# 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

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URL: https://www.frixlaw.com/law-library/documents/fr%3A97-33231

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** December 19, 1997
- **Citation:** 62 FR 66699

## Text

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, Pa.
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 Pa
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 Pa for Unit 2 is unchanged due
to conservatism in the original analysis. The revised value of
Pa 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
Pa 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
Pa 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 Pa 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

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A97-33231. Public record. Not legal advice.