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 Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2

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

Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2

The U.S. Nuclear Regulatory Commission (the Commission) is

considering issuance of an amendment 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, located in Ogle County, Illinois and Braidwood Station, Units 1 and

2, located in Will County, Illinois.

The proposed amendment would revise technical specification (TS)

1.0, ``Definitions'', TS 3/4.6.1, ``Primary Containment'' and

associated Bases; and TS 5.4.2, ``Reactor Coolant System Volume'' for

Byron and Braidwood to support the steam generator replacement for Unit

1 at each site. The replacement steam generators increase the reactor

coolant system volume which results in a higher calculated peak

containment pressure (Pa) value. The staff's proposed no significant

hazards consideration determination for the requested change was

published on April 23, 1997 (62 FR 19826).

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. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Each of the RSGs has a larger RCS primary side volume than the

original steam generators (OSGs). As a result of the RCS volume

increase, the mass and energy release during the blowdown phase of

the large break loss of coolant accident (LBLOCA) is increased.

Additionally, the heat transfer rate of the RSGs is greater than the

OSGs, and the RSGs will operate at a slightly higher pressure than

that for the OSGs. Consequently, the steam enthalpy exiting the

break during the reflood period, for the RSGs, will be greater than

for the OSGs. This results in an increase in the containment

building peak pressure, Pa.

The proposed revisions to the Technical Specifications involve

the corrected value of the current Unit 1 and Unit 2 RCS volume and

the incremental change in RCS volume for the RSGs. The proposed

revisions also involve the defined value of Unit 1 Pa

following installation of the RSGs. Several editorial changes are

also being made to improve clarity and consistency of the TS.

RCS volume is not an initiator for any event and an increase in

volume does not affect any operating margin or requirements.

Therefore, increasing the primary volume does not increase the

probability of any event previously analyzed.

The current value of Pa for Unit 2 is unchanged due

to conservatism in the original analysis. The revised value of

Pa for Unit 1 continues to be less than the design basis

pressure for the containment structure. The change represents only a

revision to the containment test pressure for containment leakage

testing. Such testing is only performed with the affected unit in

the shutdown condition. Therefore, the proposed change in

Pa for Unit 1 does not involve a significant increase in

the probability of an accident previously evaluated.

All accidents in the Updated Final Safety Analysis Report

(UFSAR) were evaluated to determine the effect of an increase in

primary volume on accident consequences. The events identified that

may be impacted by an increase in primary volume are the Waste Gas

System Leak or Failure and LBLOCA. For the Waste Gas System Leak or

Failure, the activity of the decay tank is controlled to Technical

Specification limits which are unaffected by RCS volume. Therefore,

an increase in RCS volume would not increase the offsite dose.

The offsite dose calculation for the LBLOCA is unaffected by the

proposed

[[Page 66700]]

change. The license basis offsite dose calculation is in accordance

with NRC Reg Guide 1.4 ``Assumptions Used for Evaluating The

Potential Radiological Consequences of a Loss of Coolant Accident

for Pressurized Water Reactors.'' This Regulatory Guide states, in

part, ``* * * a number of appropriately conservative assumptions,

based on engineering judgment and on applicable experimental results

from safety research programs conducted by the AEC.'' These

conservatisms include (but are not limited to) the following

assumptions:

Twenty five percent of the equilibrium full power radioactive

iodine inventory is immediately available for leakage from the

primary containment. 100% of the equilibrium full power radioactive

noble gas inventory is immediately available for leakage from the

primary containment. The primary containment should be assumed to

leak at the (maximum) leak rate specified in the technical

specifications for the first 24 hours and at 50% of this value for

the remaining 29 days of the accident duration.

The design basis leakage corresponding to a peak containment

pressure of 50 psig utilized in the design basis accident analysis

is 0.10% per day of the containment free air mass. Therefore, the

offsite dose calculation was performed with a leakage of .1% per day

for day one and .05% per day for days 2 through 30. Isotopic

inventories are unaffected by the increase in reactor coolant

volume. Thus, the offsite dose is unaffected by the increase in the

peak containment pressure. Therefore, this proposed change to

Pa does not involve a significant increase in the

consequences of an accident previously evaluated.

The editorial changes proposed are for clarity and consistency

within the Technical Specifications and do not affect either the

probability or consequences of an accident previously evaluated.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed change in RCS volume is a change in a plant

parameter within the ``Design Features'' section of the Technical

Specifications. Increasing the RCS volume does not create any new or

different failure modes. The existing RCS design requirements

continue to be met.

The revised value of Pa for Unit 1 following

replacement of steam generators continues to be less than the design

basis pressure for the containment building structure. The change

represents only a revision to the test pressure for containment

leakage testing. Such testing is only performed with the affected

unit in the shutdown condition. Therefore, no new or different

failure modes are being introduced by modification of the testing

parameters.

The editorial changes proposed are for clarity and consistency

within the Technical Specifications and do not result in any

physical changes to the facility or how it is operated. No new or

different failure modes are being introduced by these changes.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Changing the RCS volume in the Technical Specifications does not

reduce the margin of safety. RCS volume is a design feature. An

evaluation of all UFSAR accidents was performed to determine the

effect of an increase in RCS volume. This evaluation is summarized

as follows:

An evaluation of the Chemical and Volume Control System

Malfunction was performed to determine the effect of the increased

RCS volume. The larger RCS volume reduces the reactivity insertion

for a given dilution flow rate. Therefore, the UFSAR analyses remain

bounding for Byron and Braidwood and there is no reduction in the

margin of safety.

