[Federal Register Volume 59, Number 184 (Friday, September 23, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-23730]


[[Page Unknown]]

[Federal Register: September 23, 1994]


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NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-454 and STN 50-455]

 

Consideration of Issuance of Amendment to Facility Operating 
License, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing; Commonwealth Edison Co.

    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of an amendment to Facility Operating License Nos. 
NPF-37 and NPF-66, issued to Commonwealth Edison Company (the 
licensee), for operation of Byron Station, Units 1 and 2, located in 
Ogle County, Illinois.
    The proposed amendment would revise the technical specifications 
(TS) to incorporate a 1.0 volt steam generator tube interim plugging 
criteria (IPC) for Unit 1 beginning with Cycle 7, which has begun. This 
supplements the information that was published in the Federal Register 
on August 31, 1994 (59 FR 45019).
    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.
    Consistent with Regulatory Guide (RG) 1.121, ``Basis for 
Plugging Degraded PWR Steam Generator Tubes,'' Revision 0, August 
1976, the traditional depth-based criteria for SG tube repair 
implicitly ensures that tubes accepted for continued service will 
retain adequate structural and leakage integrity during normal 
operating, transient, and postulated accident conditions. It is 
recognized that defects in tubes permitted to remain in service, 
especially cracks, occasionally grow entirely through-wall and 
develop small leaks. Limits on allowable primary-to-secondary 
leakage established in Technical Specifications ensure timely plant 
shutdown before the structural and leakage integrity of the affected 
tube is challenged.
    The proposed license amendment request to implement voltage 
amplitude [steam generator] SG tube support plate Interim Plugging 
Criteria for Byron Unit 1 Cycle 7 meets the requirements of RG 
1.121. The IPC methodology demonstrates that tube leakage is 
acceptably low and tube burst is a highly improbable event during 
either normal operation or the most limiting accident condition, a 
postulated main steam line break (MSLB) event. Requesting a single 
cycle applicability is more conservative than the guidance contained 
in the draft Generic Letter on voltage-based repair criteria issued 
for comment on August 12, 1994.
    Adequate SG tube leakage integrity during normal operating 
conditions is assured by limiting allowable primary-to-secondary 
leakage to 150 gpd per SG or 600 gpd total. Currently, this limit is 
administratively controlled.
    However, a license amendment request was submitted on June 3, 
1994, to incorporate this limit into the Byron Technical 
Specifications. During normal operating conditions, the tube support 
plate constrains the [outer diameter stress corrosion cracking] 
ODSCC affected area of the tube to provide additional strength that 
precludes burst. Any leakage of a tube exhibiting ODSCC at the [tube 
support plate] TSP is fully bounded by the existing SG tube rupture 
analysis included in the Byron [Updated Final Safety Analysis 
Report] UFSAR. Therefore, probability of failure of a tube left in 
service or consequences of tube failure during normal operating 
conditions is not significantly increased by the application of IPC.
    During transients, the TSP is conservatively assumed to displace 
due to the thermal-hydraulic loads associated with the transient. 
This may partially expose a crack which is within the boundary of 
the TSP during normal operations to free span conditions. Burst is 
therefore conservatively evaluated assuming the crack is fully 
exposed to free span conditions. The structural eddy current bobbin 
coil voltage limit for free-span burst is 4.54 volts. This limit 
takes into consideration a 1.43 safety factor applied to the steam 
line break differential pressure that is consistent with RG 1.121 
requirements. With additional considerations for growth rate 
assumptions and an upper 95% confidence estimate on voltage 
variability, the maximum voltage indication that could remain in 
service is reduced to 2.7 volts. For added conservatism, the 
allowable indication voltage is further reduced in the proposed 
amendment to a 1.0 volt confirmed ODSCC indication limit. All 
indications greater than 1.0 volt will be subject to an RPC 
examination. Tubes with RPC confirmed ODSCC indications will be 
plugged or sleeved. Any ODSCC indications between 1.0 volt and 2.7 
volts which are not confirmed as ODSCC will be allowed to remain in 
service since these indications are not as likely to affect tube 
structural integrity or leakage integrity over the next operating 
cycle as the indications that are detectable by both bobbin and RPC 
inspections.
    The eddy current inspection process has been enhanced to address 
RG 1.83, ``Inservice Inspection of PWR Steam Generator Tubes,'' 
Revision 1, July 1975. consideration as well as the EPRI SG 
Inspection Guidelines. Enhancements in accordance with NUREG-1477 
and Appendix A of the Catawba IPC report (WCAP-13698) are in place 
to increase detection of ODSCC indications and to ensure reliable, 
consistent acquisition and analysis of data. Based on the 
conservative selection of the voltage criteria and the increased 
ability to identify ODSCC, the probability of tube failure during an 
accident is also not significantly increased due to application of 
requested IPC.
    For consistency with current offsite dose limits, the site 
allowable leakage limit during a MSLB has been conservatively 
calculated to be 12.8 gpm. This leakage limit includes maximum 
allowable operational leakage from the unaffected SGs and the 
accident leakage from the affected SG. As a requirement for 
operation following application of IPC, the projected distribution 
of crack indications over the operating period must be verified to 
result in primary to secondary accident leakage less than the site 
allowable leakage limit. Thus, the consequences of a MSLB remain 
unchanged.
    For an unscheduled mid-cycle inspection as a result of leakage 
due to mechanisms other than ODSCC at support plates or some other 
cause, the ODSCC indication limit is represented by the following 
equation:


