[Federal Register Volume 59, Number 60 (Tuesday, March 29, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-7326]


[[Page Unknown]]

[Federal Register: March 29, 1994]


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NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-454, STN 50-455, STN 50-456, and STN 50-457]

 

Commonwealth Edison 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 amendments to Facility Operating License Nos. 
NPF-37, NPF-66, NPF-72, and NPF-77, issued to Commonwealth Edison 
Company (the licensee), for operation of the Byron Station, Units 1 and 
2, and Braidwood Station, Units 1 and 2 located in Ogle County, 
Illinois and Will County, Illinois respectively.
    The proposed amendments would add a one-time revision to Technical 
Specification (TS) 4.7.1.2 and Table 3.7-2 to permit continued 
activities at all four units with main stream Code safety valve 
tolerances of 3 percent until the setpoints can be reset to 
within 1 percent. Specifically, the licensee proposed 
making two changes to the TSs. One change would be the addition of a 
statement to TS 4.7.1.1 for Braidwood stating that the provisions of TS 
4.0.4 are not applicable for Braidwood, Unit 1 Cycle 5 until initial 
entry into MODE 2. Braidwood, Unit 1, is currently in refueling. The 
second change would be a statement to Table 3.7-2 allowing main steam 
line Code safety valve lift settings to have a 3 percent 
tolerance until May 9, 1994, by which time the lift settings will be 
reset to 1 percent.
    The exigent circumstances could not be avoided because the licensee 
only recently became aware of the possible out-of-tolerance setpoints 
through a letter from the vendor dated March 10, 1994. Upon discovery, 
the licensee requested a Notice of Enforcement Discretion by telephone 
followed by a formal request on March 11, 1994. Upon review, the NRC 
concluded that exercising enforcement discretion involved minimal or no 
safety impact and exercised discretion not to enforce compliance with 
the TS for the period from March 10, 1994, until the approval of the TS 
amendment request which was to be submitted for NRC review by March 21, 
1994.
    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.
    Pursuant to 10 CFR 50.91(a)(6) for amendments to be granted under 
exigent circumstances, the NRC staff must determine 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:

