Commonwealth Edison Company; Consideration of Issuance of Amendments to Facility Operating Licenses Proposed no Significant Hazards Consideration Determination, and Opportunity for a Hearing

NUCLEAR REGULATORY COMMISSION

Commonwealth Edison Company; Consideration of Issuance of

Amendments to Facility Operating Licenses 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.

NPF-37, NPF-66, NPF-72, and NPF-77, issued to Commonwealth Edison

Company (ComEd, 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.

In a letter of August 13, 1993, and as supplemented on September

15, 1993, September 16, 1993, December 17, 1993, January 19, 1994,

February 11, 1994, and February 24, 1994, ComEd submitted requests for

amendments for steam generator (SG) tube sleeving in accordance with

(1) Westinghouse and (2) Babcock & Wilcox processes. By letter dated

March 4, 1994, the NRC granted the proposed sleeving methods contingent

upon four conditions which the licensee accepted in their letter of

February 24, 1994.

Three of the four changes will be reflected in the plants'

Technical Specifications (TS). By letter dated June 3, 1994, the

licensee requested changes to TS 3.4.5 and 3.4.6.2 to include the three

conditions, which are:

1. Amend the Byron and Braidwood licenses to reflect a primary-to-

secondary leakage rate limit of 150 gallons per day (gpd) through any

one SG.

2. Amend the Byron and Braidwood licenses to reflect an inservice

inspection of a minimum of 20 percent of a random sample of the sleeves

for axial and circumferential indication at the end-of-cycle. In the

event that an imperfection of 40 percent or greater depth is detected,

an additional 20 percent (minimum) of the unsampled sleeves should be

inspected, and if an imperfection of 40 percent or greater depth is

detected in the second sample, all remaining sleeves should be

inspected.

3. Add a condition to the Byron and Braidwood licenses to conduct

additional corrosion testing to establish the design life for the

kinetically or laser welded sleeved tubes in the presence of a crevice.

Collectively, these conditions will enable the licensee to have:

1. Further assurance that the integrity of the SGs will be

maintained in the event of a main steam line break or under loss-of-

coolant accident (LOCA) conditions;

2. Increased monitoring of the SG tube sleeves for any degradation;

and

3. Increased confidence that SG sleeve integrity will be maintained

for extended operations.

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.

The original amendment requested [approval] of tubesheet sleeves

and tube support plate sleeves as an alternate tube repair method

for Byron and Braidwood Units 1 and 2. The steam generator sleeves

approved for installation use the Westinghouse process (laser welded

joints) and the Babcock and Wilcox (B&W) process of kinetically

welded joints. The sleeve configuration was designed and analyzed in

accordance with the criteria of Regulatory Guide (RG) 1.121 and the

design requirements of Section III of the American Society of

Mechanical Engineers (ASME) Code. Fatigue and stress analyses of the

sleeved tube assemblies for both processes produced acceptable

results as documented in the Westinghouse and the B&W topical

reports submitted in the original sleeving package. Mechanical

testing has shown that the structural strength of the sleeves under

normal, faulted, and upset conditions is within acceptable limits.

Leakage rate testing for the tube sleeves has demonstrated that

primary-to-secondary leakage is not expected during all plant

conditions.

Any leakage through the sleeved region of the tube is fully

bounded by the leak-before-break considerations and, ultimately, the

existing steam generator tube rupture analysis included in the Byron

and Braidwood Updated Final Safety Analysis Report (UFSAR). The

reduction in TS leakage rate requirements from 500 gpd allowable per

SG to 150 gpd further ensures that SG tube integrity is maintained

in the event of a main steam line break (MSLB) or under Loss Of

Coolant Accident (LOCA) conditions. The RG 1.121 criteria for

establishing operational leakage rate limits require a plant

shutdown based upon a leak-before-break consideration to detect a

free span crack before a potential tube rupture. The 150 gpd limit

will continue to allow for early leakage detection and require a

plant shutdown in the event of the occurrence of an unexpected crack

resulting in leakage that exceeds the revised Technical

Specification limit.

The sleeve sample size has been increased to a minimum of twenty

(20) percent of the inservice sleeves. Increasing the sample size of

the sleeves to be inspected will increase the monitoring of tubes

using sleeves for any further degradation while they remain

inservice. If the sample identifies a sleeve with an imperfection of

greater [than] 40 percent depth, an additional 20 percent of the

sleeves shall be inspected. The sleeves that have identified

imperfections of greater than 40 percent shall be evaluated and

removed from service. The inservice inspections and additional

corrosion testing for the sleeves and welded joints will continue

until the corrosion resistance is demonstrated acceptable to the

NRC. If conformance with the acceptance criteria of section 4.4.5.4

for tube structural integrity is not confirmed, the tubes containing

the sleeves in question shall be removed from service. Increasing

the monitoring of the sleeved tubes will decrease the probability of

occurrence of an accident previously evaluated in the UFSAR.

