[Federal Register Volume 62, Number 223 (Wednesday, November 19, 1997)]
[Notices]
[Pages 61836-61855]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-30217]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice

Applications and Amendments to Facility Operating Licenses 
Involving No Significant Hazards Considerations

I. 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 October 27, 1997, through November 6, 1997. 
The last biweekly notice was published on November 5, 1997 (62 FR 
59912).

Notice Of Consideration Of Issuance Of Amendments ToFacility 
Operating Licenses, Proposed No Significant Hazards 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

[[Page 61837]]

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 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 December 19, 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.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to 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.

[[Page 61838]]

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: October 22, 1997
    Description of amendments request: The proposed amendment 
incorporates both steady state and transient degraded voltage setpoints 
into Technical Specifications, as opposed to the current single 
degraded voltage setpoint. The proposed changes ensure adequate 
terminal voltage to all safety-related equipment during steady state 
and transient voltage conditions. Additionally, the 4 kV voltage range 
required during testing of the emergency diesel generators (EDGs) will 
be decreased to ensure the new steady state degraded voltage relays are 
not actuated during testing and to ensure the 4 kV motors are operated 
within their voltage rating.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed changes revise the current degraded voltage 
setpoint and adds an additional steady state undervoltage 
requirement to Unit 1 and 2 Technical Specifications. The current 
degraded voltage relays will be referred to as
    <gr-thn-eq>transient degraded voltage relays.'' The new settings 
allow for calibration tolerances, potential transformer correction 
factors, test equipment uncertainties, and relay drift. The nominal 
settings account for the above factors, plus additional margin to 
the analytical limit. The acceptable voltage range during EDG 
surveillance testing is also being decreased. The setpoint and time 
delay associated with the 4 kV bus loss of voltage relays is 
unaffected by this amendment request.
    The accident analyses credit the loading of the EDGs based on 
loss of offsite power. The 4 kV emergency bus loss of voltage and 
degraded voltage relays initiate starting and loading of the 
emergency diesel generators (EDGs) when the preferred power source 
voltage is lost or drops below a predetermined value. The relays 
also initiate disconnection of the preferred power source from the 4 
kV emergency busses. These actions ensure adequate terminal voltage 
to all safety-related electrical equipment required to support 
accident mitigation. The required voltage necessary to ensure 
safety-related motors are capable of starting is 75 percent of 
nominal rated equipment voltage. The required voltage necessary to 
ensure these motors continue running for extended periods is 90 
percent of nominal rated equipment voltage.
    The degraded (transient) voltage setpoint is being changed from 
3628 [plus or minus] 25 Volts to 3710 [plus or minus] 80 Volts. 
Based on the most recent calculations, a minimum voltage of 3630 
Volts is required to ensure at least 75 percent of the nominal 
voltage is available to No. 13 Charging Pump, which is the most 
limiting electrical load.
    The new steady state degraded voltage relay setpoint will be 
established at 3900 [plus or minus] 80 Volts. The setpoint ensures 
that there is at least 90 percent of nominal voltage available to 
No. 13 Charging Pump. The time delay associated with this actuation 
is 101 [plus or minus] 3.5 seconds. The time delay provides adequate 
time for the voltage regulator to recover bus voltage following a 
voltage swing on the 500 kV system and time for the EDG voltage 
regulator to stabilize. The steady state degraded voltage relays 
will be tested in the same manner, and at the same frequency, as the 
loss of voltage and transient degraded voltage relays.
    The required voltage range during EDG surveillance testing is 
being revised from 4160 [plus or minus] 420 Volts to 4160 +240, -100 
Volts. The surveillance requirement verifies that the EDG voltage 
regulator is maintaining an acceptable voltage. The
    new value ensures the 4 kV motors are operated within their 
rated voltage and prevents actuation of the steady state degraded 
voltage relay during surveillance testing.
    The degraded voltage relays are not initiators in any previously 
evaluated accidents. Additionally, decreasing the acceptable voltage 
range during EDG testing does not affect the initiation of any 
previously analyzed accidents. Therefore, the proposed changes do 
not involve an increase in the probability or consequences of an 
accident previously analyzed.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The license amendment request revises the current degraded 
voltage setpoint and adds an additional steady state degraded 
voltage requirement. Additionally, the acceptable voltage range 
during EDG surveillance testing is being decreased. The proposed 
changes ensure adequate starting and running terminal voltage to all 
safety-related electrical equipment during steady state and 
transient degraded voltage conditions. The addition of the steady 
state degraded voltage relays provide an extra scheme of protection 
against sustained degraded voltage conditions. The facility 
currently relies upon degraded voltage relays to start and load the 
EDGs and to disconnect the preferred power source from the 4 kV 
emergency busses. Therefore, revising the relay setpoint, adding 
additional steady state degraded voltage protection, and decreasing 
the acceptable voltage range during EDG testing does not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The safety function of the degraded voltage relays is to ensure 
that the preferred power source is disconnected from the 4 kV 
emergency busses during loss of voltage or degraded voltage 
conditions. The relays also ensure the EDGs are started and loaded. 
Ultimately, these actions ensure the minimal terminal voltage 
necessary to start and run all safety-related electrical equipment 
is maintained. The proposed changes revise the current degraded 
voltage setpoint and adds an additional steady state undervoltage 
requirement. Additionally, the acceptable voltage range during EDG 
surveillance testing is being decreased to ensure actuation of the 
steady state degraded voltage relays does not occur during
    EDG testing, and to ensure the 4 kV motors are operated within 
their rated voltage range.
    Because the proposed changes ultimately ensure adequate terminal 
voltage to all safety-related electrical equipment during transient 
and steady state undervoltage conditions, the safety function of the 
degraded voltage relays, as well as the margin of safety afforded by 
these relays is unchanged.
    Therefore, the changes 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: S. Singh Bajwa, Director

