[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
