[Federal Register Volume 62, Number 38 (Wednesday, February 26, 1997)]
[Notices]
[Pages 8780-8783]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-4701]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-317 and 50-318]
Baltimore Gas and Electric Company; Notice of Consideration of
Issuance of Amendment to Facility Operating License, Proposed No
Significant Hazards Consideration Determination, and Opportunity for a
Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of an amendment to Facility Operating License Nos.
DPR-53 and DPR-69 issued to Baltimore Gas and Electric Company, for
operation of the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2,
located in Calvert County, Maryland.
The proposed amendment revises the Technical Specifications (TSs)
to reduce the minimum Reactor Coolant System (RCS) total flow rate from
370,000 gpm to 340,000 gpm; reduce the Reactor Protective
Instrumentation trip setpoint for Reactor Coolant Flow--Low from
greater than or equal to 95% to greater than or equal to 92% of design
reactor coolant flow; adjust the reactor core thermal margin safety
limit lines to reflect the reduced RCS flow rate; and reduce the lift
setting range for the eight Main Steam Safety Valves (MSSVs) with the
highest allowable lift setting from the current range of 935 to 1065
psig to a more restrictive range of 935 to 1050 psig. In addition to
the changes to the TSs necessary to support an increased number of
plugged SG tubes, reanalysis of the accident analyses affected by this
change identified an Unreviewed Safety Question (USQ) associated with
these changes. The USQ results from the determination that the Main
Steam Line Break (MSLB) and Seized Rotor Event analyses involve an
increased percentage of failed fuel cladding. Finally, three reanalyzed
events (MSLB, Loss of Coolant Flow, and Boron Dilution) will require
Nuclear Regulatory Commission (NRC) approval due to changes to the
methodology or assumptions used to analyze these events.
Before issuance of the proposed license amendment, the Commission
will have made findings required by the Atomic Energy Act of 1954, as
amended (the Act) and the Commission's regulations.
The Commission has made a proposed determination that the amendment
request involves no significant hazards consideration. Under the
Commission's regulations in 10 CFR 50.92, this means that operation of
the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. As
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration, which is presented
below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed amendment defines changes to the operating licenses
for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, necessary to
support increased steam generator tube plugging. The effects of
increased steam generator tube plugging include reduced steam
generator pressure and RCS flow rate, and increased core outlet (hot
leg) temperature. The Technical Specification changes necessary to
account for these effects are reducing the minimum RCS total flow
rate from 370,000 gpm to 340,000 gpm; reducing the Limiting Safety
System Setting for reactor coolant flow trip function from greater
than or equal to 95% to greater than or equal to 92% of design
reactor coolant flow; revising the Reactor Core Thermal Safety Limit
lines to indicate operation at the lower reactor coolant flow rate;
and decreasing the maximum allowable lift settings for the eight
highest set Main
[[Page 8781]]
Steam Safety Valves from 1065 psig to 1050 psig. The Design Basis
Events (DBEs) affected by these changes were reanalyzed to determine
if the effects of increased steam generator tube plugging, and the
associated changes to the Technical Specifications, could result in
exceeding the acceptance criteria applicable to each of these
events. Although it was determined that the DBE acceptance criteria
would not be exceeded as a result of increased steam generator tube
plugging, the analyses for the Main Steam Line Break and Seized
Rotor Events indicated an increased percentage of fuel cladding
failure as a result of the lower RCS total flow rate; therefore, it
was determined that this activity involves a USQ.
Technical Specification 2.1.1 will be changed to establish more
restrictive limits on core thermal power and reflect a lower minimum
RCS flow of 340,000 gpm. Making the core thermal power limits more
restrictive does not initiate a change to plant conditions that
would affect other plant components. Therefore, the probability of a
previously evaluated accident is not significantly increased.
Additionally, the Limiting Conditions for Operation and Limiting
Safety System Settings based on these limits remain adequately
conservative or will be changed in the Core Operating Limits Report,
as appropriate. Therefore, the consequences of a previously
evaluated accident are not significantly increased.
