[Federal Register Volume 59, Number 196 (Wednesday, October 12, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-25158]


[[Page Unknown]]

[Federal Register: October 12, 1994]


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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