[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]


=======================================================================
-----------------------------------------------------------------------

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