An evaluation of the Inadvertent Actuation of the Emergency Core

Cooling System During Power Operation Event was performed to

determine the effect of the increased RCS volume due to the RSGs.

For this event, the injection of borated water causes a negative

reactivity insertion, which increases DNBR. For a given Refueling

Water Storage Tank (RWST) boron concentration, the larger RCS volume

will cause a reduction in the negative reactivity insertion rate as

compared to the current UFSAR analysis. However, negative reactivity

would still be inserted and no fuel pins would experience DNB.

Additionally, the increased RCS volume was evaluated to determine

the effect on pressurizer level following the inadvertent actuation

of ECCS and was found to be acceptable. Therefore, there is no

reduction in the margin of safety.

An evaluation of the Small Break LOCA was performed to determine

the effect of increased RCS volume. The additional RCS volume will

cause a delay in the loop seal clearing which in turn delays the

core uncovery as compared with the UFSAR analysis. A delay in core

uncovery reduces the amount of core heatup which results in a lower

peak clad temperature (PCT) because the core decay heat would be

less than in the UFSAR analysis. The benefit is considered small,

but there is still a benefit. Therefore, the increased RCS volume

does not result in a reduction in the margin of safety.

An evaluation of the Large Break LOCA was performed to determine

the effect of increased RCS volume for the RSGs. For a LB LOCA, the

increased RCS volume causes the blowdown phase of the event to be

longer. Increased blowdown phase, alone, could potentially result in

a higher PCT. However, the RSGs also have less resistance to flow

due to increased primary side steam generator flow area, which

results in a higher blowdown flow compared to the OSGs. The

increased blowdown flow will compensate for the longer blowdown

phase associated with the increased RCS volume. The net effect is

that the blowdown time (end of bypass) for the RSG will be the same

or decrease compared to the OSG. Reduced resistance to break flow

for the RSG compared to the OSG will result in a lower PCT for the

RSG compared to the OSG.

The increase in the current value of RCS volume in Unit 2 is

significantly less than the increase associated with the replacement

of the steam generators in Unit 1. The small increase in the RCS

volume will likely result in a slight increase in the blowdown

period. This slight increase in the blowdown period will have no

significant impact on the peak clad temperature (PCT) calculation

for Unit 2. Any small changes in the PCT due to this small increase

in the RCS volume can be easily accommodated for Unit 2 because of

the significant margin in the PCT (over 100 degrees) available to

the Appendix K 10 CFR 50.46 acceptance criteria of 2200 deg.F.

Therefore, there is no reduction in the margin of safety.

An evaluation of the Gas Waste System Leak or Failure was

performed to determine the effect of the increased RCS volume.

Because the activity of the decay tank is controlled within

Technical Specification limits, an increase in RCS volume would not

change the results of the event. Therefore, there is no reduction in

the margin of safety.

An evaluation was performed to determine the effect of the

increased RCS volume (associated with the RSGs) on the peak

containment pressure following a LBLOCA. The increased RCS volume

caused the peak containment pressure to increase to 47.8 psig. This

is still below the containment design pressure of 50.0 psig.

Therefore, there is no reduction in the margin of safety. The

increase in RCS volume for the existing units (without RSGs) remains

within the conservative volume used in the calculation of the

current peak containment pressure value of 44.4 psig. Therefore,

there is no reduction in the margin of safety.

This proposed change involves testing requirements designed to

demonstrate acceptable leakage rates are maintained. If acceptable

leakage rates are maintained as outlined in the Technical

Specifications, there will be no reduction in the margin of safety.

In the event of degradation of a containment seal that results in

unacceptable leakage, plant shutdown will occur as required by

Technical Specifications and administrative requirements in

accordance with approved plant procedures. Therefore, this proposed

change does not involve a significant reduction in a margin of

safety. The editorial changes proposed are for clarity and

consistency within the Technical Specifications and do not result in

any physical changes to the facility or how it is operated.

Therefore, the changes have no effect on the margin of safety.

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

[[Page 66701]]

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 January 20, 1998, 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 for Byron, the Byron Public Library District, 109

N. Franklin, P.O. Box 434, Byron, Illinois 61010; for Braidwood, 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, 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.

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: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,

by the above date. 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 Michael I. Miller, Esquire; Sidley

and Austin, One First National Plaza, Chicago, Illinois 60603, 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 30, 1997, as revised on

December 9, 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 rooms: for Byron,

located at the Byron Public Library District, 109 Franklin, P.O. Box

434, Byron, Illinois 61010; for Braidwood, the Wilmington Public

Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

Dated at Rockville, Maryland, this 15th day of December, 1997.

[[Page 66702]]

For the Nuclear Regulatory Commission.

George Dick, Jr.,

Project Manager, Project Directorate III-2, Division of Reactor

Projects--III/IV, Office of Nuclear Reactor Regulation.

[FR Doc. 97-33231 Filed 12-18-97; 8:45 am]

BILLING CODE 7590-01-P