of the use of RELAP5YA for SBLOCA analyses for Maine Yankee. 
Specifically, the Licensee has not demonstrated that the code will 
reliably calculate the peak cladding temperature for all break sizes in 
the small-break LOCA spectrum for Maine Yankee, nor has the Licensee 
submitted the justification for the code options selected, in 
accordance with Condition 7 of the staff's SE, nor has the Licensee 
submitted other justifications and sensitivity studies to satisfy 
Conditions 4, 8, 9, and 12 of the January 30, 1989, SE. Because the 
Licensee did not satisfy the conditions specified in the NRC's 
approval, the plant-specific application of RELAP5YA, is not acceptable 
at Maine Yankee for SBLOCA. Therefore, the SBLOCA portion of the 
emergency core cooling analyses performed by Maine Yankee for Cycle 15 
does not conform with the requirement of 10 CFR Section 50.46. For the 
same reasons, the staff also concludes, that TMI Action Plan Items 
II.K.3.30, II.K.3.31, and II.K.3.5 are likewise not satisfied.
    Accordingly, the staff considers operation of Maine Yankee at 2700 
MWt unacceptable.
    The staff does, however, consider operation of Maine Yankee at 2440 
MWt, using core operating limit parameters based upon analyses 
performed for operation at 2700 MWt acceptable because:
    1. The operating limits in Revision 1 to the Core Operating Limits 
Report (COLR) submitted December 1, 1995, are restricted by non-LOCA 
transient analyses and large-break LOCA analyses which have been 
performed using NRC-approved methods and assuming power levels up to 
2700 MWt. The power level of 2440 MWt is within this range.
    2. The relatively low small-break LOCA peak cladding temperature 
(PCT), explicitly calculated with NRC-approved SBLOCA methods in 
previous cycles at power levels greater than 2440 MWt, met the 
requirements of 10 CFR Section 50.46 with substantial margin (e.g., 
Cycle 4 calculated PCT of 1348 deg. F is substantially less than the 
2200 deg. F required limit at a power level of 2630 MWt). The power 
reduction to 2440 MWt provides additional margin to account for SBLOCA 
modeling uncertainties such as those identified in NUREG-0737.
    3. Review of the analysis performed for other CE and Westinghouse 
plants related to NUREG-0737 Item II.K.3.5 have demonstrated that 
manual tripping of the RCPs meets the requirements of 10 CFR Section 
50.46. Based on the similarity of the initial Maine Yankee plant 
response to a SBLOCA to other CE and Westinghouse plants, the staff 
concludes that the manual tripping of the RCPs is acceptable for Maine 
Yankee.
    Therefore, since operating limits have been developed for power 
levels up to 2700 MWt based upon limiting events that have been 
analyzed using approved methods, and a power reduction margin is being 
imposed to account for SBLOCA modeling uncertainties, the staff finds 
that Maine Yankee operation at 2440 MWt does not pose an undue health 
or safety risk to the public.
    The staff has reviewed the results of containment peak accident 
pressure analysis performed by the Licensee for a licensed thermal 
power level of 2700 MWt, with initial containment pressure limited to 2 
psig. The calculated pressure is 54.8 psig, and is within the 
containment design pressure of 55 psig. The 54.8 psig value was 
generated using sensitivity analysis in conjunction with the original 
licensing basis results. The sensitivity studies were performed by YAEC 
using a CE mass and energy analysis and the CONTEMPT computer program. 
All known, relevant changes to the facility (e.g., spray system 
changes, power uprates, and containment maximum temperature increase) 
were considered, in addition to certain effects not encompassed in the 
original analyses (e.g., reactor coolant system (RCS) thermal 
expansion, use of lower bound containment volume assumption, and 
increased containment operating pressure of 2 psig).
    The staff further notes that there is substantial margin beyond 
containment design pressure. Specifically, containment was successfully 
tested to a pressure of 63 psig upon completion of construction and a 
finite element analysis performed by Sandia Laboratories for the staff 
calculated a lower bound on the ultimate strength of the Maine Yankee 
containment of 96 psig.
    The Licensee recently performed calculations of the leakage 
expected at the maximum containment internal pressure (Pa) for a DBA of 
54.8 psig. Extrapolating from previous Appendix J testing to this 
revised Pa, the Licensee confirmed that the revised leakage was within 
the required acceptance criteria for Type A tests as specified in 10 
CFR Part 50 Appendix J.
    The staff concludes that operation with initial containment 
pressure limited to 2.0 psig and power limited to 2440 MWt does not 
pose an undue health or safety risk to the public.