Technical Specification 2.2 will be changed to reduce the
Reactor Coolant Flow--Low reactor trip setpoint from [greater than
or equal to] 95% to [greater than or equal to] 92%, thereby
providing additional operating margin to this trip setpoint and the
associated pre-trip alarm. Reducing this setpoint does not initiate
a change to plant conditions that would affect other plant
components. Therefore, the probability of a previously evaluated
accident is not significantly increased.
As demonstrated by the revised Loss of Coolant Flow analysis,
the proposed Reactor Coolant Flow--Low reactor trip setpoint will
continue to provide adequate core protection. A trip setpoint of
[greater than or equal to] 92% ensures fuel is not damaged, and the
site boundary dose remains a small fraction of the 10 CFR Part 100
guidelines. Therefore, the consequences of a previously evaluated
accident are not significantly increased.
Technical Specification 3.2.5.c will be changed to reduce the
minimum RCS total flow rate from 370,000 gpm to 340,000 gpm. This
change reduces the core heat removal rate and slightly increases the
core outlet and average coolant temperatures. This change involves a
USQ, as the Main Steam Line Break and Seized Rotor Event analyses
have indicated an increase in the number of failed fuel pins during
these events as a result of reducing the initial RCS flow rate. The
probability of malfunction of equipment important to safety (i.e.,
fuel pin cladding) during these accidents increases. However, this
malfunction is not an accident initiator. Rather, it is a
consequence of an accident. Therefore, the probability of a
previously evaluated accident is not significantly increased. The
consequences of the Main Steam Line Break and Seized Rotor Events
are not significantly increased, as the results of the analyses of
these events are within the current acceptance criteria established
by the NRC.
Analyses and evaluations have been performed to demonstrate that
the new flow and temperature conditions are acceptable:
Fuel and core performance remain within acceptable limits.
Analysis and evaluation of fuel mechanical design, core physics,
parameters, fuel pin performance, fuel assembly thermal/hydraulic
performance, and fuel pin corrosion all demonstrate acceptable
results.
The effect of the slightly elevated core outlet and average
coolant temperature on the structural integrity of the RCS is
acceptable. The RCS penetration inspection program and the steam
generator tube inspection program will continue to identify and
repair or isolate Alloy 600 cracks prior to inservice failure of
these components. The stress analysis for the reactor vessel and
piping remain bounding.
The performance of control systems (i.e., feedwater, pressurizer
level, and pressurizer pressure) will maintain RCS and steam
generator parameters within appropriate limits by periodic
adjustment, as necessary. Reactor coolant pump operation will be
maintained within acceptable limits by periodic adjustment of the
operating curves.
Therefore, the probability of a previously evaluated accident is
not significantly increased.
Analyses and evaluations of the DBEs have been performed
demonstrating that the NRC acceptance criteria for these events are
met. The revised analyses and evaluations consider reduced RCS flow,
increased reactor coolant temperature, and increased steam generator
tube plugging conditions.
The results of analyses and evaluations of the Postulated
Accidents demonstrate that the site boundary dose is within 10 CFR
Part 100 guidelines and the core geometry remains coolable. Loss-of-
Coolant Accident analysis results meet the acceptance criteria
stipulated in 10 CFR 50.46(b).
The results of analyses and evaluations of Anticipated
Operational Occurrences demonstrate that fuel parameters do not
exceed the specified acceptable fuel design limits and site boundary
dose is a small fraction of 10 CFR Part 100 guidelines. Primary and
secondary system pressure remain below the pressure upset limits for
the RCS and steam generators, respectively.
Therefore, the consequences of a previously evaluated accident are
not significantly increased.
Technical Specification 4.7.1.1. will be changed to reduce the
maximum allowable lift setting for the eight Main Steam Safety
Valves with the highest lift setpoint. This change will place more
restrictive limits on the allowable range of lift settings for these
eight valves. The allowable range of lift settings for the proposed
change is also allowed by current Technical Specification.
Therefore, the probability of a previously evaluated accident
occurring is not significantly increased.
The revised safety analyses will credit the highest lift setting
for these eight valves as being 1050 psig. The more restrictive
limit on the maximum lift setting is required in order to make this
Technical Specification consistent with the revised safety analyses.
