[Federal Register Volume 62, Number 48 (Wednesday, March 12, 1997)]
[Notices]
[Pages 11483-11505]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-5999]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from February 14, 1997, through February 28,
1997. The last biweekly notice was published on February 26, 1997.
Notice of Consideration of Issuance of Amendments to Facility Opeating
Licenses, Proposed No Significant Harzards Consideration determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve 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. The basis
for this proposed determination for each amendment request is shown
below.
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
before action is taken. 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 a hearing and
petitions for leave to intervene is discussed below.
By April 11, 1997, 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
[[Page 11484]]
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 for the particular facility involved. 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 a 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 a 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: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
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 (Project Director): 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 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
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 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 room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendment requests: December 4, 1996
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TS) to reflect a change in the
method for detecting a reactivity anomaly described in TS 3.1.2 and TS
Surveillance Requirement 4.1.2. Actual keff will be compared to
predicted core keff instead of comparing actual and predicted
control rod density to determine if a reactivity anomaly exists.
Additionally, editorial changes to the Bases for TS 3/4.1.2 are
proposed to support the TS amendments.
Basis for proposed no significant hazards consideration
determination: 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 amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed license amendments modify the method of
detecting a reactivity anomaly. The proposed license amendments
allow using core keff to detect a reactivity anomaly instead of
control rod density. The correlation between core
[[Page 11485]]
reactivity and control rod density depends on predicting core
keff. Core keff can be readily monitored with the new
plant process computer program and core keff can more
accurately detect a reactivity anomaly in the core (assumptions are
minimized). A reactivity anomaly is not considered an initiator of
any previously analyzed accident. As such, changing the method of
detecting a reactivity anomaly will not increase the probability of
any accident previously evaluated. Although, a reactivity anomaly
could impact the consequences of a previously analyzed accident, the
consequences of an event occurring using the proposed method of
detecting a reactivity anomaly are the same as the consequences of
an event occurring using the current method of detecting a
reactivity anomaly. As a result, the proposed amendments do not
involve a significant increase in the consequences of any accident
previously evaluated.
2. The proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The proposed license amendments do not involve a physical
modification to the plant. The proposed license amendments also
continue to verify that the reactivity difference between predicted
and actual are such that a reactivity anomaly does not exist. In
addition, core keff can more accurately detect a reactivity
anomaly in the core (assumptions are minimized) and can be readily
monitored with the new plant process computer program. Therefore,
the change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety. The proposed license amendments
modify the method of detecting a reactivity anomaly. The proposed
license amendments allow using core keff to detect a reactivity
anomaly instead of control rod density. The correlation between core
reactivity and control rod density depends on predicting core
keff. Core keff can be readily monitored with the new
plant process computer, and core keff can more accurately
detect a reactivity anomaly in the core (assumptions are minimized).
Therefore, the proposed license amendments do not involve a
significant reduction in a margin of safety.
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.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart, Acting
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant (BSEP), Units 1 and 2, Brunswick County,
North Carolina
Date of amendment requests: January 7, 1997.
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TS) to: (1) exchange the reactor
pressure vessel pressure-temperature (P-T) limits curves currently
located in the Unit 1 and 2 TS; and (2) delete the current 8, 10, and
12 effective full power year (EFPY) hydrostatic test P-T limits curves
and incorporate new 14 and 16 EFPY hydrostatic test P-T limits curves
for the Unit 1 and 2 reactor pressure vessels. As reported in Licensee
Event Report (LER) 1-94-05 dated March 22, 1994, and LER supplements
dated April 29, 1994, and September 23, 1994, the licensee, the
Carolina Power & Light Co. (CP&L), determined that the Unit 1 and 2 P-T
limits curves had been inadvertently transposed and evaluated the
effects of the transposition. The proposed amendments correct this
transposition error. The proposed changes to the hydrostatic test P-T
limits curves are required because it is anticipated that both units
will exceed 12 EFPY during 1997.
Basis for proposed no significant hazards consideration
determination: 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:
This Technical Specification Change Request makes the following
changes:
1. Exchanges the pressure-temperature limits curves currently
located in the Unit 1 and Unit 2 Technical Specifications. In
Licensee Event Report 1-94-05, CP&L reported that the Unit 1 and
Unit 2 pressure-temperature limits curves had been inadvertently
transposed. This request is an administrative change to relocate the
pressure-temperature limits curves to Technical Specifications of
the unit to which they correctly correspond.
2. Deletes the current 8, 10 and 12 effective full power year
(EFPY) hydrostatic test pressure-temperature limits curves and
incorporates new 14 and 16 effective full power year (EFPY)
hydrostatic test pressure-temperature limits curves for the
Brunswick Unit 1 and 2 reactors. The current reactor vessel
pressure-temperature limits curves contained in the technical
specifications for hydrostatic pressure tests are suitable for up to
12 effective full power years (EFPY) of reactor operation. It is
anticipated that both units will surpass this threshold during 1997.
Based on this, new pressure-temperature limits curves for 14 and 16
EFPY were developed. Commensurate changes to the references in
Technical Specification 3/4.4.6.1 and Bases 3/4.6 are also proposed
to reflect the deletion of current Technical Specification Figure
3.4.6.1-3c.
3. Reformat[s] the pressure-temperature limits curves in
Technical Specification Figures 3.4.6.1-1, 3.4.6.1-2, 3.4.6.1-3a,
and 3.4.6.1-3b. The changes associated with reformatting the Figures
are administrative in nature.
Items 1, 2, and 3 do not involve a significant increase in the
probability or consequences of an accident previously evaluated
because of the following reasons:
1. Item 1 will exchange the Unit 1 and Unit 2 pressure-
temperature limits curves. This change is considered administrative
in nature. The pressure-temperature limits curves were developed
based on design and materials information for the reactor vessel;
however, due to an administrative error during the development of
the curves, the materials information for the Unit 1 and Unit 2
reactor vessels was inadvertently reversed. Proposed change 1 is
being made to exchange the reactor coolant system pressure-
temperature limits curves. Therefore, since this proposed change
does not involve a change to the pressure-temperature limits curves
nor a change to the configuration of the facility, the probability
of an accident previously evaluated is not increased.
Item 2 deletes the current Technical Specification hydrostatic
test pressure-temperature limits curves and replaces them with
updated curves. The current hydrostatic test pressure-temperature
limits curves, which are valid through 12 EFPY are expected to
expire during 1997; therefore, new hydrostatic test pressure-
temperature limits curves were developed through 16 EFPY. These new
hydrostatic test pressure-temperature limits curves will ensure that
the integrity of the Brunswick Units 1 and 2 reactor pressure
vessels is maintained during hydrostatic and leak tests up to 16
effective full power years of operation. The calculations used to
generate the new pressure-temperature limits curves were performed
using Appendix G to Section XI of the ASME Boiler and Pressure
Vessel Code, Welding Research Council Bulletin 175, and Appendix A
to Section XI of the ASME Boiler and Pressure Vessel Code, and
[incorporate] the requirements of 10 CFR 50, Appendix G, Section
IV.A.2. For pressure-temperature limit curve development, the
methods described in Appendix G to Section XI of the ASME Boiler and
Pressure Vessel Code are equivalent to the methods described in
Appendix G to Section III of the ASME Boiler and Pressure Vessel
Code. The proposed pressure-temperature limits curves, for
hydrostatic and leak tests, take into consideration the effects of
neutron irradiation on reactor vessel materials and provide the
necessary margin, as specified by Appendix G of 10 CFR 50, to assure
the structural integrity of the reactor coolant pressure boundary.
Based on the above, it is concluded that this change will not
increase the probability of an accident previously evaluated.
[[Page 11486]]
Item 3 reformats each of the Technical Specification Figures
containing the pressure-temperature limits curves. The changes
associated with the reformatting of proposed Technical Specification
Figures 3.4.6.1-1, 3.4.6.1-2, 3.4.6.1-3a, and 3.4.6.1-3b reflect
presentation preferences and do not result in technical changes
(either actual or interpretational) to the requirements of the
pressure-temperature limits curves. Therefore, the changes
associated with reformatting the Technical Specification Figures
containing the pressure-temperature limits curves are considered to
be administrative in nature. Based on the above, it is concluded
that this change will not increase the probability of an accident
previously evaluated.
The proposed license amendments do not alter Limiting Safety
System Settings nor Safety Limits. The proposed license amendments
do not revise the technical bases from which the pressure-
temperature limits curves were derived, and do not affect stresses
and fatigue for transients and design basis events for which the
reactor vessels were designed. The operation of plant equipment is
not significantly impacted by the proposed license amendments. The
proposed pressure-temperature limits curves provide the necessary
margin to ... assure the structural integrity of the reactor coolant
pressure boundary is maintained. This margin is designed to preclude
the probability of a reactor coolant pressure boundary failure. In
addition, since the proposed pressure-temperature limits curves are
based on current regulatory requirements and fluence data, the
consequences of a reactor coolant pressure boundary failure are not
impacted by the proposed license amendments. Therefore, the proposed
license amendments do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated. The proposed license amendments will ensure
that acceptable pressure-temperature limits are imposed on the
reactor pressure vessels during all phases of plant operation,
thereby ensuring the structural integrity of the reactor pressure
vessels. The pressure-temperature limits curves are designed to
provide fracture protection for the reactor coolant pressure
boundary and do not create any new accident modes. Accident modes
for the reactor coolant pressure boundary, due to nonductile
failure, are well understood by the industry. The proposed pressure-
temperature limits curves and the Technical Specifications continue
to provide controls to preclude such a failure. In addition, the
proposed license amendments do not result in physical changes to the
facility, nor do the proposed license amendments alter safety-
related equipment, or safety functions. Therefore, the proposed
license amendments do not create a new or different kind of accident
from any previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety. The pressure-temperature limits
curves are designed to provide a specific margin of safety. This
margin is required to be at least as great as that specified in
Appendix G to Section III of the ASME Boiler and Pressure Vessel
Code and Appendix G to 10 CFR 50. The proposed pressure-temperature
limits curves were developed based on design and materials
information for the reactor vessels, current regulatory requirements
and fluence data. The proposed pressure-temperature limit curves are
based on analyses that ensure that the fracture toughness margins of
10 CFR Part 50, Appendix G are not exceeded. Therefore, the proposed
license amendments do not involve a significant reduction in the
margin of safety.
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.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart, Acting.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. STN
50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: April 29, 1996, as supplemented on
January 21, 1997.
Description of amendment request: The proposed amendment would:
1. Revise Technical Specification (TS) 3.7.1.1, Action a., to
require the unit to be in hot shutdown, rather than cold shutdown, for
consistency with NUREG-1431, ``Standard Technical Specifications for
Westinghouse Plants,'' and add a new Action b. to clarify the shutdown
requirements when there are more than three inoperable main steam line
Code safety valves on any one steam generator.
2. Revise TS Surveillance Requirement 4.7.1.1 to clarify that
Specification 4.0.4 does not apply for entry into Mode 3 for Byron and
Braidwood and, for Braidwood only, delete the one-time requirements for
Unit 1, Cycle 5 and Unit 2 after outage A2F27.
3. Revise the maximum allowable power range neutron flux high trip
setpoints in Table 3.7-1.
4. Revise Table 3.7-2 to increase the as-found main steam safety
valve (MSSV) lift setpoint tolerance to plus/minus 3%, provide an as-
left setpoint tolerance of plus/minus 1%, and change a table notation.
5. Delete the orifice size column from Table 3.7-2.
6. Revise the Bases for TS 3.7.1.1 to be consistent with the
proposed changes to TS 3.7.1.1.
Basis for proposed no significant hazards consideration
determination: 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:
A. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The text describing reactor coolant loops and steam generators
is redundant. TS 3.4.1.1, ``Reactor Coolant Loops and Coolant
Circulation--Startup and Power Operation,'' and 3.4.1.2, ``Reactor
Coolant Loops and Coolant Circulation--Hot Standby,'' provide
restrictions on the number of operating reactor coolant loops and
steam generators. Therefore, deleting the text that requires having
four reactor coolant loops and associated steam generators in
operation from TS 3.7.1.1, Action a., has no impact on any analyzed
accident.
The proposed change to TS 3.7.1.1, Action a., to require the
final mode to be hot shutdown rather than cold shutdown is
consistent with the Applicability section of the specification,
which does not require the MSSVs to be operable in hot shutdown.