Implementation of a corrosion testing program should determine

the effects that material microstructure, chemistry, and joint

crevices will have on primary water stress corrosion cracking

initiation and growth. This program will not cause an increase in

the probability or consequence of an accident previously evaluated

because the testing program is conducted in laboratory conditions.

If the results of the testing program do not confirm the structural

integrity of the tubes, the tubes containing the sleeves in question

shall be removed from service. These changes do not involve a

significant increase in 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 implementation of the proposed amendment will not introduce

significant or adverse changes to the plant design basis. The

proposed changes do not involve plant modification or changes to

equipment, and consist of: reduction in allowable steam generator

leakage limits, increase in the sample size of the steam generator

tube sleeved and the addition of a commitment to perform a corrosion

testing program on the sleeved tubes.

The reduction in TS leakage rate requirements from 500 gpd

allowable per SG to 150 gpd further ensures that SG tube integrity

is maintained in the event of a MSLB or under LOCA conditions. The

150 gpd limit is designed to provide for leakage detection and a

plant shutdown in the event of the occurrence of an unexpected

single crack resulting in excessive tube leakage. The limit provides

for early detection and a plant shutdown prior to a postulated crack

reaching critical crack lengths for Main Steam Line Break

conditions.

Increasing the sample size of tubes sleeved during each

scheduled inservice inspection will increase the monitoring of these

tubes for any further degradation. The improved monitoring and

evaluation of the tube and the sleeves assures tube structural

integrity is maintained or the tube is removed for service.

Additionally, corrosion testing to establish sleeve design life

and corrosion resistance to confirm tube structural integrity will

be performed. If the tube structural integrity is not confirmed, the

tubes containing the sleeves in question shall be removed from

service.

With these actions the possibility of a new or different type of

accident from any accident previously evaluated is not created.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Implementation of the proposed changes will not reduce the

margin of safety. This amendment involves the reduction of steam

generator leakage limit, and increase in the amount of sleeved tubes

inspected and the incorporation of a corrosion testing program for

sleeved tubes. All of these actions will help ensure steam generator

tube integrity.

Reduction of the leakage rate requirement from 500 to 150

gallons per day (gpd) per steam generator will continue to ensure

steam generator tube integrity is maintained in the event of main

steam line break or under LOCA conditions. The reduction to 150 gpd

also limits the allowable primary-to-secondary leakage from 1 gallon

per minute to 600 gpd for all steam generators not isolated from the

Reactor Coolant System (RCS). This previous leakage limit, used in

UFSAR accident analysis, ensured the dosage contribution from tube

leakage would be limited to a small fraction of the 10 CFR Part 100

dose guideline values in the event of either a steam generator tube

rupture or steam line break. Reducing these limits will not result

[in] a reduction in the margin of safety.

The portions of the installed sleeve assembly which represent

the reactor coolant pressure boundary can be monitored for the

initiation and progression of sleeve/tube wall degradation, thus

satisfying the requirement of Regulatory Guide 1.83. The portion of

the tube bridged by the sleeve joints is effectively removed from

the pressure boundary, and the sleeve then forms the new pressure

boundary. The sleeve enhances the safety of the plant by increasing

the protective boundaries of the steam generator. Keeping the tube

in service with the use of a sleeve instead of plugging the tube and

removing it from service increases the heat transfer efficiency of

the steam generator. Monitoring for any increased degradation of a

repaired steam generator tube shall be implemented at Byron and

Braidwood by increasing the sampling size of inservice sleeves to

include an additional twenty (20) percent of the sleeves inservice.

During each scheduled in service inspection, each sampled sleeve

evaluated and found to have unacceptable degradation shall be

removed from service.

Implementation of a corrosion testing program should determine

the effects that material microstructure, chemistry, and joint

crevices will have on primary water stress corrosion cracking

initiation and growth. This program is conducted in laboratory

setting; therefore, [it] will not involve a significant reduction in

a margin of safety. In addition, the corrosion testing program will

be performed to establish sleeve design life and corrosion

resistance to confirm tube structural integrity. If the tube

structural integrity is not confirmed, the tubes containing the

sleeves in question shall be removed from service. These actions

[do] not involve a significant reduction in a margin of safety.

Based on the preceding analysis it is concluded that operation of

Byron and Braidwood Units 1 and 2, in accordance with the proposed

amendment does not increase the probability of an accident previously

evaluated, does not create the possibility of a new or different kind

of accident previously evaluated, nor reduce any margins to plant

safety. Therefore, the license amendment does not involve a Significant

Hazards Consideration as defined in 10 CFR 50.92.

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 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 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

20555.

The filing of requests for hearing and petitions for leave to

intervene is discussed below.

By November 14, 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, Byron, Illinois 61010; and for Braidwood is

located at the Wilmington Township 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, 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 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. Robert A. Capra: 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 June 3, 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 Township Public Library, 201 S.

Kankakee Street, Wilmington, Illinois 60481.

Dated at Rockville, Maryland, this 4th day of October 1994.

For the Nuclear Regulatory Commission.

Ramin R. Assa,

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

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

[FR Doc. 94-25158 Filed 10-11-94; 8:45 am]

BILLING CODE 7590-01-M