[Federal Register Volume 62, Number 244 (Friday, December 19, 1997)]
[Notices]
[Pages 66699-66702]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-33231]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. STN 50-454, STN 50-455, STN 50-456, and STN 50-457]


Commonwealth Edison Company; Notice of Consideration of Issuance 
of Amendment to Facility Operating License, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing 
Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2

    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 (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.
    The proposed amendment would revise technical specification (TS) 
1.0, ``Definitions'', TS 3/4.6.1, ``Primary Containment'' and 
associated Bases; and TS 5.4.2, ``Reactor Coolant System Volume'' for 
Byron and Braidwood to support the steam generator replacement for Unit 
1 at each site. The replacement steam generators increase the reactor 
coolant system volume which results in a higher calculated peak 
containment pressure (Pa) value. The staff's proposed no significant 
hazards consideration determination for the requested change was 
published on April 23, 1997 (62 FR 19826).
    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.
    Each of the RSGs has a larger RCS primary side volume than the 
original steam generators (OSGs). As a result of the RCS volume 
increase, the mass and energy release during the blowdown phase of 
the large break loss of coolant accident (LBLOCA) is increased. 
Additionally, the heat transfer rate of the RSGs is greater than the 
OSGs, and the RSGs will operate at a slightly higher pressure than 
that for the OSGs. Consequently, the steam enthalpy exiting the 
break during the reflood period, for the RSGs, will be greater than 
for the OSGs. This results in an increase in the containment 
building peak pressure, P<INF>a</INF>.
    The proposed revisions to the Technical Specifications involve 
the corrected value of the current Unit 1 and Unit 2 RCS volume and 
the incremental change in RCS volume for the RSGs. The proposed 
revisions also involve the defined value of Unit 1 P<INF>a</INF> 
following installation of the RSGs. Several editorial changes are 
also being made to improve clarity and consistency of the TS.
    RCS volume is not an initiator for any event and an increase in 
volume does not affect any operating margin or requirements. 
Therefore, increasing the primary volume does not increase the 
probability of any event previously analyzed.
    The current value of P<INF>a</INF> for Unit 2 is unchanged due 
to conservatism in the original analysis. The revised value of 
P<INF>a</INF> for Unit 1 continues to be less than the design basis 
pressure for the containment structure. The change represents only a 
revision to the containment test pressure for containment leakage 
testing. Such testing is only performed with the affected unit in 
the shutdown condition. Therefore, the proposed change in 
P<INF>a</INF> for Unit 1 does not involve a significant increase in 
the probability of an accident previously evaluated.
    All accidents in the Updated Final Safety Analysis Report 
(UFSAR) were evaluated to determine the effect of an increase in 
primary volume on accident consequences. The events identified that 
may be impacted by an increase in primary volume are the Waste Gas 
System Leak or Failure and LBLOCA. For the Waste Gas System Leak or 
Failure, the activity of the decay tank is controlled to Technical 
Specification limits which are unaffected by RCS volume. Therefore, 
an increase in RCS volume would not increase the offsite dose.
    The offsite dose calculation for the LBLOCA is unaffected by the 
proposed

[[Page 66700]]