There are no credible transients requiring the MSSVs in modes 4 and
5. The steam generators are not normally used for heat removal in
modes 5 and 6, and thus cannot be overpressurized. The change also
eliminates the unnecessary transient that had been imposed on the
unit by forcing entry into cold shutdown.
The new Action b. for TS 3.7.1.1 and text changes to Action a.
clarify the shutdown requirement times based on the number of
inoperable valves. There are no changes to these times.
Changing TSSR 4.7.1.1 to delete the one-time requirements
imposed by previous amendments and allow entry into Mode 3 prior to
performing the requirements of TSSR 4.0.5 has no impact on any
accident. The change permits testing the MSSVs in accordance with
the applicable codes and allows a reasonable amount of time for
completion of the surveillance. The conditions requiring the one-
time requirements have been corrected, so the one-time requirements
are no longer required.
The proposed setpoints in Table 3.7-1 are more limiting than
those currently allowed in Specification 3.7.1.1. Westinghouse
[[Page 11487]]
determined that the current setpoints are non-conservative for some
combinations of reduced MSSV availability and reactor power levels.
By reducing the setpoints, the original design margins for safety
will be met. Reduced reactor trip setpoints due to reduced
availability of the MSSVs are not precursors to any accidents, but
are used in the safety analysis to establish that plant response
will be within required margins for accidents of concern.
Increasing the as-found valve setpoint tolerance from plus/minus
1% to plus/minus 3% does not have a significant impact on any
accident. The peak primary and secondary pressures remain below 110%
of design at all times. The departure from nucleate boiling ratio
and peak cladding temperature values remain within the specified
limits of the licensing basis. All of the applicable loss-of-coolant
accident (LOCA) and non-LOCA design basis acceptance criteria remain
valid.
The MSSVs are actuated after accident initiation to protect the
secondary systems from overpressurization. Increasing the as-found
setpoint tolerance will not result in any hardware modification to
the MSSVs. Therefore, there is not an increase in the probability of
the spurious opening of a MSSV. Sufficient margin exists between the
normal steam system operating pressure and the valve setpoint with
the increased tolerance to preclude an increase in the probability
of actuating the valves. The MSSVs also remain capable of relieving
any unlikely system overpressure during all applicable operating
modes.
Although increasing the as-found valve setpoint tolerance may
increase the steam release from the ruptured steam generator above
the Updated Final Safety Analysis Review (UFSAR) value by
approximately 2%, the steam generator tube rupture analysis
indicates that the calculated break flow is still less than the
value reported in the UFSAR. Therefore, the radiological analysis
indicates that the slight increase in the steam release is offset by
the decrease in the break flow such that the offsite radiation doses
are less than those reported in the UFSAR. The evaluation also
concluded that the existing mass releases used in the offsite dose
calculation for the remaining transients (i.e., steam line break,
rod ejection) are still applicable. Therefore, based on the above,
there is no increase in the dose releases.
Neither the mass and energy release to the containment following
a postulated LOCA, nor the analysis of containment response
following the LOCA credit the MSSVs in mitigating the consequences
of an accident. Therefore, changing the MSSV lift setpoint
tolerances would have no impact on the containment integrity
analysis. In addition, based on the conclusion of the transient
analysis, the change to the MSSV tolerance will not affect the
calculated steam line break mass and energy releases inside
containment.
Deleting the orifice size column from Table 3.7.1-2 has no
impact on previously evaluated accidents. There is no change to the
orifice size, which is stated in the UFSAR and incorporated as
needed in the accident analyses.
The proposed changes do not introduce any new equipment,
equipment modifications, or any new or different modes of plant
operation. The MSSVs are not precursors to any analyzed accident.
The proposed changes will not affect the operational characteristics
of any equipment or systems.
Therefore, these proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Deleting the text describing reactor coolant loops and steam
generators from TS 3.7.1.1 Action a. has no impact on plant
operation since the specific restrictions on the number of operating
reactor coolant loops and steam generators are provided in TS
3.4.1.1 and 3.4.1.2.
The proposed change to TS 3.7.1.1, Action a., to require the
final mode to be hot shutdown rather than cold shutdown is
consistent with the Applicability section of the specification,
which does not require the MSSVs to be operable in hot shutdown.
There are no credible transients requiring the MSSVs in Modes 4 and
5. The steam generators are not normally used for heat removal in
Modes 5 and 6, and thus cannot be overpressurized. NUREG-1431 does
not include requirements for the MSSVs to be operable in these
modes. The change will also eliminate the unnecessary transient that
had been imposed on the unit by forcing entry into cold shutdown.
The new Action b. for TS 3.7.1.1 and text changes to Action a.
clarify the shutdown requirement times based on the number of
inoperable valves. There are no changes to the times.
The proposed change to TSSR 4.7.1.1 to clarify that TSSR 4.0.4
does not apply for entry into Mode 3 will allow ComEd to continue to
perform MSSV testing at normal operating pressure and temperature as
required by the applicable codes. The change precludes having to
enter an action statement to perform the testing and eliminates
severe time restrictions on the valve testing and conflicts with
other plant startup requirements.
The proposed recalculated setpoints of Table 3.7-1 are more
limiting than those currently allowed in the Specification and
ensure that the original design margins for safety are met. The
secondary system pressure remains within design limits.
Increasing the as-found tolerance on the MSSV setpoint to plus/
minus 3% will not increase the challenge to the MSSVs or result in
increased actuation of the valves. The changes to the Bases document
the method for calculating the reduced reactor trip setpoints based
on reduced availability of MSSVs.
Deleting the orifice size column from Table 3.7-2 and the
obsolete one-time requirements in TSSR 4.7.1.1 are administrative
changes only.
Increasing the lift setpoint tolerance on the MSSVs does not
introduce a new accident initiator mechanism. The proposed change
does not introduce any new equipment, equipment modifications, or
any new or different modes of plant operation. No new failure modes
have been defined for any system or component important to safety
nor has any new limiting single failure been identified. This change
will not affect the operational characteristics of any equipment or
systems. Thus, there is no change in the margin for safety.
Therefore, these proposed changes will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
C. The proposed change does not involve a significant reduction
in a margin of safety.
Deleting the text describing reactor coolant loops and steam
generators has no impact on plant operation since the specific
restrictions on the number of operating reactor coolant loops and
steam generators are provided in TS 3.4.1.1 and 3.4.1.2.
The change requiring hot shutdown instead of cold shutdown entry
is more appropriate than the existing specification since the action
statement places the plant in a mode where operability of the MSSVs
is not required. The Technical Specification is applicable in Modes
1, 2, and 3, therefore, entering Mode 4 places the plant in a
condition where the MSSVs are not required to be operable. There are
no credible transients requiring the MSSVs in Modes 4 and 5. The
steam generators are not normally used for heat removal in Modes 5
and 6, and thus cannot be overpressurized. NUREG-1431 does not
include requirements for the MSSVs to be operable in these modes.
Changing the mode in which the MSSVs are tested will not change
the operational characteristics of the MSSVs. ComEd will continue to
test the MSSVs at normal operating pressure and temperature as
required by the applicable codes.
The proposed reactor trip setpoints in Table 3.7-1 are more
limiting than the current setpoints in the Specification. Reactor
trip settings were calculated using a revised methodology to account
for the non-linear relationship of reactor trip setpoints and
reduced MSSV availability. The revised setpoints ensure the original
design margin of safety is maintained. The proposed changes to the
Bases include the revised equation used to calculate the reduced
reactor trip setpoints.
Increasing the as-found lift setpoint tolerance on the MSSVs
will not adversely affect the operation of the reactor protection
system, any of the protection setpoints, or any other device
required for accident mitigation. The proposed increase in the
setpoint tolerance does not invalidate the LOCA and non-LOCA
conclusions presented in the UFSAR accident analyses. In letter CAE-
91-209/CAE 91-219, Westinghouse concluded that the new loss of load/
turbine trip analysis satisfied all applicable acceptance criteria
and demonstrated that the conclusion presented in the UFSAR remains
valid. For all the UFSAR non-LOCA transients, the departure from
nucleate boiling design basis, primary and secondary pressure
limits, and dose release limits continue to be met. Peak cladding
temperatures remain well below the limits specified in the 10 CFR
50.46.
[[Page 11488]]
Deleting the orifice size column from Table 3.7-2 and the
obsolete one-time requirements in TSSR 4.7.1.1 are administrative
changes.
The proposed changes do not introduce any new equipment,
equipment modifications, or any new or different modes of plant
operation. These changes will not affect the operational
characteristics of any equipment or systems. Therefore, no reduction
in the margin of safety will occur as a result of changes.
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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: 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.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of amendment request: August 23, 1996.
Description of amendment request: The proposed amendment would
revise the technical specifications to reflect the design lineup for
the Non-Accessible Area Exhaust Filter Plenum Ventilation System, and
to make provisions for the performance of maintenance and testing on
the system.
Basis for proposed no significant hazards consideration
determination: 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.
Non-Accessible Area Exhaust Filter Plenum Ventilation (VA)
System lineups are not considered as the precursors to any accident.
The additional provisions added to the action statement for TS 3.7.7
accommodates required maintenance and surveillance activities. No
new equipment is being installed and no existing equipment is being
modified. Thus, these proposed changes will not result in an
increase in the probability of occurrence of an accident previously
evaluated.
On the postulated Loss Of Coolant Accident (LOCA) with Loss Of
Offsite Power (LOOP), the operating plenum will either realign
immediately or following the re-energization of its ESF bus which
will occur within 10 seconds. Thus, there will always be at least
one plenum operating immediately during an accident. The emergency
procedures direct the realignment of the standby plenum. This
direction is contained in the Byron and Braidwood Emergency
Procedures (BEP/BwEP)-0, ``Reactor Trip or Safety Injection,'' and
is performed prior to conducting event diagnostic steps.
Filtration of the air from the Emergency Core Cooling System
(ECCS) equipment cubicles becomes critical when the ECCS pumps begin
pumping accident water from the containment recirculation sumps.
Prior to this the water flowing in these pumps is Refueling Water
Storage Tank (RWST) water. This swap over from the RWST to the
containment recirculation sump is expected to occur, at the
earliest, 11 minutes following accident initiation leaving time to
open the inlet damper on the standby VA plenum. Thus, since the
standby plenum can be realigned before filtration of the ECCS
equipment cubicle air is required, the Updated Final Safety Analysis
Report (UFSAR) assumptions, and offsite dose calculation assumptions
remain valid. There will be no significant change in the types or
significant increase in the amounts of any effluent that may be
released offsite, and there will be no significant increase in
individual or cumulative occupational radiation exposure.
Observations conducted on licensed operators undergoing simulator
training verified that the VA system is realigned well before the
swap-over to the containment recirculation sump under these
conditions. Therefore, these proposed changes will not result in a
significant increase in the consequences of an accident previously
evaluated.
A review of the Byron and Braidwood Probabilistic Risk
Assessment (PRA) shows that these proposed changes will have no
effect on either Core Damage Frequency (CDF) or Uncontrolled Release
Frequency (URF).
Therefore, these changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
These proposed changes continue to ensure that, following a
LOCA, the air being exhausted from the ECCS equipment rooms is
properly filtered before being released to the environment.
These changes will not result in the installation of any new
equipment or the modification of any existing equipment. No new
operating modes or system interfaces will be created. The VA system
will continue to operate as designed during normal and post accident
conditions. All of the accident analysis assumptions and conditions
will remain satisfied.
Thus this 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.
These proposed changes reflect the design lineup for the VA
system and provide action requirements to accommodate required
maintenance and surveillance testing. The VA system will continue to
ensure that following a LOCA, the air being exhausted from the ECCS
equipment rooms is properly filtered before being released to the
environment.
Filtration of the ECCS equipment cubicle air does not become
critical until the suction of the ECCS pumps is switched from the
RWST to the containment recirculation sumps. This is postulated to
occur, at the earliest, 11 minutes following accident initiation. On
the postulated LOCA with LOOP, at lease one VA plenum will be in
operation immediately and the emergency procedures direct the
realignment of the standby plenum well before the ECCS pump suction
swap-over. Observations conducted on licensed operators undergoing
simulator training have verified this fact. Therefore, these
proposed changes do not alter or affect any UFSAR or off-site dose
calculation assumptions, and the margin of safety is not reduced.
