[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