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: October 2, 1997
    Description of amendment request: The proposed amendment would 
address an unreviewed safety question associated with the analysis of a 
fuel handling accident in the Fuel Storage Building as described in 
Section 15.7.4, ``Design Basis Fuel Handling Accidents,'' of the H.B. 
Robinson Steam Electric Plant (HBR) Updated Final Safety Analysis 
Report (UFSAR). Carolina Power & Light Company (the licensee) 
determined that an assumption used in the accident analysis for depth 
of water above the top of irradiated fuel in the spent fuel pit was 
non-conservative. The accident analysis assumed a depth of 23 feet 
instead of the correct value of 21 feet. The licensee has submitted a 
revised accident analysis using the correct assumption and has proposed 
that the UFSAR be

[[Page 61839]]

changed to incorporate the results of the revised analysis.
    Basis for proposed no significant hazards 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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change to the UFSAR is to change assumptions 
associated with the evaluation of a fuel handling accident in the 
Fuel Storage Building. The change in assumptions is to reduce the 
decontamination factor associated with the removal of elemental 
iodine from the spent fuel pool water. Because the decontamination 
factor for elemental iodine is reduced, the consequences of a fuel 
handling accident in the Fuel Storage Building is [sic] increased. 
However, because the radiological consequences remain well within 
the exposure guideline values of 10 CFR 100, paragraph 11 (i.e., 25% 
or less of the values), the increase in consequences is not 
significant. The change in assumptions for the fuel handling 
accident in the Fuel Storage Building do [sic] not affect operation, 
maintenance, or design of equipment associated with the handling of 
fuel in the Fuel Storage Building, therefore, the probability of a 
fuel handling accident as previously evaluated is not changed.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components, changes in parameters 
governing normal plant operation, or methods of operation. The 
proposed change does not introduce a new mode of operation or 
changes in the method of normal plant operation. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change to the UFSAR to change the assumptions 
associated with a fuel handling accident in the Fuel Storage 
Building is to change the assumption for the decontamination factor 
for elemental iodine to a smaller value. The new assumption for 
elemental iodine decontamination factor preserves the approximate 
factor of 24 margin between experimental data for elemental iodine 
decontamination factor and the assumed value provided in NRC Safety 
Guide 25. Therefore, the 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 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: James E. LyonsCommonwealth Edison Company, 
Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 
and 3, Grundy County, Illinois Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois Date of application for amendment request: September 30, 1997
    Description of amendment request: This request changes the 
Technical Specifications (TS) by adding a new Section 3/4.12.C, 
``Inservice Leak and Hydrostatic Testing Operation,'' to allow certain 
reactor coolant pressure tests to be performed in MODE 4 when the 
metallurgical characteristics of the reactor pressure vessel require 
the pressure testing at or approaching temperatures 
<ls-thn-eq>212 deg.F, which normally correspond with MODE 3.
    Basis for proposed no significant hazards 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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    The proposed amendment represents the addition of a Special Test 
Exception to perform Pressure Testing Operations consistent with the 
requirements of Section 3.10.1 of the Improved Standard Technical 
Specifications (NUREG-1433). The proposed changes are consistent 
with the current plant safety analyses. Implementation of these 
changes will provide continued assurance that specified parameters 
associated with Pressure Testing Operations will remain within their 
acceptance limits, and as such, will not significantly increase the 
probability or consequences of a previously evaluated accident.
    The proposed changes are based on requirements specified by 
Section 3.10.1 of NUREG-1433. Any such changes are consistent with 
the current plant safety analyses and have been determined to 
represent sufficient requirements for the assurance and reliability 
of equipment assumed to operate in the safety analyses, or provide 
continued assurance that specified parameters associated with 
Pressure Testing Operations remain within their acceptance limits. 
As such, these changes will not significantly increase the 
probability or consequences of a previously evaluated accident.
    The associated systems affecting Pressure Testing Operations 
related to this proposed amendment are not assumed in any analyses 
to initiate any accident sequence; therefore, the probability of any 
accident previously evaluated is not increased by this proposed 
amendment which incorporates the requirements of Section 3.10.1 of 
NUREG-1433. In addition, the proposed limiting conditions for 
operation and surveillance requirements for the proposed amendment 
ensure a level of equipment operability sufficient to mitigate any 
operational occurrences which could occur while operating under this 
Special Test Exception. Furthermore, any operational occurrence 
postulated during operation under this Special Test Exception is 
bounded by the Design Basis Accidents. Therefore, the proposed 
amendment does not increase the consequences of nay accident 
previously evaluated.
    There is no change to the consequences of an accident previously 
evaluated because Pressure Testing Operations does not adversely 
affect either the on-site or off-site does consequences resulting 
from an accident. In addition, Pressure Testing Operations is not an 
accident initiator. As such, there is no adverse impact on the 
probability of accident initiators. Thus, there is no significant 
increase in the probability of any previously analyzed accident.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The proposed amendment represents the conversion of current 
Technical Specification requirements to maintain consistency with 
those requirements specified in Section 3.10.1 of NUREG-1433. The 
proposed changes are consistent with the current plant safety 
analyses. These proposed changes do not involve revisions to the 
design of the station. In addition, the proposed limiting conditions 
for operation and surveillance requirements for the proposed 
amendment ensure a level of equipment operability sufficient to 
mitigate any operational occurrences which could occur while 
operating under the Special Test Exception. Some of the changes may 
involve revision in the testing of components at the station; 
however, these are in accordance with the current plant safety 
analyses. The proposed changes will not introduce new failure 
mechanisms beyond those already considered In the current plant 
safety analyses.
    The associated systems that affect Pressure Testing Operations 
related to the proposed amendment, are not assumed in any plant 
safety analysis to initiate any accident sequence. In addition, the 
proposed surveillance requirements for any such affected systems are 
consistent with the requirements of Section 3.10.1 of NUREG-1433. 
Therefore, the possibility of a new or different kind of accident 
from any accident previously evaluated is not created.
    3) Involve a significant reduction in the margin of safety 
because:
    ComEd proposes to revise the Technical Specifications to be 
consistent with those provisions specified in Section 3.10.1 of 
NUREG-1433. The proposed changes are consistent with the current 
plant safety analyses. In addition, these proposed changes do not 
involve revisions to the design of the station. As such, the 
proposed individual changes will maintain the same level of