A review of the Byron and Braidwood PRA shows that these
proposed changes will have no effect on either CDF or URF.
No new equipment is being installed, and no existing equipment
is being modified. The VA system will continue to operate as
designed during normal and post accident conditions. All of the
accident analysis assumptions remain satisfied.
Therefore this proposed change does not involve a significant
reduction in a margin of safety.
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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: 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.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
[[Page 11489]]
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of amendment request: January 20, 1997.
Description of amendment request: The proposed amendment would
change Technical Specification Table 3.6-1 to reflect planned changes
in the plant configuration. As a result of the planned replacement of
the Westinghouse D4 steam generators at Byron, Unit 1, and Braidwood,
Unit 1, changes will be made to the containment isolation piping
arrangements at the penetrations associated with the Feedwater (FW) and
Auxiliary Feedwater (AF) systems. As a result of these changes, there
will be no split FW flow with the replacement steam generators. AF flow
will be fed into the main FW piping outside of containment and the
existing FW tempering penetration will be used for a new steam
generator recirculation system to be used during periods of extended
shutdown. Additionally, since the replacement steam generators use a
feedring design rather than a preheater design, the FW Isolation Bypass
line and associated containment isolation valves will no longer be
required. Table 3.6-1 of the Technical Specifications (TS) must be
updated to reflect these changes. These changes do not affect the
containment isolation capability originally designed to the criteria in
10 CFR 50, Appendix A, General Design Criteria (GDC) 54 through 57 as
reflected in the Byron/Braidwood Updated Final Safety Analysis Report
(UFSAR).
Basis for proposed no significant hazards consideration
determination: 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.
Technical Specification 3/4.6.3 establishes the operability
requirements for containment isolation valves as required by the
Byron and Braidwood Operating Licenses in compliance with General
Design Criteria 54 through 57 of Appendix A to 10 CFR 50. The
operability of the containment isolation valves ensure that the
containment atmosphere will be isolated from the outside environment
in the event of a release of radioactive material to the containment
atmosphere. Table 3.6-1 identifies these isolation valves and
captures relevant information to ensure these valves remain operable
under required conditions.
These proposed changes result in the elimination of the FW
Isolation Bypass isolation valves. These isolation valves are not
required with the replacement steam generator design. The remaining
isolation valves have not been altered in any way, only the piping
associated with them has been altered to the revised configuration.
These changes do not result in alteration of any containment
penetrations.
Failure of the piping between the isolation valve and the
containment penetration is considered as an accident initiator.
However, all piping changes between the isolation valve and the
containment penetrations meet the requirements of the original
design.
Therefore, since all original piping design criteria are met and
the actual number of containment isolation valves is reduced, the
proposed change does not involve a significant increase in the
probability of an accident previously evaluated.
Each penetration identified in the proposed change is associated
with a closed system inside containment and, as such, is provided
containment isolation in accordance with the applicable requirements
of GDC 54 through 57. There are four analyzed transients which take
credit for feedwater isolation and are, therefore, relevant to this
proposed change. These accidents are: (1) feedwater system
malfunctions that result in an increase in FW flow, (2) inadvertent
opening of a steam generator relief or safety valve, (3) steam
system piping failure, and (4) FW system pipe break. All operability
requirements for the affected containment isolation valves are
unaffected by this proposed change.
The containment isolation valves' functions, system operating
conditions, and accident responses are unchanged as a result of the
new configuration. Therefore, since all original design criteria are
met and each remaining isolation valve continues to provide the same
degree of containment isolation as the original design, the proposed
change does not involve a significant increase in the 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.
All modifications associated with the proposed changes will be
outside of containment and can be characterized as the rearrangement
of piping systems. All piping changes will comply with the original
design of the plant and will retain required containment isolation
capabilities per the requirements of GDC 54 through 57 as required
by the current design basis. Piping configurations within the area
of the containment penetration and the containment isolation valves
are required to minimize branch connections per guidance in the
Standard Review Plan (SRP) Section 3.6.2.
Therefore, since there are no unique configurations or
reductions in design requirements, this proposed change does not
create the possibility of any new or different kinds of accidents
from those previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes to the containment isolation arrangement
are being made consistent with the same codes, standards, and
isolation criteria as are currently in use at Byron and Braidwood.
The containment isolation valves remaining in place following the
steam generator replacement are unchanged with regard to their
function, capability, reliability, or physical requirements.
Containment isolation capability in accordance with GDC 54 through
57 is maintained at current levels of protection for the health and
safety of the general public. Therefore, this proposed change does
not involve a significant reduction in the margin of safety.
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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: 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.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: January 31, 1997.
Description of amendment request: The proposed amendments would
revise the maximum allowable value in the Byron, Unit 1, Technical
Specifications (TS), of the dose equivalent (DE) iodine-131
concentration in the primary coolant from the present value of 0.35
microcuries per gram of coolant to a maximum allowable of 0.20
microcuries per gram. This reduction in the DE iodine-131 concentration
would be applicable only for the remainder of the present Byron, Unit
1, operating cycle (i.e., fuel cycle 8) which the licensee has
previously stated will end in December 1997. The subject amendments are
proposed by the licensee in order to provide additional margin with
respect to the maximum Byron Station site allowable primary-to-
secondary leakage limit from the Byron, Unit 1, steam generators (SG).
This proposed Byron, Unit 1, TS revision to increase this margin is
being proposed in conjunction
[[Page 11490]]
with the proposed operating interval of 540 days above a Thot
temperature of 500 degrees Fahrenheit, between eddy current inspections
(ECI) of the Byron 1 SGs. The last Byron, Unit 1, ECI was initiated in
November 1995. This margin increase is being sought by the licensee to
address staff concerns regarding potential SG tube leakage under
postulated accident conditions due to SG tube circumferential cracking
at the top of the tubesheet in the roll transition zone.
While the proposed revision to the DE iodine-131 is applicable only
to Byron, Unit 1, the pending request for license amendments involves
both Byron, Units 1 and 2, in that both units have a common set of TSs.
Basis for proposed no significant hazards consideration
determination: 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.
Generic Letter 95-05, ``Voltage Based Repair Criteria For
Westinghouse Steam Generator Tubes Affected By Outside Diameter
Stress Corrosion Cracking,'' allows lowering of the RCS DE I-131
activity as a means for accepting higher projected leak rates if
justification for equivalent I-131 below 0.35 microcuries/gm is
provided. Four methods for determining the impact of a release of
activity to the public were reviewed to provide the justification.
They are as follows:
Method 1: NRC NUREG 0800, Standard Review Plan (SRP) Methodology
Method 2: Methodology described in a report by J.P. Adams and C.L.
Atwood, ``The Iodine Spike Release Rate During a Steam Generator
Tube Rupture,'' Nuclear Technology, Vol. 94 p. 361 (1991), using
Byron Station reactor trip data.
Method 3: Methodology described in Adams and Atwood report, using
normalized industry reactor trip data.
Method 4: Methodology described in draft EPRI Report TR-103680,
Revision 1, November 1995, ``Empirical Study of Iodine Spiking in
PWR Plants''.
The effect of reducing the RCS DE I-131 limit on the amount of
activity released to the environment remains unchanged when the
maximum site allowable primary-to-secondary leakage limit is
proportionately increased. With a DE I-131 limit of 1.0 microcuries/
gm, the maximum site allowable leakage limit was calculated in
accordance with the NRC SRP methodology to be 12.8 gpm. The
corresponding calculated activity released during a MSLB is 15.8 Ci.
ComEd has evaluated the reduction of the DE I-131 to 0.20
microcuries/gm along with the increase of the allowable leakage to
64 gpm and has concluded:
--The maximum activity released is not changed, and
--The offsite dose including the iodine spiking factor is bounded by
method 1.
Therefore, the offsite dose assessment and conclusions
previously reached remain valid and continue to meet the
requirements of 10 CFR 100.
An evaluation of Control Room dose attributed to a MSLB
concurrent with steam generator primary-to-secondary leakage at the
site allowable leakage limit was performed in support of a license
amendment request for application of 1.0 volt Interim Plugging
Criteria. This evaluation concluded that Control Room dose due to
the MSLB scenario is bounded by the existing loss of coolant
accident analysis. Therefore, the maximum site allowable primary-to-
secondary leakage limit continues to be based on offsite dose at the
Exclusion Area boundary due to MSLB leakage. This conclusion was
previously submitted to the Staff in a September 22, 1994,
transmittal in support of the 1.0 volt Interim Plugging Criteria
license amendment request.
Based on the NRC SRP methodology for dose assessments, the
Control Room dose, the Low Population Zone dose, and the dose at the
Exclusion Area Boundary continue to satisfy the appropriate fraction
of the 10CFR100 dose limits.
The Adams and Atwood report concluded that the NRC SRP
methodology, which specifies a release rate spike factor of 500 for
iodine activity from the fuel rod to the RCS, is conservative. In
order to justify that a release rate spike factor of 500 is
conservative, actual operating data from the previous reactor trips
of Byron Unit 1 and Unit 2, with and without fuel failures, were
reviewed and analyzed using the methodology presented Section II.C
of the Adams and Atwood report (Method 2). The same five data
screening criteria described in the Adams and Atwood report were
applied to the Byron data to ensure consistency and validity when
comparing the Byron results to the data in the Adams and Atwood
report. Of the twenty-eight (28) reactor trip events at Byron Units
1 and 2, twelve (12) met the five data screening criteria.
Three of the Byron trips occurred during cycles with no failed
fuel. In all three of these instances, the calculated spike factor
was less than the spike factor of 500 assumed in the NRC SRP
methodology. Byron, Unit 1, Cycle 8 is currently operating with no
failed fuel and a DE I-131 activity of approximately 6E-4
microcuries/gm. The three previous trips with no fuel failures had
steady-state iodine values that are relatively close to current
operating conditions. It is therefore reasonable to conclude that
the calculated spike factors from those trips would reflect the
spike factor expected from an actual trip during the current cycle.
Based on the data in the Adams and Atwood report, the NRC SRP
release rate spike factor of 500 may seem non-conservative since the
Adams and Atwood factor was typically greater than 500 when initial
concentrations were less than 0.3 microcuries/gm. The primary reason
for these high ratios (up to 12,000) is not because the absolute
post-trip release rate is high (factor numerator), but rather
because the steady-state release rate (factor denominator) is low.
The Byron specific data only resulted in one trip with a calculated
release rate spike factor greater than 500, a value of 603.9. The
trip occurred during the first operating cycle of Unit 2 which
experienced failed fuel and a very low steady-state release rate. It
is not expected based upon the current fuel cycle conditions that a
spiking factor of greater than 500 would occur.
In order to compare the Byron specific data to the NRC SRP
methodology, the release rate for a steady-state RCS DE I-131
activity of 1.0 microcuries/gm was calculated. Using the Byron
specific data, the steady-state release rate is 17.6 Ci/hr. Using a
release rate factor of 500 for the accident initiated spike, the
post-trip maximum release rate would be 8797 Ci/hr. This is
significantly higher than the largest iodine release rate of 127 Ci/
hr from the Byron data. This demonstrates that, although a data
point shows an iodine spike factor greater than 500, the resulting
post-trip RCS DE I-131 fuel rod iodine release rate is less than the
fuel rod iodine release rate from the NRC SRP methodology.
In the fourth method, the results from Draft EPRI Report TR-
103680, Rev. 1, November 1995, ``Empirical Study of Iodine Spiking
In PWR Power Plants'' were applied. The objective of the EPRI study
was to quantify the iodine spiking in postulated Main Steam Line
Break/Steam Generator Tube Rupture (MSLB/SGTR) sequences. In the
EPRI report, an iodine spike factor between 40 and 150 was
determined to match data from existing plant trips. The maximum
iodine spike factor value of 150 was applied to a steady-state
equilibrium RCS DE I-131 activity of 0.33 microcuries/gm. The
resulting 2-hour average iodine concentration for a postulated MSLB/
SGTR sequence was determined to be 3.1 microcuries/gm. Since the
EPRI report is based on industry data and the EPRI method predicted
a post-accident iodine activity which is a small fraction of the
activity predicted by the NRC SRP methodology, it can be expected
that, for the proposed 0.2 microcuries/gm limit under a MSLB/SGTR
sequence, the post-accident iodine activity would be a small
fraction of the RCS DE I-131 activity predicted by the NRC SRP
methodology.