V

    On Monday, December 18, 1995, a transcribed public meeting was held 
at NRC Headquarters, Rockville, MD, to discuss with the Licensee the 
findings of the review and evaluation team and to seek any additional 
information the Licensee or its agent, YAEC, could provide. In 
concluding the meeting, the NRC advised the Licensee that the NRC had 
concerns regarding the adequacy of proprietary computer code RELAP5YA, 
applied by the Licensee for Cycle 15 SBLOCA analysis, and that this 
analysis is not adequate for demonstrating compliance with 10 CFR 
Section 50.46, ``Acceptance Criteria for Emergency Core Cooling Systems 
for Light Water Nuclear Power Reactors,'' and NUREG-0737, 
``Clarification of TMI Action Plan Requirements,'' Items II.K.3.30 and 
II.K.3.31. This determination led the 

[[Page 738]]
staff to conclude that operation at 2700 MWt was not supported, and 
that the Licensee should evaluate operation at the 2440 MWt level 
established in the original license issued on September 15, 1972. The 
staff indicated that operation at a lower power level could be found 
acceptable if operation is based upon methods previously found 
acceptable by the staff, and not dependent on RELAP5YA for SBLOCA 
analysis. Further, the NRC advised the Licensee that the NRC would 
identify terms and conditions under which the Licensee could propose 
resumption of power operation of its facility.
    On Tuesday, December 19, 1995, the Licensee informed the NRC staff 
that they intended to use RELAP5YA to analyze transients not associated 
with core operating limits. In a December 20, 1995, telephone call the 
NRC advised the Licensee that, based on this broader use of RELAP5YA, 
the NRC would require additional time to determine its further actions. 
In addition, the Licensee committed to not restart the facility until 
NRC had completed its review of new information regarding the use of 
RELAP5YA and containment pressure limits. A letter summarizing events 
of the week of December 18, 1995, was sent to the Licensee on December 
21, 1995.
    By letter dated December 22, 1995, the Licensee committed to: (1) 
limit thermal power output of the plant at or below 2440 MWt until a 
SBLOCA analysis specific to the Maine Yankee plant has been submitted 
to the NRC and written approval from the NRC staff for operation at a 
higher power has been received, (2) develop and document the 
justification for the use of Cycle 15 operating limits using methods 
approved for Maine Yankee without reliance on the RELAP5YA computer 
code prior to achieving initial criticality for Cycle 15 operation, (3) 
limit the maximum internal containment operating pressure to 2 psig 
prior to Cycle 15 initial criticality, and (4) conduct a thorough 
review in order to identify any other applications where RELAP5YA would 
be relied on for Cycle 15 operation.

VI

    I find that implementation of the Licensee's commitments to limit 
power to 2440 MWt and initial containment pressure to 2 psig as set 
forth in the Licensee's letter of December 22, 1995, is acceptable and 
necessary, and that with implementation of these commitments, the 
public health and safety are reasonably assured. In view of the 
foregoing, I have determined that public health and safety require that 
such commitments be confirmed by this Order and Demand. The Licensee 
has agreed to this action. Pursuant to 10 CFR 2.202, I have also 
determined, based on the Licensee's commitment and on the significance 
of the concerns regarding the adequacy of the Licensee's small-break 
LOCA and containment analyses supporting operations described above, 
that the public health and safety require that this Order be 
immediately effective.

VII

    Accordingly, pursuant to sections 103, 161b, 161i, 161o, 182 and 
186 of the Atomic Energy Act of 1954, as amended, and the Commission's 
regulations in 10 CFR 2.202 and 10 CFR Part 50, It is hereby ordered, 
effective immediately, that:
    1. Authority to operate Maine Yankee at 2700 MWt maximum power is 
suspended and Maine Yankee shall limit power to 2440 MWt, until the NRC 
has reviewed and approved the SBLOCA analysis described in Section IX, 
item 5, below.
    2. Authority to operate Maine Yankee at maximum internal 
containment pressure at 3 psig is suspended and Maine Yankee shall 
limit containment pressure to 2 psig, until the NRC has reviewed and 
approved the DBA analysis of containment pressure response required by 
Section IX, item 6, below.
    The Director, Office of Nuclear Reactor Regulation, may relax or 
rescind, in writing, any provisions of this Confirmatory Order upon a 
showing by the Licensee of good cause.

VIII

    Any person adversely affected by this Confirmatory Order, other 
than the Licensee, may request a hearing within 20 days of its 
issuance. Where good cause is shown, consideration will be given to 
extending the time to request a hearing. A request for extension of 
time must be made in writing to the Director, Office of Nuclear Reactor 
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 
and include a statement of good cause for the extension. Any request 
for a hearing shall be submitted to the Secretary, U.S. Nuclear 
Regulatory Commission, ATTN: Chief, Docketing and Service Section, 
Washington, DC 20555. Copies of the hearing request shall also be sent 
to the Director, Office of Nuclear Reactor Regulation, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555, to the Assistant General 
Counsel for Hearings and Enforcement at the same address, to the 
Regional Administrator, NRC Region I, 475 Allendale Road, King of 
Prussia, PA 19406-1415, and to the Licensee. If such a person requests 
a hearing, that person shall set forth with particularity the manner in 
which his/her interest is adversely affected by this Order and shall 
address the criteria set forth in 10 CFR 2.714(d).
    If the hearing is requested by a person whose interest is adversely 
affected, the Commission will issue an Order designating the time and 
place of any hearing. If a hearing is held, the issue to be considered 
at such hearing shall be whether this Confirmatory Order should be 
sustained.
    Pursuant to 10 CFR 2.202(c)(2)(i), any person other than the 
Licensee adversely affected by this Order, may, in addition to 
demanding a hearing, at the time the answer is filed or sooner, move 
the presiding officer to set aside the immediate effectiveness of the 
Order on the ground that the Order, including the need for immediate 
effectiveness, is not based on adequate evidence but on mere suspicion, 
unfounded allegations, or error.
    In the absence of any request for hearing, or written approval of 
an extension of time in which to request a hearing, the provisions 
specified in Section VII above shall be final 20 days from the date of 
this Order without further order or proceedings. If an extension of 
time for requesting a hearing has been approved, the provisions 
specified in Section VII shall be final when the extension expires if a 
hearing request has not been received. An answer or a request for 
hearing shall not stay the immediate effectiveness of this order.