change. The license basis offsite dose calculation is in accordance 
with NRC Reg Guide 1.4 ``Assumptions Used for Evaluating The 
Potential Radiological Consequences of a Loss of Coolant Accident 
for Pressurized Water Reactors.'' This Regulatory Guide states, in 
part, ``* * * a number of appropriately conservative assumptions, 
based on engineering judgment and on applicable experimental results 
from safety research programs conducted by the AEC.'' These 
conservatisms include (but are not limited to) the following 
assumptions:
    Twenty five percent of the equilibrium full power radioactive 
iodine inventory is immediately available for leakage from the 
primary containment. 100% of the equilibrium full power radioactive 
noble gas inventory is immediately available for leakage from the 
primary containment. The primary containment should be assumed to 
leak at the (maximum) leak rate specified in the technical 
specifications for the first 24 hours and at 50% of this value for 
the remaining 29 days of the accident duration.
    The design basis leakage corresponding to a peak containment 
pressure of 50 psig utilized in the design basis accident analysis 
is 0.10% per day of the containment free air mass. Therefore, the 
offsite dose calculation was performed with a leakage of .1% per day 
for day one and .05% per day for days 2 through 30. Isotopic 
inventories are unaffected by the increase in reactor coolant 
volume. Thus, the offsite dose is unaffected by the increase in the 
peak containment pressure. Therefore, this proposed change to 
P<INF>a</INF> does not involve a significant increase in the 
consequences of an accident previously evaluated.
    The editorial changes proposed are for clarity and consistency 
within the Technical Specifications and do not affect either 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 proposed change in RCS volume is a change in a plant 
parameter within the ``Design Features'' section of the Technical 
Specifications. Increasing the RCS volume does not create any new or 
different failure modes. The existing RCS design requirements 
continue to be met.
    The revised value of P<INF>a</INF> for Unit 1 following 
replacement of steam generators continues to be less than the design 
basis pressure for the containment building structure. The change 
represents only a revision to the test pressure for containment 
leakage testing. Such testing is only performed with the affected 
unit in the shutdown condition. Therefore, no new or different 
failure modes are being introduced by modification of the testing 
parameters.
    The editorial changes proposed are for clarity and consistency 
within the Technical Specifications and do not result in any 
physical changes to the facility or how it is operated. No new or 
different failure modes are being introduced by these changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Changing the RCS volume in the Technical Specifications does not 
reduce the margin of safety. RCS volume is a design feature. An 
evaluation of all UFSAR accidents was performed to determine the 
effect of an increase in RCS volume. This evaluation is summarized 
as follows:
    An evaluation of the Chemical and Volume Control System 
Malfunction was performed to determine the effect of the increased 
RCS volume. The larger RCS volume reduces the reactivity insertion 
for a given dilution flow rate. Therefore, the UFSAR analyses remain 
bounding for Byron and Braidwood and there is no reduction in the 
margin of safety.
    An evaluation of the Inadvertent Actuation of the Emergency Core 
Cooling System During Power Operation Event was performed to 
determine the effect of the increased RCS volume due to the RSGs. 
For this event, the injection of borated water causes a negative 
reactivity insertion, which increases DNBR. For a given Refueling 
Water Storage Tank (RWST) boron concentration, the larger RCS volume 
will cause a reduction in the negative reactivity insertion rate as 
compared to the current UFSAR analysis. However, negative reactivity 
would still be inserted and no fuel pins would experience DNB. 
Additionally, the increased RCS volume was evaluated to determine 
the effect on pressurizer level following the inadvertent actuation 
of ECCS and was found to be acceptable. Therefore, there is no 
reduction in the margin of safety.
    An evaluation of the Small Break LOCA was performed to determine 
the effect of increased RCS volume. The additional RCS volume will 
cause a delay in the loop seal clearing which in turn delays the 
core uncovery as compared with the UFSAR analysis. A delay in core 
uncovery reduces the amount of core heatup which results in a lower 
peak clad temperature (PCT) because the core decay heat would be 
less than in the UFSAR analysis. The benefit is considered small, 
but there is still a benefit. Therefore, the increased RCS volume 
does not result in a reduction in the margin of safety.
    An evaluation of the Large Break LOCA was performed to determine 
the effect of increased RCS volume for the RSGs. For a LB LOCA, the 
increased RCS volume causes the blowdown phase of the event to be 
longer. Increased blowdown phase, alone, could potentially result in 
a higher PCT. However, the RSGs also have less resistance to flow 
due to increased primary side steam generator flow area, which 
results in a higher blowdown flow compared to the OSGs. The 
increased blowdown flow will compensate for the longer blowdown 
phase associated with the increased RCS volume. The net effect is 
that the blowdown time (end of bypass) for the RSG will be the same 
or decrease compared to the OSG. Reduced resistance to break flow 
for the RSG compared to the OSG will result in a lower PCT for the 
RSG compared to the OSG.
    The increase in the current value of RCS volume in Unit 2 is 
significantly less than the increase associated with the replacement 
of the steam generators in Unit 1. The small increase in the RCS 
volume will likely result in a slight increase in the blowdown 
period. This slight increase in the blowdown period will have no 
significant impact on the peak clad temperature (PCT) calculation 
for Unit 2. Any small changes in the PCT due to this small increase 
in the RCS volume can be easily accommodated for Unit 2 because of 
the significant margin in the PCT (over 100 degrees) available to 
the Appendix K 10 CFR 50.46 acceptance criteria of 2200  deg.F. 
Therefore, there is no reduction in the margin of safety.
    An evaluation of the Gas Waste System Leak or Failure was 
performed to determine the effect of the increased RCS volume. 
Because the activity of the decay tank is controlled within 
Technical Specification limits, an increase in RCS volume would not 
change the results of the event. Therefore, there is no reduction in 
the margin of safety.
    An evaluation was performed to determine the effect of the 
increased RCS volume (associated with the RSGs) on the peak 
containment pressure following a LBLOCA. The increased RCS volume 
caused the peak containment pressure to increase to 47.8 psig. This 
is still below the containment design pressure of 50.0 psig. 
Therefore, there is no reduction in the margin of safety. The 
increase in RCS volume for the existing units (without RSGs) remains 
within the conservative volume used in the calculation of the 
current peak containment pressure value of 44.4 psig. Therefore, 
there is no reduction in the margin of safety.
    This proposed change involves testing requirements designed to 
demonstrate acceptable leakage rates are maintained. If acceptable 
leakage rates are maintained as outlined in the Technical 
Specifications, there will be no reduction in the margin of safety. 
In the event of degradation of a containment seal that results in 
unacceptable leakage, plant shutdown will occur as required by 
Technical Specifications and administrative requirements in 
accordance with approved plant procedures. Therefore, this proposed 
change does not involve a significant reduction in a margin of 
safety. The editorial changes proposed are for clarity and 
consistency within the Technical Specifications and do not result in 
any physical changes to the facility or how it is operated. 
Therefore, the changes have no effect on the margin of safety.

    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

[[Page 66701]]

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 Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, 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.
    The filing of requests for hearing and petitions for leave to 
intervene is discussed below.
    By January 20, 1998, 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, and at the local public 
document room located for Byron, the Byron Public Library District, 109 
N. Franklin, P.O. Box 434, Byron, Illinois 61010; for Braidwood, the 
Wilmington 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 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to Michael I. Miller, Esquire; Sidley 
and Austin, One First National Plaza, Chicago, Illinois 60603, 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 January 30, 1997, as revised on 
December 9, 1997, which is available for public inspection at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC, and at the local public document rooms: for Byron, 
located at the Byron Public Library District, 109 Franklin, P.O. Box 
434, Byron, Illinois 61010; for Braidwood, the Wilmington Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

    Dated at Rockville, Maryland, this 15th day of December, 1997.


[[Page 66702]]


    For the Nuclear Regulatory Commission.
George Dick, Jr.,
Project Manager, Project Directorate III-2, Division of Reactor 
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 97-33231 Filed 12-18-97; 8:45 am]
BILLING CODE 7590-01-P