Lowering the Unit 1 RCS DE I-131 activity limit is conservative
and remains bounded by the NRC SRP methodology. Thus, all offsite
and control room dose assessment conclusions satisfy the appropriate
limits of 10 CFR 100 and GDC 19. These proposed changes do not
result in a significant increase in the consequences of an accident
previously analyzed.
The RCS DE I-131 activity limit is not considered as a precursor
to any accident. Therefore, this proposed change does not result in
a significant increase in the probability of an accident previously
analyzed.
The correction of the typographical error is administrative in
nature and has no impact on either the probability or consequences
of an accident previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
[[Page 11491]]
The changes proposed in this amendment request conservatively
reduce the Unit 1 DE I-131 limit at which action needs to be taken
and correct a typographical error. The changes do not directly
affect plant operation. These changes will not result in the
installation of any new equipment or systems or the modification of
any existing equipment or systems. No new operating procedures,
conditions or modes will be created by this proposed amendment.
Thus, this 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.
NRC Generic Letter 95-05 allows lowering of the dose equivalent
iodine as a means for accepting higher projected leakage rates
provided justification for equivalent I-131 below 0.35 microcuries/
gm is provided. Four methods for determining the fuel rod iodine
release rates and spike factors during an accident were reviewed.
Each of these methods utilized actual industry data, including
Byron, Unit 1 and Unit 2, for pre-and post-reactor trip DE I-131
activities. Each of the methods demonstrated that the actual fuel
rod iodine release rates are a small fraction of the release rate as
calculated using the NRC SRP methodology. All design basis and off-
site dose calculation assumptions remain satisfied. This proposed
change will not result in a reduction in a margin of safety.
Correction of the typographical error is administrative in
nature and does not impact the margin of safety. Therefore, the
proposed changes do not result in a significant reduction in a
margin of safety.
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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Byron Public Library District,
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of amendment request: February 18, 1997.
Description of amendment request: The proposed amendment would
revise Byron and Braidwood Technical Specification (TS) Table 2.2-1
(functional unit 13.a), ``Reactor Trip System Instrumentation Trip
Setpoint: Steam Generator Water Level Low-Low'; TS Table 3.3-4
(functional unit 5.b.1), ``Engineered Safety Features Actuation System
Instrumentation Trip Setpoints: Steam Generator Water Level-High-High';
TS Table 3.3-4 (6.c.1), ``Engineered Safety Features Actuation System
Instrumentation Trip Setpoints: Steam Generator Water Level-Low-Low
Start Motor-Driven Pump and Diesel-Driven Pump'; TS Surveillance
Requirement (TSSR) 4.4.1.2.2, required steam generator inventory during
hot standby; TSSR 4.4.1.3.2, required steam generator inventory during
hot shutdown; and TS Section 3.4.1.4.1.b, limiting condition for
operation during cold shutdown with loops filled.
The installation of Babcock and Wilcox International (BWI),
replacement steam generators (RSGs) at Byron, Unit 1, and Braidwood,
Unit 1, necessitates an increase to the operating range of the steam
generators due to the decrease in narrow range span from 233 inches for
the original Westinghouse Model D4 steam generators (OSGs) to 180
inches for the BWI RSGs. The increase in operating range will minimize
the possibility of inadvertent plant trips following load changes and
feedwater transients.
ComEd also proposes to eliminate notations from page 2-5 for both
Braidwood and Byron and pages 3/4 3-25 and 3/4 3-26 (for Braidwood
only) since they are related to cycles already completed and,
therefore, are no longer valid.
Basis for proposed no significant hazards consideration
determination: 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.
This proposed change includes changing the low-low and high-high
SG level setpoints. The setpoints are being changed to increase the
SG level operating range. The change in acceptable operating range
will decrease the possibility of inadvertent plant trips following
load changes and feedwater transients. Therefore, the probability of
inadvertent plant trips will decrease with this change.
The minimum setpoint change proposed in this request establishes
controls to ensure that an adequate heat sink is maintained by
providing an adequate secondary liquid mass to remove primary system
sensible heat and core decay heat shortly after reactor trip and
initiating auxiliary feedwater flow for long-term cooling. The
accidents evaluated for this requirement are the Loss of Normal
Feedwater and Feedwater Line Break transients.
The maximum setpoint ensures the steam lines and turbine remain
undamaged from the introduction of low quality, two-phase flow from
the steam generators into the steam lines. The accident evaluated
for this requirement is the Feedwater System Malfunction that
results in an increase in feedwater to one or more steam generators.
The steam generator water level setpoints are not considered a
precursor to any of the analayzed accidents, and, therefore, these
proposed changes do not result in an increase in the probability of
occurrence of any accident previously analyzed.
The accidents evaluated for the low-low setpoint are the Loss of
Normal Feedwater and Feedwater Line Break transients. These
accidents were both analyzed using approved methodologies. All
acceptance criteria were shown to be met for both these events. In
addition, it was demonstrated that the Feedwater System Pipe Break
response with the RSGs and the proposed low-low setpoint were
bounded by the response with the original Model D4 steam generators.
Therefore, the proposed low-low level setpoint change is
demonstrated not to result in an increase in the consequences for
these accidents.
The accident evaluated for the high-high setpoint is the
Feedwater System Malfunction that results in an increase in
feedwater to one or more Steam Generators. All acceptance criteria
were shown to be met. In addition, it was shown that the RSGs do not
completely fill with liquid. This assures that the steam lines and
turbine remain undamaged with no introduction of low quality, two-
phase flow from the steam generators into the steam lines during the
transient. With all acceptance criteria met, the proposed high-high
level setpoint change is demonstrated not to result in an increase
in the consequences for these accidents.
TSSR 4.4.1.2.2, TSSR 4.4.1.3.2, and TS 3.4.1.4.1.b assure a
minimum inventory (i.e., level) to provide decay heat removal. The
requirement for a minimum inventory to remove decay heat is met with
assurance that the tube bundle is completely covered. The steam
generator operating water level during shutdown conditions are not
considered a precursor to any accident, and, therefore, these
proposed changes do not result in an increase in the probability of
occurrence of any accident previously analyzed.
The elimination of outdated cycle specific notations from page
2-5 for both Braidwood and Byron and pages 3/4 3-25 and 3/4 3-26
(Braidwood only) are only administrative and does not impact the
probability or consequences of any accidents previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed setpoint changes do not create any new operating
conditions or modes. The proposed change only revises the setpoints
for the Reactor Trip System and Engineered Safety Features Actuation
System. The actions of these systems will continue to be performed
in accordance with
[[Page 11492]]
existing requirements which are sufficient to ensure plant safety is
maintained.
Shutdown conditions steam generator water level is necessary to
assure adequate decay heat removal capacity. Assurance that the tube
bundle is completely covered along with existing technical
specification controls on the Auxiliary Feedwater System and on the
Condensate Storage Tank ensure adequate heat removal capacity is
maintained and that plant safety is maintained.
Thus, this proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The elimination of outdated cycle specific notations from page
2-5 for both Braidwood and Byron and pages 3/4 3-25 and 3/4 3-26
(Braidwood only) are only administrative and does not create the
possibility of a new or different accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
A safety evaluation was performed to determine the effect of the
RSGs with the revised setpoints.
The accidents potentially affected by the change in the Reactor
Trip Steam Generator Water Level low-low setpoint (TS 2.2.1, Table
2.2-1, functional unit 13.a) and Engineered Safety Features
Actuation System low-low AFW start setpoint (TS 3.3.2, Table 3.3-4,
functional unit 6.c.1) are the Loss of Normal Feedwater and
Feedwater Line Break transients. These accidents were both analyzed
using approved methodologies. All acceptance criteria were shown to
be met for both these events.
In addition, it was demonstrated that the Feedwater System Pipe
Break response with the RSGs with the proposed low-low setpoint were
bounded by the response with the OSGs. Therefore, the proposed low-
low level setpoint change is demonstrated not to result in an
reduction in the margin of safety for these accidents.
The accident potentially affected by the change in the
Engineered Safety Features Actuation System high-high SG level trip
(TS 3.3.2, Table 3.3-4, functional unit 5.b.1) is a Feedwater System
Malfunction that results in an increase in feedwater to one or more
steam generators. This accident was analyzed using an approved
methodology. In the evaluation of the Feedwater System Malfunction,
all acceptance criteria were shown to be met. In addition, it was
shown that the RSGs do not completely fill with liquid. This assures
that the steam lines and turbine remain undamaged with no
introduction of low quality, two-phase flow from the steam
generators into the steam lines during the transient. With all
acceptance criteria met, the proposed high-high level setpoint
change is demonstrated not to result in a reduction in the margin of
safety.
There are no design basis accidents involving shutdown condition
steam generator water level. Existing TS controls on the Auxiliary
Feedwater System and on the Condensate Storage Tank ensure adequate
heat removal capacity is maintained and that plant safety is
maintained during shutdown conditions. Therefore, a change to the
shutdown condition steam generator water level does not result in a
reduction in the margin of safety.
The elimination of outdated cycle specific notations from page
2-5 for both Braidwood and Byron and pages 3/4 3-25 and 3/4 3-26
(for Braidwood only) are only administrative and does not result in
a reduction in the margin of safety for any analyzed event.
Therefore, this amendment request does not result in a
significant decrease in a margin of safety.
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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: 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.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company,
Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power Plant,
Unit No. 1, Lake County, Ohio
Date of amendment request: January 31, 1997.
Description of amendment request: The proposed amendment will
insert, by general reference, in the Perry Nuclear Power Plant
Technical Specifications, the implementation document that the licensee
will use to implement Option B, ``Performance-Based Requirements,'' to
10 CFR 50, Appendix J, ``Primary Reactor Containment Leakage Testing
for Water-Cooled Power Reactors.'' Option B to 10 CFR 50 Appendix J is
an option that became effective on October 26, 1995.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The changes involved in this license amendment request revise
the criteria for determining the Containment leak rate testing
interval based upon past component performance. The revised criteria
are based on the guidance contained in Regulatory Guide 1.163,
``Performance-Based Containment Leak-Test Program.'' When the
containment or containment penetrations have performed
satisfactorily on a historical basis, this guidance permits the use
of extended testing frequencies.
Since the allowable leakage rates are not being affected, the
performance of the primary containment and systems and components
penetrating the primary containment remains within acceptable
limits. The functions and operation of these components will remain
unchanged. Since the components are utilized to mitigate the
consequences of accidents that require containment isolation, they
are not considered to be accident initiators. Additionally, there
are no accidents associated with implementation of a performance-
based testing frequency for the primary containment and systems and
components penetrating the primary containment.
As discussed previously, the components are utilized to mitigate
the consequences of accident scenarios which rely upon the primary
containment and systems and components penetrating the primary
containment, to prevent the release of radioactive effluents. The
implementation of Option B to 10 CFR 50 Appendix J is not intended
to provide relief from the leakage criteria. The components will
still be required to meet the leakage requirements as discussed in
USAR Section 6.2.6 and Technical Specifications 3.6.1.1, 3.6.1.2,
and 3.6.1.3. The primary containment isolation system is designed to
limit leakage to La, which is defined by the Perry Technical
Specifications to be 0.20 percent of primary containment air weight
per day at the calculated peak containment pressure (Pa) for
the design basis loss of coolant accident. The limitation on the
rate of primary containment leakage is designed to ensure that the
total leakage volume will not exceed the value assumed in the
accident analyses at Pa. The La value is not being
modified by this proposed change. Based on this, the primary
containment and system and components penetrating the primary
containment will remain capable of maintaining radioactive effluent
releases within the limits of 10 CFR 100.
Because the proposed change does not alter the plant design,
including the primary containment and primary containment
penetrations, the proposed change does not directly result in an
increase in primary containment leakage. Since the frequency will be
based on the performance of the subject components, only those
components that have satisfactorily maintained the actual leakage
less than the allowable leakage will be tested less frequently. The
testing frequency for components which have not satisfactorily
limited leakage, or have not performed satisfactorily in the past,
will not be altered. Other programs are also in place to ensure that
proper maintenance and repairs are performed during the service life
of the primary containment and systems and components penetrating
the primary containment.
[[Page 11493]]
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of previously evaluated
accidents.