IX

    Additionally, further information is needed to determine whether 
the Commission can continue to have reasonable assurance that the 
Licensee is conducting its activities in accordance with the 
Commission's requirements.
    Accordingly, pursuant to sections 161c, 161o, 182 and 186 of the 
Atomic Energy Act of 1954, as amended, and the Commission's regulations 
in 10 CFR 2.204 and 10 CFR 50.54(f), in order for the Commission to 
determine whether your license should be modified, suspended or 
revoked, or other enforcement action taken to ensure compliance with 
NRC regulatory requirements, you are required to submit to the 
Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, the following information, in writing 
and under oath or affirmation, in the form and according to the 
schedule indicated below:

[[Page 739]]

    1. A description of evaluations that have been completed that 
provide justification for the use of Cycle 15 operating limits, as 
established in the Cycle 15 Core Operating Limits Report, using methods 
approved for Maine Yankee and without reliance on the RELAP5YA computer 
code for SBLOCA analysis and assuming a reactor thermal rating of 2440 
MWt. Details related to analyses performed, significant assumptions, 
and conclusions drawn shall be provided;
    2. A description of all other applications where RELAP5YA is relied 
on for Cycle 15 operation identifying the details of the application, 
and conclusions drawn with respect to any facility modifications or 
procedure changes. For each application, document the determination 
that operability, as defined in Maine Yankee Technical Specifications, 
of affected structures, systems and components is maintained. For plant 
procedures required by Maine Yankee Technical Specifications that rely 
on RELAP5YA analysis for operator action, document the determination as 
to why the affected operator action continues to be appropriate or, if 
necessary, evaluate the affected procedures in accordance with 10 CFR 
Section 50.59 and provide a summary of that evaluation. If any 
procedures are changed, confirm that appropriate training has been 
provided;
    3. A description of measures taken to limit reactor operation to a 
maximum thermal power of 2440 MWt (90.37% of 2700 MWt);
    4. A description of measures taken to limit containment internal 
operating pressure to a maximum of 2 psig;
    5. A SBLOCA analysis that is specific to Maine Yankee for operation 
at power levels up to 2700 MWt. The analysis must meet the requirements 
of 10 CFR Section 50.46, ``Acceptance criteria for emergency core 
cooling systems for light water nuclear power reactors,'' and NUREG-
0737, ``Clarification of TMI Action Plan Requirements,'' Items 
II.K.3.30 and 31, ``SBLOCA Methods'' and ``Plant-specific Analysis,'' 
respectively, and NUREG-0737, Item II.K.3.5, ``Automatic Trip of 
Reactor Coolant Pumps During LOCA;''
    6. An integrated containment analysis, accounting for relevant 
changes to the facility (e.g., spray system changes, power uprates, and 
containment maximum temperature and pressure changes), during a DBA 
that demonstrates the maximum calculated DBA containment pressure meets 
the design basis pressure for Maine Yankee (55 psig). Assumptions used 
for these analyses that are different from those specified in NUREG-
0800, the NRC Standard Review Plan, Section 6.2.1.1.A, shall be 
described.
    Information required by items 1, 2, 3, and 4, above, shall be 
documented and submitted to the NRC prior to criticality. Detailed 
files and supporting computer analyses shall be available on site or at 
the corporate office.
    A schedule for producing the information required by items 5 and 6 
above, shall be provided to the NRC within 30 days of the date of the 
Demand for Information.
    Copies of the response regarding items 1, 2, 3, and 4, and the 
schedule for producing the information required by items 5 and 6, shall 
also be sent to the Assistant General Counsel for Hearings and 
Enforcement at the same address, and to the Regional Administrator, NRC 
Region I, 475 Allendale Road, King of Prussia, PA 19406-1415.
    After reviewing your response, the NRC will determine whether 
further action is necessary to ensure compliance with regulatory 
requirements.

    Dated at Rockville, Maryland, this 3rd day of January 1996.

    For the Nuclear Regulatory Commission.
William T. Russell,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 96-348 Filed 1-9-96; 8:45 am]
BILLING CODE 7590-01-P