TN23SE94.003

where:

V=measured voltage
VBOC=voltage at BOC
t=time period of operation to unscheduled outage
CL=cycle length (full operating cycle length where operating cycle 
is the time between two scheduled steam generator inspections)
VSL=4.5 volts

    Assuming linear flaw growth from BOC to EOC and a maximum 
structural limit of 4.5 volts, the voltage expected for an 
identified flaw at any time in the cycle can be predicted. The 
allowed voltage limit for an unscheduled inspection, as identified 
by the equation given above, reduces the predicted straightline 
growth voltage to ensure conservatism in the limit. A flaw which has 
not exceeded the predicted voltage growth at any point in the cycle 
would not be expected to exceed the structural limit at end of cycle 
or negatively impact the burst probability calculated based on 
results from the last scheduled inspection. Therefore, it is 
acceptable to leave the tube in service.
    Therefore, as implementation of the 1.0 volt IPC for Byron Unit 
1 Cycle 7 does not adversely affect steam generator tube integrity 
and results in acceptable dose consequences, the proposed license 
amendment request does not result in any significant increase in the 
probability or consequences of an accident previously evaluated 
within the Byron Updated Final Safety Analysis Report.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed SG tube IPC for Byron Unit 1 
Cycle 7 does not introduce any significant changes to the plant 
design basis. Use of the criteria does not provide a mechanism which 
could result in an accident outside the tube support plate 
elevations since industry experience indicates that ODSCC 
originating within the tube support plate does not extend 
significantly beyond the thickness of the support plate. This 
criteria only applies to ODSCC contained within the region of the 
tube bounded by the tube support plate.
    In addressing the combined effects of Loss of Coolant Accident 
(LOCA) coincident with a Safe Shutdown Earthquake (SSE) on the SG 
(as required by General Design Criterion 2), it has been determined 
that tube collapse of select tubes may occur in the SGs at some 
plants, including Byron Unit 1. There are two issues associated with 
SG tube collapse. First, the collapse of SG tubing reduces the RCS 
flow area through the tubes. The reduction in flow area increases 
the resistance to flow of steam from the core during a LOCA which, 
in turn, may potentially increase Peak Clad Temperature (PCT). 
Second, there is a potential that partial through-wall cracks in 
tubes could progress to through-wall cracks during tube deformation 
or collapse.
    A number of tubes have been identified, in the ``wedge'' 
locations of the SG TSPs, that demonstrate the potential for tube 
collapse during a LOCA + SSE event. Because of this potential, these 
tubes have been excluded from application of the voltage-based SG 
TSP IPC.
    Therefore, neither a single nor multiple tube rupture event 
would be expected in a steam generator in which IPC has been 
applied.
    ComEd has implemented a maximum primary to secondary leakage 
limit of 150 gpd through any one SG at Byron to help preclude the 
potential for excessive leakage during all plant conditions. The 150 
gpd limit provides for leakage detection and plant shutdown in the 
event of an unexpected single crack leak associated with the longest 
permissible free span crack length. The 150 gpd limit provides 
adequate leakage detection and plant shutdown criteria in the event 
an unexpected single crack results in leakage that is associated 
with the longest permissible free span crack length. Since tube 
burst is precluded during normal operation due to the proximity of 
the TSP to the tube and the potential exists for the crevice to 
become uncovered during MSLB conditions, the leakage from the 
maximum permissible crack must preclude tube burst at MSLB 
conditions. Thus, the 150 gpd limit provides a conservative limit to 
prompt plant shutdown prior to reaching critical crack lengths under 
MSLB conditions.
    Upon implementation of the 1.0 volt IPC for Byron Unit 1 Cycle 
7, steam generator tube integrity continues to be maintained through 
inservice inspection and primary-to-secondary leakage monitoring. 
Therefore, the possibility of a new or different kind of accident 
from any previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of the voltage based bobbin coil probe SG TSP IPC for 
Byron Unit 1 Cycle 7 will maintain steam generator tube integrity 
commensurate with the criteria of RG 1.121 as discussed above. Upon 
implementation of the criteria, even under the worst case 
conditions, the occurrence of ODSCC at the TSP elevations is not 
expected to lead to a steam generator tube rupture event during 
normal or faulted plant conditions. The distribution of crack 
indications at the TSP elevations result in acceptable primary-to-
secondary leakage during all plant conditions and radiological 
consequences are not adversely impacted by the application of IPC.
    The installation of SG tube plugs and sleeves reduces the RCS 
flow margin. As noted previously, implementation of the SG TSP IPC 
will decrease the number of tubes which must be repaired by plugging 
or sleeving. Thus, implementation of IPC will retain additional flow 
margin that would otherwise be reduced due to increased tube 
plugging. Therefore, no significant reduction in the margin of 
safety will occur during Cycle 7 as a result of the implementation 
of this proposed license amendment request.
    Although not relied upon to prove adequacy of the proposed 
amendment request, the following analyses demonstrate that 
significant conservatism exists in the methods and justifications 
described above:

Limited Tube Support Plate Displacement

    An analysis was performed to verify the extent of limited TSP 
displacement during accident conditions (MSLB). Application of 
minimum TSP displacement assumptions reduce the likelihood of a tube 
burst to negligible levels.
    Consideration of limited TSP displacement would also reduce 
potential MSLB leakage when compared to the leakage calculated 
assuming free span indications.

Probability of Detection

    The Electric Power Research Institute (EPRI) Performance 
Demonstration Program analyzed the performance of approximately 20 
eddy current data analysts evaluating data from a unit with \3/4\'' 
inside diameter and 0.049'' wall thickness tubes. The results of 
this analysis clearly show that the detectability of larger voltage 
indications is increased which lends creditability for application 
of a POD of >0.6 for ODSCC indications larger than 1.0 volt.

Risk Evaluation of Core Damage

    As part of ComEd's evaluation of the operability of Byron Unit 1 
Cycle 7, a risk evaluation was completed. The objective of this 
evaluation was to compare core damage frequency under containment 
bypass conditions, with and without the interim plugging criteria 
applied at Byron Unit 1.
    The total Byron core damage frequency is estimated to be 3.09E-5 
per reactor year with a total contribution from containment bypass 
sequences of 3.72E-8 per reactor year according to the results of 
the current individual plant evaluation (IPE). Operation with the 
requested IPC resulted in an insignificant increase in core damage 
frequency resulting from MSLB with containment bypass conditions.

    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, 11455 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 October 24, 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 room located at the Byron Public Library, 109 N. 
Franklin, Byron, Illinois 61010. 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 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. 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 
applications for amendment dated September 7, 1994, and September 17, 
1994, (two letters) which are 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 room located 
at the Byron Public Library, 109 N. Franklin, P.O. Box 434, Byron, 
Illinois 61010.

    Dated at Rockville, Maryland, this 21st day of September 1994.

    For the Nuclear Regulatory Commission.
Robert A. Capra,
Director Project Directorate III-2, Division of Reactor Projects--III/
IV, Office of Nuclear Reactor Regulation.
[FR Doc. 94-23730 Filed 9-22-94; 8:45 am]
BILLING CODE 7590-01-M