Several administrative/editorial changes have been incorporated
(e.g., the clarification of the ``less than'' and ``less than or
equal to'' signs on the Technical Specification acceptance criteria,
and the retention of the standard frequency for the Drywell visual
inspections). Such administrative/editorial changes do not impact
initiators of analyzed events or assumed mitigation of accident or
transient events. Therefore, these changes also do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not involve a change to the plant
design or operation, or new system interfaces. Consequently, the
proposed change does not affect the parameters or conditions that
could contribute to initiation of accidents. This change involves
adopting a performance-based method for determining Type A, B, and C
test frequencies. Except for the method of defining the test
frequency, the methods for performing the actual tests are not
changed. No new accident modes would be created by extending testing
intervals. No safety related equipment or safety functions are
altered as a result of this change. The change in testing frequency
will not create any different types of accidents since the primary
containment and systems and components penetrating the primary
containment will continue to operate within their design bases.
Therefore, reducing the test frequency would have no influence on,
nor contribute to, the possibility of a new or different kind of
accident or malfunction from those previously analyzed.
Based on the above discussions, the proposed change would not
create the possibility of a new or different kind of accident than
those previously evaluated.
The proposed administrative/editorial changes do not involve a
physical alteration of the plant (no new or different type of
equipment will be installed) or changes in methods governing normal
plant operation. Thus, these changes also do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
This request does not involve a significant reduction in a
margin of safety. The proposed change adopts a performance-based
method for determining frequency of Type A, B, and C testing.
Except for the method of defining test frequency, no change in
the method of testing is proposed. Since the frequency will be based
on the performance of the subject components, only those components
that have satisfactorily maintained actual leakage less than the
allowable leakage will be tested less frequently. Other programs are
also in place to ensure that proper maintenance and repairs are
performed during the service life of the primary containment and
systems and components penetrating the primary containment.
The margin of safety associated with the proposed change
involves the offsite dose consequences of postulated accidents,
which are directly related to the rate of primary containment
leakage. The primary containment isolation system is designed to
limit leakage to La, which is defined by the Perry Technical
Specifications to be 0.20 percent of primary containment air weight
per day at the calculated peak containment pressure (Pa) for
the design basis loss of coolant accident. The limitation on the
rate of primary containment leakage is designed to ensure that the
total leakage volume will not exceed the value assumed in the
accident analyses at Pa. The margin of safety for the offsite
dose consequences of postulated accidents directly related to the
primary containment leakage rate is maintained by continuing to meet
La. The La value is not being modified by this proposed
change. Based on this, the primary containment and systems and
components penetrating the primary containment will remain capable
of maintaining radioactive effluent releases within the limits of 10
CFR 100.
Therefore, the changes associated with this license amendment
request do not involve a significant reduction in the margin of
safety.
The proposed administrative/editorial changes will not reduce
the margin of safety because they have no impact on safety analysis
assumptions. These changes do not involve questions regarding safety
issues, and therefore also do not involve a significant reduction in
a margin of safety.
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.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
Dairyland Power Cooperative (DPC), Docket No. 50-409, LaCrosse Boiling
Water Reactor (LACBWR), Vernon County, Wisconsin
Date of amendment request: April 10, 1996.
Description of amendment request: This is a corrected notice that
was first issued on August 1, 1996. The proposed amendment would update
the facility Possession Only License and Technical Specifications to
reflect the permanently shutdown and defueled condition of the plant.
The amendment would also serve to remove the fire protection
requirements, radiological effluent controls, quality assurance program
controls and administrative controls for the emergency and security
plans from the Technical Specifications to other inspectable and
enforceable documents.
Basis for proposed no significant hazards consideration
determination: 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:
DPC proposes to modify the LACBWR Technical Specifications to
more accurately reflect the permanently shutdown, defueled,
possession-only status of the facility.
Analysis of no significant hazards consideration:
1. The proposed changes do not create a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes delete system requirements that are no
longer necessary to prevent, or mitigate the consequences of, a
credible SAFSTOR accident as described in our current SAFSTOR
Accident Analysis.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are either administrative in nature or were
made based on the analysis of previously evaluated accident
scenarios. In no other way do they change the design or operation of
the facility and therefore do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed changes do not result in a significant reduction
in the margin of safety.
The changes incorporate into the proposed Technical
Specifications the margin of safety associated with the current
SAFSTOR accident analysis and thus don't involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: LaCrosse Public Library, 800
Main Street, LaCrosse, Wisconsin 54601.
Attorney for licensee: Wheeler, Van Sickle and Anderson, Suite 801,
25 West Main Street, Madison, Wisconsin 53703-3398.
NRC Project Director: Seymour H. Weiss.
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: February 10, 1997 (TSC 95-04).
Description of amendment request: The proposed changes would revise
the
[[Page 11494]]
Technical Specifications (TS) to reduce the allowable reactor building
volume leakage rate per-day limit to permit removal of consideration of
the penetration room contribution to the limit and the requirement to
maintain the penetration room at a negative pressure with respect to
all adjacent areas. Also, the penetration room ventilation system would
be removed from the description of the containment in TS 5.2, and a
surveillance requirement to perform a refueling outage test of the
penetration room ventilation system would be added to TS 4.5.4. In
addition, related changes would be made to the appropriate Bases
sections.
Basis for proposed no significant hazards consideration
determination: 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:
A. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No.
The following requirements are being removed from Technical
Specifications regarding the PRVS [Penetration Room Ventilation
System]:
(1) The requirement to measure reactor building leakage in
excess of 50% of the total allowed containment leakage to the
penetration room.
(2) The requirement, as specified in the design features, for
the PRVS to maintain the penetration room at a negative pressure
with respect to all adjacent areas. In addition, the design features
description for the PRVS will be completely removed from Technical
Specification 5.2 and replaced with a surveillance requirement in
Technical Specification 4.5.4.
To demonstrate the inconsequential effects of the removal of the
above requirements, a dose analysis was performed to conservatively
demonstrate that PRVS adds margin, but is not necessary to meet
10CFR100 limits. The analysis assumes that the PRVS is completely
unavailable for offsite dose reduction. However, the PRVS will be
available, and all of the relevant operability and surveillance
requirements for the PRVS will be retained in the Technical
Specifications. Therefore, it is highly unlikely that the actual
dose consequences would increase from 167 Rem thyroid to 240 Rem
thyroid, since all surveillance and operability requirements for
PRVS, other than the two requirements specified above, will be
retained in Technical Specifications.
The specified Technical Specification requirements for PRVS are
not accident initiators, nor will these requirements impact the
probability of an accident. The purpose of these requirements is to
ensure that the PRVS can reduce offsite dose to the public in the
event of an accident which results in radioactive effluents leaking
from the Reactor Building (RB) into the Penetration Room (PR).
In the initial ONS [Oconee Nuclear Station] design basis, the
PRVS was credited to reduce offsite dose to the public in the event
of certain accidents, such as a loss of coolant accident (LOCA) or
Maximum Hypothetical Accident (MHA), where there is airborne leakage
of radioactivity from the RB into the PR. The PRVS was credited to
reduce the MHA two-hour Exclusion Area Boundary (EAB) dose to less
than the 10CFR100 limit of 300 Rem thyroid. The current ONS dose
analysis, which takes credit for the PRVS, calculates the MHA two-
hour EAB dose to be 167 Rem thyroid. With a reduction in the
allowable leakage from the Reactor Building (La) from 0.25 w%/
day to 0.20 w%/day, while taking no credit for the PRVS, the two
hour EAB MHA dose is calculated to be 240 Rem thyroid. This new dose
analysis result meets the acceptance criterion of 10CFR100.
In addition to conducting a detailed dose analysis without
taking credit for PRVS, a detailed review of PRA [probabilistic risk
analysis] risk significance of the PRVS was conducted. The PRVS was
determined to have virtually no PRA risk significance and no
significant impact on consequences.
A review of the impact on control room habitability due to the
proposed Technical Specification changes was conducted for credible
UFSAR [Updated Final Safety Analysis Report] Chapter 15 accident
scenarios. The operability requirements of the PRVS which are being
retained in the Technical Specifications will ensure operability
requirements are met to support the Control Room Ventilation System
(CRVS). Therefore, removal of the identified statements pertaining
to PRVS operability from Technical Specifications will not
significantly impact control room habitability.
Based on the above information, the removal of the specified
requirements for PRVS from Technical Specifications will not
significantly increase the probability or consequences of an
accident previously evaluated. The original design basis for offsite
dose will still be met without any credit taken for the PRVS.
A change has been proposed to the Technical Specifications to
reduce the allowable leakage from the Reactor Building (La)
from 0.25 w%/day to 0.20 w%/day. This proposed change is
conservative in nature since it will result in a potential reduction
in the consequences of any accidents previously evaluated. Past
integrated leak rate tests (ILRTs) for all three Oconee units have
been reviewed by engineering and it has been concluded that this
reduction in allowable leakage will have no impact on future station
operation. This reduction is possible since the actual leakage of
the ONS reactor buildings is far less than the original allowable
design leakage.
B. Create the possibility of a new or different kind of accident
from the accident previously evaluated?
No.
As stated previously, the proposed Technical Specification
changes for the PRVS are not accident initiators, nor will these
changes create the possibility of new or different kinds of
accidents. The purpose of the PRVS is to reduce offsite dose to the
public in the event of an accident which results in leakage from the
RB into the PR.
Therefore, the proposed changes to the Technical Specifications
will not create the possibility of a new or different kind of
accident from the accidents previously evaluated.
C. Involve a significant reduction in a margin of safety?
No.
By reducing the allowable La to 0.20 w%/day, ONS meets
10CFR100 limits for off-site dose without taking any credit for the
PRVS.
Although the margin to 10CFR100 limits is reduced by not taking
credit for PRVS, it is concluded that the reduction in margin of
safety is insignificant because:
(1) PRVS operability and surveillance requirements are being
retained in Technical Specifications with the exception of two items
which do not significantly degrade the ability of PRVS to perform
its function.
(2) The reduction in the margin of safety is being offset by a
reduction in La.
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.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
GPU Nuclear, Inc. and Saxton Nuclear Experimental Corporation, Docket
No. 50-146, Saxton Nuclear Experimental Facility (SNEF), Bedford
County, Pennsylvania
Date of amendment request: November 25, 1996.
Description of amendment request: The proposed amendment would
allow decommissioning of the SNEF. The proposed changes to the license
and technical specifications (TSs) would (1) accommodate
decommissioning activities at the SNEF, (2) establish specific TS
controls such as administrative controls and inspection requirements
over decommissioning activities, (3) establish limiting conditions for
performing decommissioning activities, (4) extend exclusion area
controls to include the SNEF Decommissioning Support Building, (5)
establish requirements for a Radiological Environmental Monitoring
Program, an Off-Site Dose Calculation Manual and a Process Control
Program, and (6) establish requirements for Technical and Independent
Safety Reviews. In addition, the licensees have proposed other
administrative and editorial
[[Page 11495]]
changes to the TSs associated with the changes proposed above.
Basis for Proposed No Significant Hazards Consideration
Determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve a significant hazards
consideration because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Accidents which might occur during the active decommissioning
phase of the SNEF are bounded by the twelve accidents addressed in
section 3.0 of the Updated Safety Analysis Report (USAR). The
accident analyses addressed in the USAR demonstrate that no adverse
public health and safety impacts are expected from accidents that
might occur during decommissioning operations at the SNEF. The
highest calculated dose to an individual located at the site
boundary is less than 1.5 mrem to the whole body during a postulated
materials handling accident. The dose to an individual located at
the site boundary for other on-site accidents is at or below this
value. The limiting accident case represents less than 0.15% of the
EPA lower whole body dose limit for radiological accidents. Based on
the analyses of postulated credible accidents that might occur
during the planned decommissioning operations at the SNEF, it is
concluded that no significant increase in the probability or
consequences of an accident previously evaluated would be involved.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
There are three general categories of accidents. These scenarios
evaluate different methods of dispersing radioactive material to the
environment which include a loss of support systems and external
events. The first includes accident scenarios associated with
decommissioning tasks. These were identified and evaluated as
described in Section 3.0 of the USAR. The radiological effects of
these accident scenarios are discussed in item 1 above. They do not,
therefore, reflect a new or different kind of accident previously
evaluated. The second category, loss of support systems, does not
directly lead to an accident situation. Therefore, this category of
event does not create the possibility of a new or different kind of
accident. The final category of accidents involves external events.
