[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