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

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