Since these types of events can occur whether the SNEF is being
decommissioned or not, the act of decommissioning does not create
the possibility of a new or different kind of external event. Any
potential radiological hazard that may occur as a result of an
external event is addressed in item 1 above.
3. Involve a significant reduction in a margin of safety.
The TSs currently in place at the SNEF were developed to
maintain a shutdown facility in a secured condition with occasional
monitoring. These specifications were designed to ensure that the
approximately 4 megacuries of radioactive material left on site
following shutdown in 1972 as identified in the Saxton
Decommissioning Plan and Safety Analysis Report dated April 1972,
would remain safely contained. In the ensuing years, natural decay
of these radioactive materials has resulted in a remainder of
approximately 1500 curies of radioactive material at the facility
(93% of which is activation contained within the steel structures of
the reactor vessel). These proposed decommissioning TSs were
developed in order to ensure this remaining radioactive material is
safely contained and disposed of and that the environment
surrounding the facility is monitored. These actions will assure
that there is no reduction in the margin of safety during the active
decommissioning of the facility. The final result of these efforts
will be the removal of any potential radiological hazard from the
site and the release of the site for unrestricted use.
The NRC staff has reviewed the analysis of the licensees and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Saxton Community Library,
Front Street, Saxton, Pennsylvania 16678.
Attorney for the Licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts, and Trowbridge, 2300 N Street, N.W., Washington, D.C.
20037.
NRC Project Director: Seymour H. Weiss.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: January 28, 1997.
Description of amendment request: The proposed amendment would
relocate the details of Technical Specification (TS) Section 6.2.3 on
the Independent Safety Engineering Group (ISEG) from the Administration
Controls section of the TSs and place these details in the Updated
Final Safety Analysis Report (UFSAR) for South Texas Project, Units 1
and 2. This relocation is administrative only, and would not render any
changes to the existing plant philosophy toward the ISEG or any safety
analysis. Section 6.2.3 would be deleted from the TSs and removed from
the table of contents for Administrative Controls. Currently UFSAR
Section 13.4.2.2 describes the ISEG, but not in the detail as the
current TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes move details from the Technical
Specifications [TSs] to the Updated Final Safety Analysis Report
(UFSAR). The changes do not result in any hardware or operating
procedure changes. The details being removed from the Technical
Specifications [TSs] are not assumed to be an initiator of any
analyzed event. The UFSAR, which will contain the removed Technical
Specification [TS] details, will be maintained using the provisions
of 10 CFR 50.59 and is subject to the change control process in the
Administrative Controls Section of the Technical Specifications
[TSs]. [In addition] any changes to the UFSAR will be evaluated per
10 CFR 50.59, no increase in the probability or consequences of an
accident previously evaluated will be allowed without prior NRC
[Nuclear Regulatory Commission] approval. Therefore, the changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes move details from the technical
Specifications [TSs] to the Updated Final Safety Analysis Report
(UFSAR). The changes will not alter the plant configuration (no new
or different type of equipment will be installed) or make changes in
methods governing plant operation. The changes will not impose
different requirements, and adequate control of information will be
maintained. The changes will not alter assumptions made in the
safety analysis and licensing basis. Therefore, the changes will 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.
The proposed changes move detail from the Technical
Specifications [TSs] to the Updated Final Safety Analysis Report
(UFSAR). The changes do not reduce the margin of safety since the
relocation of details [is an administrative action and] has no
impact on any safety analysis assumptions. In addition, the detail
transposed from the Technical Specifications [TSs] to the UFSAR are
the same as the existing Technical Specification [TS] [6.2.3]. [In
addition] any future changes to the FSAR will be evaluated per the
requirements of 10 CFR 50.59, no reduction in a margin of safety
will be allowed without prior NRC approval. [Therefore, the licensee
concluded that the
[[Page 11496]]
changes will not involve a significant reduction in a margin of
safety.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: William D. Beckner.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: February 18, 1997.
Description of amendment request: The proposed amendment would
change the reactor core fuel assembly design features requirements
contained in Technical Specification 5.3.1, Fuel Assemblies. The
proposed change would allow for the limited replacement of failed or
damaged fuel rods in fuel assemblies with solid stainless steel or
zirconium alloy filler rods in accordance with NRC-approved
applications of fuel rod configurations. Reconstituted fuel assemblies
would be limited to those fuel designs that have been analyzed with
applicable NRC-staff-approved codes and methods and shown by tests or
analyses to comply with all fuel safety design bases. A limited number
of lead test assemblies that have not completed representative testing
would be allowed to be placed in nonlimiting core regions.
The proposed change would be in accordance with the guidance
provided in NRC Generic Letter 90-02, Supplement 1, issued July 31,
1992.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10 CFR
50.92(c)(1)) because the fuel assemblies would continue to meet the
same fuel assembly and fuel rod design bases as the current fuel
assemblies, the acceptance criteria for emergency core cooling systems
would continue to be satisfied for all fuel assemblies, there would be
no changes to reload design and safety analysis limits, and the
radiological consequences of accidents previously evaluated would
remain valid.
B. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because the fuel assemblies would continue to satisfy the
same design bases previously used. Since the original design criteria
would be met, no new accident initiators would be introduced. All
design and performance criteria would continue to be met for the use of
reconstituted assemblies containing the approved filler rods.
Furthermore, the use of reconstituted fuel assemblies does not affect
the manner by which the facility is operated.
C. The changes do not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)) because the core reload design and
safety analysis limits would be unchanged by the use of fuel assemblies
containing approved filler rods. The use of all fuel assemblies would
continue to be limited by the normal core operating conditions defined
in the Technical Specifications. Reconstituted fuel assemblies would be
evaluated specifically for each cycle reload core using approved reload
design methods and approved fuel rod design models and methods.
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.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270.
NRC Project Director: Patrick D. Milano.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: March 4, 1996.
Description of amendment request: The proposed amendment would
modify Surveillance Requirements 4.8.1.1.2.a.6, 4.8.1.1.2.b, and
4.8.1.1.2.g.7 by specifying load bands in loading the diesel generator
(DG) in lieu of the present requirement to load the DG greater than or
equal to a given value. A footnote is being added to the three
surveillance requirements to indicate that a momentary transient
outside the load range shall not invalidate the test. The associated
Bases sections have been revised to reflect the above changes.
Basis for proposed no significant hazards consideration
determination: 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:
NNECO has reviewed the proposed changes in accordance with 10
CFR 50.92 and has concluded that the changes do not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10 CFR 50.92(c) are not
compromised. The proposed changes do not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The purpose of the proposed changes to Surveillance Requirements
4.8.1.1.2.a.6, 4.8.1.1.2.b, and 4.8.1.1.2.g.7 is to provide the load
bands for loading the DG during the monthly, 184 days and 18-month
surveillances. Specifically, for monthly (Surveillance
4.8.1.1.2.a.6) and once per 184 days (Surveillance 4.8.1.1.2.b)
surveillances, the load band is between 4800-5000 kW. For the 18-
month surveillance (Surveillance 4.8.1.1.2.g.7), the load band is
between 5400-5500 kW during the first 2 hours and between 4800-5000
kW during the remaining 22 hours. The specified load bands account
for instrumentation inaccuracies using the plant computer and for
the operational control capabilities and human factor
characteristics. The proposed changes will keep the actual upper
load limit of the DG below the manufacturer's recommended limit and
the actual lower limit enveloping the accident load requirements.
The proposed changes will reduce unnecessary engine stress and wear,
while potentially improving overall diesel generator reliability and
availability. The changes to the Bases section reflect the changes
made to the surveillance requirements and, therefore, have no
adverse impact on plant safety. Since the proposed changes serve to
enhance overall safety, these changes do not increase the
probability or consequences of any accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes regarding the load band for the DGs do not
affect the operation or response of any plant equipment, including
the DG, or introduce any new failure mechanism. The proposed changes
will reduce unnecessary engine stress and wear, while potentially
improving overall DG reliability and availability. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
[[Page 11497]]
3. Involve a significant reduction in a margin of safety.
The proposed changes specifying the load bands for diesel
testing will keep the actual upper load limit of the DG below the
manufacturer's recommended limit, and the actual lower limit
enveloping the accident load requirements. Therefore, the proposed
changes do not affect the capability of the diesel to perform its
intended function. The purpose of these changes is to increase the
overall DG reliability. The proposed changes do not impact the
consequences of any design basis accidents. There is no direct
impact on any of the protective boundaries. For these reasons, the
changes do not involve a reduction in the margin of safety.
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.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: January 31, 1997.
Description of amendment request: The amendments would revise
Technical Specification 3/4.6.1.5, and its associated Bases section, to
ensure that a representative average containment air temperature is
measured.
Basis for proposed no significant hazards consideration
determination: 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.
Limitations on containment average air temperature ensure that
the overall containment average air temperature does not exceed the
initial temperature condition assumed in the accident analysis for a
Loss of Coolant Accident or Steamline Break inside Containment. The
resulting DBA temperature limits are used to established the
environmental qualification envelope for safety-related electrical
equipment inside containment.
The measurement of Containment average air temperature is a
means to ensure that the design temperature normal operating limit
is not exceeded. The probability of an accident is not impacted by
the surveillance of normal temperature as it is a measurement which
involves permanently installed, static equipment. The consequences
of an accident are not impacted since the method of measurement
ensures that the design basis temperatures are maintained and the
intent of the existing surveillance specification is not changed.
The proposed change does not impact the actual containment
temperature, but specifies an acceptably accurate method for its
determination.
Therefore, the probability of and consequences of an accident
previously evaluated are not significantly increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve any modifications to
existing plant equipment, do not alter the function of any plant
systems within Containment, do not introduce any new operating
configurations or new modes of plant operation, nor change the
safety analyses. The proposed change is consistent with NUREG-1431
and provides a methodology to ensure that calculated temperature is
accurately determined.
The proposed changes will, therefore, 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.
The proposed change results in an acceptably accurate
determination of the containment average air temperature, therefore,
compliance with the TS surveillance and its associated basis is
assured. The present margin of safety is not affected since
operating parameters and conditions are unchanged.
All changes are consistent with the intent of Salem's current TS
and with the surveillance specified in NUREG-1431, Revision 1.
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.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: February 11, 1997.
Description of amendment request: The amendments would add a new
Technical Specification 3/4.7.10, ``Chilled Water System'' to address
the support function this system provides to other necessary safety
systems.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Chilled Water System is a support system providing cooling
to the Relay Rooms, the Control Room, and the affected Electrical
Equipment Rooms. The Chilled Water System is not an accident
initiator of any accident evaluated in the Safety Analysis Report.
No physical changes to the Chilled Water System result from the
proposed TS. The specified Allowed Outage Times in the TS are
commensurate with the safety significance of the Chilled Water
System as demonstrated by the PSA analysis.
Therefore, the proposed TS does not significantly increase 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 changes do not involve any modifications to the
Chilled Water System or mode of operation of the system. The
proposed TS specifies the minimum operable number of chillers and
chilled water pumps to assure that the system performs its design
function. It does not change the basic way in which the Chilled
Water System is operated. The loads that are isolated are non-safety
loads. By maintaining the minimum operable number of chillers and
chilled water pumps, adequate cooling is assured to the Relay Rooms,
the Control Room, the affected Electrical Equipment Rooms.
Therefore, the change will 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.
The Chilled Water System is a support system which provides
cooling to the Relay Rooms, the Control Room, and the affected
Electrical Equipment Rooms. The proposed changes do not involve any
modifications to the Chilled Water System or changes to the mode of
operation of the system. The proposed TS establishes controls to
better ensure that the Chilled Water System will be able to perform
its intended design function
[[Page 11498]]
and ensures that the safety functions of supported systems are
maintained.
The proposed changes establish Allowed Outage Times and do not
affect the operation of the Chilled Water System, and thus do not
involve a reduction in a margin of safety.
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.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Project Director: John F. Stolz.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: January 20, 1997.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 3/4.5.2, ``Emergency Core
Cooling Systems, ECCS Subsystems--T avg 280 deg.F,''
TS Section 3/4.5.3, ``Emergency Core Cooling Systems, ECCS Subsystems--
Tavg < 280 deg.F,'' and TS Section 3/4.7, ``Plant Systems.''