Analyses performed assuming the proposed maximum lift setting for
these valves demonstrates that secondary system pressure does not
exceed 110% of the system design pressure. Therefore, the
consequences of a previously evaluated accident are not
significantly increased.
Therefore, operation of the facility in accordance with this
amendment does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed amendment revises limiting parameters to assure
safe operation commensurate with the effects of steam generator tube
plugging, and will not change the modes of operation defined in the
facility license. The analysis of transients associated with steam
generator malfunctions are part of the design and licensing bases.
This change does not add any new equipment, modify any interfaces
with any existing equipment, or change the equipments's function, or
the method of operating the equipment. The proposed change does not
change plant conditions in a manner which could affect other plant
components. Reactor core, RCS, and steam generator parameters remain
within appropriate design limits during normal operation.
Therefore, the proposed change could not cause any existing
equipment to become an accident initiator.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The margins of safety associated with this change are defined in
the fuel and core-related analyses, the Alloy 600 stress corrosion
cracking evaluation, the RCS structural evaluation, the operational
evaluation, and in each of the transient and accident analyses
affected by the increased steam generator tube plugging.
Reanalysis of the fuel and core-related analyses for fuel
mechanical design, core physics, fuel performance, thermal
hydraulics, and fuel rod corrosion verified that the fuel and core
performance will remain within acceptable limits and will be bounded
by the current assumptions for fuel performance in the transient and
accident analyses. The Alloy 600 RCS penetration inspection program
and the steam generator tube inspection program will continue to
find and repair Alloy 600 cracks at the slightly elevated core exit
temperature prior to any postulated inservice failure of these
components. The stress analyses performed for the reactor vessel and
piping remain bounding for the slightly elevated core exit
[[Page 8782]]
temperature. Additionally, the performance of non-safety-related
control systems remains adequate to maintain RCS and steam generator
parameters within appropriate operating limits. Therefore, the
margins of safety associated with the physical and operational
effects of this change will not be significantly reduced.
An evaluation of the affected DBEs confirmed that the
established acceptance criteria for specified acceptable fuel design
limits, primary and secondary system over-pressurization, 10 CFR
50.46(b), Acceptance Criteria for Emergency Core Cooling Systems for
Light-Water Nuclear Power Reactors, and potential radiation dose
during accidents have been completed in support of this license
amendment request. The evaluation concludes that, when considering
the proposed Limiting Safety System Setting for the Reactor Coolant
Flow--Low trip, Limiting Conditions for Operation for RCS total flow
rate, and reduced lift settings for eight Main Steam Safety Valves
per unit, all applicable acceptance limits are met. Furthermore, the
USQ resulting from the reduced RCS total flow rate does not
represent a reduction in the margin of safety, as the site boundary
dose calculated in the affected DBE analyses is within the current
established radiation dose limits and the core geometry remains
coolable. Therefore, the margins of safety associated with the
transient and accident analyses affected by this change will not be
significantly reduced.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received.
Should the Commission take this action, it will publish in the Federal
Register a notice of issuance and provide for opportunity for a hearing
after issuance. The Commission expects that the need to take this
action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to
4:15 p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By March 28, 1997 the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and at the local public
document room located at the Calvert County Library, Prince Frederick,
Maryland 20678. If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of hearing or an
appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
[[Page 8783]]
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to S. Singh Bajwa: petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, and to Jay E.
Silbert, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street,
NW., Washington, DC, 20037 attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for hearing will not
be entertained absent a determination by the Commission, the presiding
officer or the presiding Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment dated January 31, 1997, as supplemented
February 13, 1997, which is available for public inspection at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC, and at the local public document room located at
the Calvert County Library, Prince Frederick, Maryland 20678.
Dated at Rockville, Maryland, this 20th day of February 1997.
For the Nuclear Regulatory Commission.
Alexander W. Dromerick,
Senior Project Manager, Project Directorate I-1, Division of Reactor
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 97-4701 Filed 2-25-97; 8:45 am]
BILLING CODE 7590-01-P