    a. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    In the analysis performed for a 3% as-found MSSV 
[main steam safety valve] setpoint, all of the applicable Loss of 
Coolant Accident (LOCA) and non-LOCA design basis acceptance 
criteria remain valid both for the transients evaluated and the 
single event analyzed, Loss of External Load/Turbine Trip.
    The MSSVs are actuated after accident initiation to protect the 
secondary systems from overpressurization. Increasing the as-found 
setpoint tolerance will not result in any hardware modification to 
the MSSVs. Therefore, there is not an increase in the likelihood of 
spurious opening of a MSSV. Sufficient margin exists between the 
normal steam system operating pressure and the valve setpoint with 
the increased tolerance to preclude an increase in the probability 
of actuating the valves.
    The peak primary and secondary pressures remain below 110% of 
design at all times. The Departure from Nucleate Boiling Ratio 
(DNBR) and Peak Clad Temperature (PCT) values remain within the 
specified limits of the licensing basis. Although increasing the 
valve setpoint tolerance may increase the steam release from the 
ruptured steam generator above the UFSAR [Updated Final Safety 
Analysis Report] value by approximately 2%, the Steam Generator Tube 
Rupture (SGTR) analysis indicates that the calculated break flow is 
still less than the value reported in the UFSAR. Therefore, the 
radiological analysis indicates that the slight increase in the 
steam release is offset by the decrease in the break flow such that 
the offsite radiation doses are less than those reported in the 
UFSAR. The evaluation also concluded that the existing mass releases 
used in the offsite dose calculation for the remaining transients 
(i.e., steamline break, rod ejection) are still applicable. 
Therefore, based on the above, there is no increase in the dose 
releases.
    The effects of increased tolerances for MSSV setpoints on the 
LOCA safety analyses has been previously performed for VANTAGE 5 
fuel. Calculations performed to determine the response to a 
hypothetical large break LOCA do not model the MSSVs, since a large 
break LOCA is characterized by a rapid depressurization of the 
reactor coolant system below the pressure of the steam generators. 
Thus, the calculated consequences of a large break LOCA are not 
dependent upon assumptions of MSSV performance. Therefore, the large 
break LOCA analysis results are not adversely affected by revising 
setpoint tolerances.
    The small break LOCA analyses presented in Appendix C of the 
Byron/Braidwood Stations Units 1 and 2 VANTAGE 5 Reload Transition 
Safety Report were performed using a 3% higher safety valve setpoint 
pressure. The standard 3% accumulation between valve actuation and 
full flow was also accounted for in the analyses. These analyses 
calculated peak cladding temperatures well below the allowed 
2200 deg. F limit as specified in 10 CFR 50.46 demonstrating that 
the change to the MSSV setpoint tolerance can be accommodated for 
small break LOCAs.
    Neither the mass and energy release to the containment following 
a postulated LOCA, nor the containment response following the LOCA 
analysis, credit the MSSV in mitigating the consequences of an 
accident. Therefore, changing the MSSV lift setpoint tolerances 
would have no impact on the containment integrity analysis. In 
addition, based on the conclusion of the transient analysis, the 
change to the MSSV tolerance will not affect the calculated 
steamline break mass and energy releases inside containment.
    The loss of load/turbine trip event was analyzed in order to 
quantify the impact of the setpoint tolerance relaxation. As was 
demonstrated in the evaluation, all applicable acceptance criteria 
for this event have been satisfied and the conclusions presented in 
the UFSAR remain valid. The conclusions presented in the 
Overpressure Protection Report remain valid. Therefore, the 
probability or consequences of an accident previously evaluated in 
the UFSAR would not be increased as a result of increasing the MSSV 
lift setpoint as found tolerance to 3% above or below the current 
Technical Specification lift setpoint value.
    The probability of an accident occurring will not be affected by 
granting this amendment request. Therefore, the requested amendment 
does not significantly increase the probability or consequences of 
an accident previously evaluated.
    b. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No new system configurations are introduced, and no equipment is 
being operated in a new or different manner than has been previously 
analyzed. Accordingly, no new or different failure modes are being 
created. Increasing the as-left setpoint tolerance on the MSSV does 
not create the possibility of an accident which is different than 
any already evaluated in the UFSAR. Increasing the as-left lift 
setpoint tolerance on the MSSVs does not introduce a new accident 
initiator mechanism. No new failure modes have been defined for any 
system or component important to safety nor has any new limiting 
single failure been identified. No accident will be created that 
will increase the challenge to the MSSVs, and result in increased 
actuation of the valves. Therefore, the possibility of an accident 
different than any already evaluated is not created.
    c. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Although the proposed amendment is requested for equipment 
utilized to prevent overpressurization on the secondary side and to 
provide an additional heat removal path, increasing the as-left lift 
setpoint tolerance on the MSSVs will not adversely affect the 
operation of the reactor protection system, any of the protection 
setpoints or any other device required for accident mitigation.
    The proposed increase in the as-left MSSV lift setpoint 
tolerance will not invalidate the LOCA and non-LOCA conclusions 
presented in the UFSAR accident analyses. The new loss of load/
turbine trip analysis concluded that all applicable acceptance 
criteria are still satisfied. For all the UFSAR non-LOCA transients, 
the DNB [departure from nucleate boiling] design basis, primary and 
secondary pressure limits and does release limits continue to be 
met. Peak cladding temperatures remain well below the limits 
specified in 10 CFR 50.46. Thus, there is no reduction in the 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 15 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 15-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 15-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. The Commission expects that the need to 
take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to room P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, 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 
20555.
    The filing of requests for hearing and petitions for leave to 
intervene is discussed below.
    By April 13, 1994, 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 20555 and at the local 
public document rooms, which for Byron is located at the Byron Public 
Library, 109 N. Franklin, P.O. 434, Byron, Illinois 61010; and for 
Braidwood is located at the Wilmington Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481. 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, designed 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 a 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 the amendment is issued before the expiration of the 30-day 
hearing period, the Commission will make a final determination on the 
issue of no significant hazards consideration. If a hearing is 
requested, the final determination will serve to decide when the 
hearing is held.
    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, 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 20555, 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 Mr. James E. Dyer: 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, and to Michael I. 
Miller, Esquire; Sidney and Austin, One First National Plaza, Chicago, 
Illinois 60690, 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 March 11, 1994, which is available for 
public inspection at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC 20555, and at the local 
public document rooms, which for Byron is located at the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; and for 
Braidwood is located at the Wilmington Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

    Dated at Rockville, Maryland, this 23rd day of March 1994.

    For the Nuclear Regulatory Commission.
George F. Dick, Jr.,
Project Manager, Project Directorate III-2; Division of Reactor 
Projects--III/IV/V, Office of Nuclear Reactor Regulation.
[FR Doc. 94-7326 Filed 3-28-94; 8:45 am]
BILLING CODE 7590-01-M