[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]
-----------------------------------------------------------------------
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