Several surveillance intervals would be changed from 18 months to once
each refueling interval.
Basis for proposed no significant hazards consideration
determination: 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:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in
accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no such accidents are affected
by the proposed revisions to increase the surveillance test
intervals from 18 to 24 months for the ECCS Subsystems (Surveillance
Requirements 4.5.2.d.2.a, 4.5.2.e, 4.5.2.g.2, and 4.5.3), Auxiliary
Feedwater System (Surveillance Requirement 4.7.1.2.1.c), Motor
Driven Feedwater Pump System (Surveillance Requirement 4.7.1.7.d),
Component Cooling Water System (Surveillance Requirement 4.7.3.1.b)
and Service Water System (Surveillance Requirement 4.7.4.1.b).
Initiating conditions and assumptions remain as previously analyzed
for accidents in the DBNPS Updated Safety Analysis Report.
These revisions do not involve any physical changes to systems
or components, nor do they alter the typical manner in which the
systems or components are operated.
A review of historical 18 month surveillance data and
maintenance records support an increase in the surveillance test
intervals from 18 to 24 months (and up to 30 months on a non-routine
basis) because no potential for a significant increase in a failure
rate of an affected system or component was identified during these
reviews.
These proposed revisions are consistent with the NRC guidance on
evaluating and proposing such revisions as provided in Generic
Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2,
1991.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the source term, containment
isolation or radiological releases are not being changed by these
proposed revisions. Existing system and component redundancy is not
being changed by these proposed changes. Existing system and
component operation is not being changed by these proposed changes.
The assumptions used in evaluating the radiological consequences in
the DBNPS Updated Safety Analysis Report are not invalidated.
A review of historical 18 month surveillance data and
maintenance records support an increase in the surveillance test
intervals from 18 to 24 months (and up to 30 months on a non-routine
basis) because no potential for a significant increase in a failure
rate of an affected system or component was identified during these
reviews.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because these
revisions do not involve any physical changes to systems or
components, nor do they alter the typical manner in which the
systems or components are operated. A review of historical 18 month
surveillance data and maintenance records support an increase in the
surveillance test intervals from 18 to 24 months (and up to 30
months on a non-routine basis) because no potential for a
significant increase in a failure rate of a system or component was
identified during these reviews. No changes are being proposed to
the type of testing currently being performed, only to the length of
the surveillance test interval.
3. Not involve a significant reduction in a margin of safety
because a review of the historical 18 month surveillance data and
maintenance records identified no potential for a significant
increase in a failure rate of a system or component due to
increasing the surveillance test interval to 24 months. Existing
system and component redundancy is not being changed by these
proposed changes.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences,
therefore, there are no significant reductions in a margin of
safety.
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.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: January 30, 1997.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 2.2, ``Limiting Safety
System Settings,'' and applicable bases, TS Section 3/4.3,
``Instrumentation,'' and applicable bases, TS Section 3/4.4, ``Reactor
Coolant System,'' and TS Section 3/4.7, ``Plant Systems.'' Several
surveillance intervals would be changed from 18 months to once each
refueling interval. In addition, several setpoints would be revised
based on an instrument drift study, and trip setpoints would be revised
based on new calculations. Administrative revisions are also proposed
consistent with these changes.
Basis for proposed no significant hazards consideration
determination: 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:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in
accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no such accidents are affected
by the proposed revisions to increase the surveillance test
intervals from 18 to 24 months for the subject Technical
Specifications (TS): TS 2.2 Limiting Safety System Settings; TS 3/
4.3.1.1, Reactor Protection System Instrumentation; TS 3/4.3.2.2,
Steam and Feedwater Rupture Control System Instrumentation; TS 3/
4.3.3.5.1, Remote Shutdown Instrumentation;
[[Page 11499]]
TS 3/4.3.3.6, Post-Accident Monitoring Instrumentation; TS 3/4.4.3,
Safety Valves and Pilot Operated Relief Valve--Operating; TS 3/
4.4.6.1, Reactor Coolant System Leakage Detection Systems; TS 3/
4.7.1.2 and Auxiliary Feedwater System. Initiating conditions and
assumptions remain as previously analyzed for accidents in the DBNPS
Updated Safety Analysis Report.
Results of the instrument drift study analysis and review of
historical 18 month surveillance data and maintenance records
support an increase in the surveillance test intervals from 18 to 24
months (and up to 30 months on a non-routine basis) because: the
projected instrument errors caused by drift are bounded by the
existing setpoint analysis or either a new analysis has been
performed incorporating a more conservative setpoint or the
calculations excess margin was reduced; projected instrument errors
caused by drift are acceptable for control of plant parameters to
effect a safe shutdown with the associated instrumentation or an
engineering evaluation has been performed to justify continued use
of the instrument string and revisions will be made to DBNPS
calculations and controlling procedures where appropriate, to offset
any adverse effect; and no potential for a significant increase in a
failure rate of a system or component was identified during
surveillance data and maintenance records reviews.
These proposed revisions are consistent with the NRC guidance on
evaluating and proposing such revisions as provided in Generic
Letter 91-04, ``Changes in Technical Specification Surveillance
Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2,
1991.
The proposed revisions to Allowable Values for Steam and
Feedwater Rupture Control System Steam Generator Level--Low are
conservative with respect to the current Allowable Values and
therefore, do not adversely affect previously analyzed accidents.
The application of the Allowable Value to the Channel Functional
Test only, the proposed deletion of the Trip Setpoint, and revision
of the Limiting Condition for Operation and Action Statement A for
TS 3.3.2.2, SFRCS Instrumentation, associated with the proposed
revision of the Allowable Values for SFRCS Steam Generator Level--
Low are consistent with NUREG-1430, Revision 1, ``Standard Technical
Specifications, Babcock and Wilcox Plants,'' dated April, 1995. The
proposed revisions will have no adverse effect on any previously
analyzed accident.
The proposed revision to the Reactor Protection System High Flux
Allowable Value was determined in accordance with the approved
setpoint methodology described in Babcock and Wilcox document BAW-
10179P, Safety Criteria for Acceptable Cycle Reload Analyses, and is
bounded by the High Flux trip of 112% rated power assumed in the
DBNPS accident analysis.
The proposed deletion of the Trip Setpoints, deletion of the
Allowable Values applicable to the Channel Calibration for RC low
pressure, and RC high pressure functional units, application of
Allowable Values to the Channel Functional Test as opposed to the
Channel Calibration, and deletion of the ``**'' and ``#'' footnotes
for Technical Specification Table 2.2-1, Reactor Protection System
Instrumentation Trip Setpoints, and the proposed revision to TS 2.2,
Limiting Safety System Settings, are consistent with NUREG-1430,
Revision 1, ``Standard Technical Specifications, Babcock and Wilcox
Plants,'' dated April, 1995. The proposed revisions have no adverse
effect on any previously analyzed accident.
The proposed revision to Technical Specification Table 4.3-10,
Post-Accident Monitoring Instrumentation Surveillance Requirements,
Instrument 6, Containment Vessel Post-Accident Radiation separates
the radiation monitors to reflect the revision to 24 month
surveillance intervals for the High Range Radiation Monitors and
that the Containment Wide Range Noble Gas monitors will remain on a
18 month surveillance frequency is an administrative change and does
not affect previously analyzed accidents.
The proposed revision to the Technical Specification Bases
2.2.1, Reactor Protection System Instrumentation Setpoints, and
Bases 3/4.3.1 and 3/4.3.2, Reactor Protection System and Safety
System Instrumentation, are administrative and do not affect
previously analyzed accidents.
Initiating conditions and assumptions remain as previously
analyzed for accidents in the DBNPS Updated Safety Analysis Report.
These revisions do not involve any physical changes to systems
or components, nor do they alter the typical manner in which the
systems or components are operated.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the source term, containment
isolation or radiological releases are not being changed by these
proposed revisions. Existing system and component redundancy is not
being changed by these proposed changes. Existing system and
component operation is not being changed by these proposed changes
and the assumptions used in evaluating the radiological consequences
in the DBNPS Updated Safety Analysis Report are not invalidated.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because these
proposed revisions do not involve any physical changes to systems or
components, nor do they alter the typical manner in which the
systems or components are operated.
No changes are being proposed to the type of testing currently
being performed, only to the length of the surveillance test
interval.
Results of the instrument drift study analysis and review of
historical 18 month surveillance data and maintenance records
support an increase in the surveillance test intervals from 18 to 24
months (and up to 30 months on a non-routine basis) because: the
projected instrument errors caused by drift are bounded by the
existing setpoint analysis or either a new analysis has been
performed incorporating a more conservative setpoint or the
calculations excess margin was reduced; projected instrument errors
caused by drift are acceptable for control of plant parameters to
effect a safe shutdown with the associated instrumentation or an
engineering evaluation has been performed to justify continued use
of the instrument string and revisions will be made to DBNPS
calculations and controlling procedures where appropriate, to offset
any adverse effect; and no potential for a significant increase in a
failure rate of a system or component was identified during
surveillance data and maintenance records reviews.
The proposed revisions to Allowable Values for Steam and
Feedwater Rupture Control System Steam Generator Level--Low are
conservative with respect to the current Allowable Values and do not
alter any testing currently being performed.
The application of the Allowable Value to the Channel Functional
Test only, the proposed deletion of the Trip Setpoint, and revision
of the Limiting Condition for Operation and Action Statement A for
TS 3.3.2.2, SFRCS Instrumentation, associated with the proposed
revision to the Allowable Values for SFRCS Steam Generator Level--
Low are consistent with NUREG-1430, Revision 1, ``Standard Technical
Specifications, Babcock and Wilcox Plants,'' dated April, 1995. The
proposed revisions do not alter any testing currently being
performed.
The proposed deletion of the Trip Setpoints, deletion of the
Allowable Values applicable to the Channel Calibration for RC
lowpressure, and RC high pressure functional units, application of
Allowable Values to the Channel Functional Test as opposed to the
Channel Calibration, and deletion of the ``**'' and ``''
footnotes for Technical Specification Table 2.2-1, Reactor
Protection System Instrumentation Trip Setpoints, and the proposed
revision to TS 2.2, Limiting Safety System Settings, are consistent
with NUREG-1430, Revision 1, ``Standard Technical Specifications,
Babcock and Wilcox Plants,'' dated April, 1995. The proposed
revisions do not alter any testing currently being performed.
The proposed revision to the Reactor Protection System High Flux
Allowable Value was determined in accordance with the approved
setpoint methodology described in Babcock and Wilcox document BAW-
10179P, Safety Criteria for Acceptable Cycle Reload Analyses, and is
bounded by the High Flux trip of 112% rated power assumed in the
DBNPS accident analysis and does not alter any testing currently
being performed.
The proposed revision to Technical Specification Table 4.3-10,
Post-Accident Monitoring Instrumentation Surveillance Requirements,
Instrument 6, Containment Vessel Post-Accident Radiation separates
the radiation monitors to reflect the revision to 24 month
surveillance intervals for the High Range Radiation Monitors and
that the Containment Wide Range Noble Gas monitors will remain on a
18 month surveillance frequency is an administrative change and does
not alter any testing currently being performed.
The proposed revision to the Technical Specification Bases
2.2.1, Reactor Protection System Instrumentation Setpoints, and
Bases 3/4.3.1 and 3/4.3.2, Reactor Protection System and Safety
System Instrumentation,
[[Page 11500]]
are administrative and do not alter any testing currently being
performed.
3. Not involve a significant reduction in a margin of safety
because The results of the instrument drift study analysis and
review of historical 18 month surveillance data and maintenance
records support an increase in the surveillance test intervals from
18 to 24 months (and up to 30 months on a non-routine basis)
because: the projected instrument errors caused by drift are bounded
by the existing setpoint analysis or either a new analysis has been
performed incorporating a more conservative setpoint or the
calculations excess margin was reduced; projected instrument errors
caused by drift are acceptable for control of plant parameters to
effect a safe shutdown with the associated instrumentation or an
engineering evaluation has been performed to justify continued use
of the instrument string and revisions will be made to DBNPS
calculations and controlling procedures where appropriate, to offset
any adverse effect; and no potential for a significant increase in a
failure rate of a system or component was identified during
surveillance data and maintenance records reviews. Existing system
and component redundancy is not being changed by these proposed
changes.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences,
consequently there are no significant reductions in a margin of
safety.
