[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