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.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of amendment request: December 21, 1995, as supplemented on
October 24, 1996.
Description of amendment request: The proposed amendments would
relocate certain cycle-specific parameter limits from the Technical
Specifications to the Operating Limits Report (ORL).
Date of publication of individual notice in Federal Register:
February 20, 1997 (62 FR 7804).
Expiration date of individual notice: March 24, 1997.
Local Public Document Room location: 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.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of amendment request: November 5, 1996.
Description of amendment request: The proposed amendments would
revise the technical specifications to allow ComEd to take credit, on a
temporary basis, for soluble boron in the spent fuel storage water in
maintaining an acceptable margin of subcriticality.
Date of publication of individual notice in Federal Register:
February 10, 1997 (62 FR 6016).
Expiration date of individual notice: March 12, 1997.
Local Public Document Room location: 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.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 1 and 2, Grundy County, Illinois
Date of amendment request: February 17, 1997.
Description of amendment request: The amendments would increase the
maximum allowable water temperature for the Containment Cooling Service
Water inlet and the Suppression Pool.
Date of publication of individual notice in Federal Register:
February 27, 1997 (62 FR 8998).
Expiration date of individual notice: March 31, 1997.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document
[[Page 11501]]
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document rooms for the particular facilities involved.
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of application for amendment: October 3, 1996.
Brief description of amendment: The amendment concerns the
provisions at Calvert Cliffs Unit 1 for receiving, possessing, and
using byproduct, source, and special nuclear material. The amendment
changed the Unit 1 license, which previously contained restrictions on
the possession and use of byproduct, source, or special nuclear
material, to be consistent with the Unit 2 license, which has no such
restrictions. The staff found this license amendment to be acceptable
since both units share the same radiation protection staff, and the
training and procedures used to control the acceptance and use of
radioactive material at Unit 2 are sufficient to control the
radioactive material at Unit 1, as well.
Date of issuance: February 19, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 220.
Facility Operating License No. DPR-53: Amendment revised the
Operating License.
Date of initial notice in Federal Register: November 6, 1996 (61 FR
57482). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application for amendments: February 20, 1996 as
supplemented October 16, 1996.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3.1.5, TS 3.1.10 and TS 4.1 to: (1) reduce the
surveillance frequency for the boron concentration in the concentrated
boric acid storage tank; (2) delete the surveillance requirements for
Sr89 and Sr90, gross beta activity, gross alpha activity and
dissolved gas concentration in the reactor coolant, and gross beta
activity in the steam generator feedwater; (3) relocate the
surveillance requirements for tritium, chloride, fluoride, and oxygen
in the reactor coolant to the Selected Licensee Commitment (SLC)
manual; and (4) delete TS 3.1.10 related to temperature and pressure
requirements to avoid gas bubble formation on depressurization.
Date of issuance: February 19, 1997.
Effective date: As of the date of issuance to be implemented within
30 days. Implementation shall include concurrent revision of the
Selected Licensee Commitment Manual in accordance with the application
of this amendment.
Amendment Nos.: 221, 221, 218.
Facility Operating License Nos. DPR-38, DPR-47 and DPR-55:
Amendments revise the Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13523). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 19, 1997.
No significant hazards consideration comments received: No
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, Michigan
Date of application for amendments: February 26, 1996.
Brief description of amendments: The amendments revise the TS to
allow an increased limit for the nominal enrichment of new
(unirradiated) Westinghouse-fabricated fuel stored in the new fuel
storage racks.
Date of issuance: February 27, 1997.
Effective date: February 27, 1997, with full implementation within
45 days.
Amendment Nos.: 213 and 198.
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18172) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 27, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 4, 1996 as supplemented by letter
dated January 8, 1997.
Description of amendment request: The amendment revises Seabrook
Appendix A Technical Specifications (TS) 1.7, ``Containment
Integrity'', 3/4.6.1, ``Primary Containment'', and 3/4.6.5,
``Containment Enclosure Building'', to incorporate the provisions of
Option B to 10 CFR Part 50, Appendix J. TS Section 6.15, ``Containment
Leakage Rate Testing Program'', has been added to establish a
Containment Leakage Rate Testing Program, as specified in Regulatory
Guide 1.163, dated September 1995, to support these changes. In
addition to the changes to incorporate the provisions of Option B, TS
3.6.1.7 and 4.6.1.7.1 have been revised to incorporate an increased
leak testing interval and to include reference to the Containment
Leakage Rate Testing Program.
Date of issuance: February 24, 1997.
Effective date: February 24, 1997.
Amendment No.: 49.
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 28, 1996 (61 FR
44359). The licensee's letter dated January 8, 1997, which provided
additional information relating to containment purge supply and exhaust
valve testing and maintenance, does not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: July 18, 1995.
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to extend the surveillance schedule from 18 months
to each refueling interval (nominally 24 months) for TS 3/4.4.4,
``Relief Valves;'' TS 3/4.4.6.1, ``Reactor Coolant System
[[Page 11502]]
Leakage;'' TS 3/4.4.6.2, ``Operational Leakage;'' TS 3/4.4.9.3,
``Overpressure Protection Systems;'' and TS 3/4.4.11, ``Reactor Coolant
System Vents.''
Date of issuance: February 19, 1997.
Effective date: As of the date of issuance, to be implemented
within 90 days.
Amendment No.: 133.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58402).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: October 25, 1996.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) to incorporate the requirements of 10
CFR Part 50, Appendix J, Option B, for containment leakage tests. In
addition, the amendments add a new section to the TSs, which
establishes the requirements of the containment leakage rate testing
program, consistent with the Improved Standard Technical
Specifications.
Date of issuance: February 19, 1997.
Effective date: February 19, 1997, with full implementation within
30 days.
Amendment Nos.: 126 and 118.
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 15, 1997 (62 FR
2191) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 16, 1995, as supplemented by
letter dated August 8, 1996.
Brief description of amendment: The amendment revises the technical
specifications to add a limiting condition for operation and
surveillance test for safety related inverters and deletes the
nonsafety related instrument buses.
Date of issuance: February 13, 1997.
Effective date: February 13, 1997, to be implemented within 60 days
from the date of issuance.
Amendment No.: 180.
Facility Operating License No. DPR-40. Amendment revised the
Technical
Specifications.
Date of initial notice in Federal Register: March 13, 1996 (61 FR
10395)
The August 8, 1996, supplemental letter provided additional
clarifying information and did not change the initial no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 13, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: August 27, 1996.
Brief description of amendments: The proposed amendments change the
minimum allowable charging water header pressure from a value of 955
psig to a value of 940 psig in Technical Specification 3.10.8,
``Shutdown Margin (SDM) Test-Refueling.''
Date of issuance: February 19, 1997.
Effective date: Both units, as of date of issuance, to be
implemented within 30 days.
Amendments Nos.: 218 and 221.
Facility Operating License Nos. DPR-44 and DPR-56: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 23, 1996 (61 FR
55036)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: February 2, 1996, as
supplemented September 23, 1996.
Brief description of amendments: These amendments change Technical
Specification 3.6.1.2 for each unit to permit primary containment
leakage testing of the main steamline isolation valves at either 22.5
psig or 45 psig according to the type of test to be conducted.
Date of issuance: February 25, 1997.
Effective date: Both units, as of date of issuance, to be
implemented within 30 days.
Amendment Nos.: 163 and 134.
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 14, 1996 (61 FR
42282). The September 23, 1996, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 25, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Southern California Edison Company, et al., Docket No. 50-362, San
Onofre Nuclear Generating Station, Unit No. 3, San Diego County,
California
Date of application for amendment: January 14, 1997.
Brief description of amendment: The amendment revises Surveillance
Requirements (SRs) 3.8.1.14 and 3.8.1.15 to temporarily restore
provisions of the emergency diesel generator surveillance requirements
as they were prior to their revision as part of NRC Amendment No. 116
(conversion to the Improved Technical Specifications).
[[Page 11503]]
Date of issuance: February 10, 1997.
Effective date: February 10, 1997.
Amendment Nos.: 125.
Facility Operating License Nos. NPF-15: The amendments revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (62 FR 3536 dated January 23, 1997). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by February 24, 1997, but indicated that if the Commission
makes a final no significant hazards consideration determination any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 1997.
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: September 19, 1996,
supplemented on November 18, 1996, revised on January 13, 1997, and
supplemented on January 27, 1997.
Brief description of amendments: These amendments revise the
reactor coolant system temperature below which the low temperature
overpressure protection (LTOP) system and pressurizer power-operated
relief valves (PORVs) shall be operable, modify the requirement to
limit operation of the high pressure safety injection pump from reactor
coolant system cold leg temperature of less than or equal to 275 deg.F
to whenever the LTOP is required to be operable, change the name of the
system from the overpressure mitigation system to the LTOP system, and
revise the PORV setpoint from 425 psig to 440 psig.
Date of issuance: February 20, 1997, with full implementation
within 45 days.
Effective date: February 20, 1997.
Amendment Nos.: 172 and 176.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (62 FR 5256, dated February 4, 1997) The
notice provided an opportunity to submit comments on the Commission's
proposed NSHC determination. No comments have been received. The notice
also provided for an opportunity to request a hearing by March 6, 1997,
but indicated that if the Commission makes a final NSHC determination,
any such hearing would take place after issuance of the amendments. The
Commission's related evaluation of the amendments, finding of exigent
circumstances, and final determination of no significant hazards
considerations are contained in a Safety Evaluation dated February 20,
1997.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 13, 1995, as supplemented by
letter dated October 10, 1996.
Brief description of amendment: The amendment revises the 125-volt
D.C. Sources (3.8.2.1 and 3.8.2.2) and Onsite Power Distribution
(3.8.3.1 and 3.8.3.2) Technical Specifications to include provisions
for installed spare battery chargers, which will be added to the plant
design before startup from the ninth refueling outage.
Date of issuance: February 10, 1997.
Effective date: February 10, 1997, to be implemented before startup
from the ninth refueling outage, currently scheduled to begin in
September 1997.
Amendment No.: 104.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1639) The October 10, 1996, supplemental letter provided additional
clarifying information and did not change the initial no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Notice of Issuance of Amendment to Facility Operating License and Final
No Significant Hazards Consideration Determination
During the period since publication of the last biweekly notice,
individual notices of issuance of amendments have been issued for the
facilities as listed below. These notices were previously published as
separate individual notices. They are repeated here because this
biweekly notice lists all amendments that have been issued for which
the Commission has made a final determination that an amendment
involves no significant hazards consideration.
In this case, a prior Notice of Consideration of Issuance of
Amendment, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing was issued, a hearing was requested, and
the amendment was issued before any hearing because the Commission made
a final determination that the amendment involves no significant
hazards consideration.
Details are contained in the individual notice as cited.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of
[[Page 11504]]
Issuance of Amendment, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
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 room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By April 11, 1997, 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 for the
particular facility involved. 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 a 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 a 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. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
[[Page 11505]]
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 20555-001, 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, 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 (Project Director): 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 20555-001, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the 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).
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Goodhue County, Minnesota
Date of application for amendments: February 6, 1997, as
supplemented February 12, 1997.
Brief description of amendments: The amendments revise Technical
Specification 3.3.A to allow safety injection pump testing and
evolutions during low-temperature shutdown conditions provided controls
for reactor coolant system conditions are in place to provide low
temperature overpressurization protection.
Date of issuance: February 20, 1997.
Effective date: February 20, 1997, with full implementation within
30 days.
Amendment Nos.: 127 and 119.
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications and Bases.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. NRC published a public notice of the
proposed amendments, issued a proposed finding of no significant
hazards consideration, and requested that any comments on the proposed
finding be provided to the staff by close of business on February 14,
1997. The notice was published in the Red Wing Republican Eagle on
February 12, 1997, the Minneapolis Star Tribune on February 9, 1997,
and the St. Paul Pioneer Press on February 10, 1997. No comments have
been received.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, consultation with the State of Minnesota, and
final determination of NSHC are contained in a Safety Evaluation dated
February 20, 1997.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Dated at Rockville, Maryland, this 5th day of March 1997.
For The Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects--III/IV Office of Nuclear
Reactor Regulation.
[FR Doc. 97-5999 Filed 3-11-97; 8:45 am]
BILLING CODE 7500-01-P