[[Page 61840]]

reliability of the equipment associated with Pressure Testing 
Operations, assumed to operate in the plant safety analysis, or 
provide continued assurance that specified parameters affecting, 
will remain within their acceptance limits. Therefore, the proposed 
changes provide continued assurance of Pressure Testing Operations 
without adversely affecting the public health and safety and as 
such, will not significantly reduce existing plant safety margins.
    The proposed amendment to the Technical Specifications 
implements present requirements, or the requirements in accordance 
with the guidelines set forth in Section 3.10.1 of NUREG-1433. The 
proposed changes have been evaluated and found to be acceptable for 
use at the stations based on system design, safety analysis 
requirements, and operational performance. Since the proposed 
changes are based on NRC accepted provisions that are applicable at 
the stations and maintain necessary levels of system or component 
reliability affecting Pressure Testing Operations, the proposed 
changes 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 are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidle and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: September 26, 1997
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to (1) prohibit the simultaneous 
opening of the drywell and suppression chamber purge system isolation 
valves, (2) upgrade the ventilation filter testing program to the 
latest industry standards, and (3) specify that the auxiliary electric 
equipment room is required to be habitable during design bases 
accidents.
    Basis for proposed no significant hazards 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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    a. Drywell and Suppression Chamber Purge System
    The purpose of the drywell and suppression chamber purge system 
isolation valves is to mitigate the consequences of a design bases 
accident. Operation of these valves will have no effect on the 
probability of a design bases accident occurring.
    The current TS 3.6.1.8 allows for the drywell and suppression 
chamber purge system isolation valves to be open simultaneously. In 
this condition, containment pressure and offsite dose during design 
bases accidents would be greater than previously evaluated. The 
proposed revision to TS 3.6.1.8 would prevent the simultaneous 
opening of the drywell and suppression chamber purge system 
isolation valves thus assuring that the consequences of design bases 
accidents previously evaluated are still bounding.
    b. Ventilation Filter Testing Program
    The SBGTS [Standby Gas Treatment System] and Control Room and 
AEER [Auxiliary Electric Equipment Room] Emergency Filtration 
Systems are designed to mitigate the radiological consequences of 
previously evaluated design bases accidents. Operation and testing 
of these systems will have no effect on the probability of a design 
bases accident occurring.
    The proposed revisions associated with this change relocate the 
requirements for SBGTS and Control Room and AEER Emergency 
Filtration System filter testing from the current TS SRs to a new TS 
administrative control program. The testing requirements are being 
upgraded to the latest industry standards. Filter testing in 
accordance with the proposed program will ensure that Title 10, Code 
of Federal Regulations, Part 50 (10 CFR 50), Appendix A, General 
Design Criteria (GDC) 19 and 10 CFR 100 limits are not exceeded.
    c. Other Control Room and Auxiliary Electric Equipment Room 
Emergency Filtration System Changes
    The SBGTS and Control Room and AEER Emergency Filtration System 
are designed to mitigate the radiological consequences of previously 
evaluated design bases accidents. Operation and testing of these 
systems will have no effect on the probability of a design bases 
accident occurring.
    The proposed revisions associated with this change acknowledge 
that the AEERs are required to be habitable during design bases 
accidents. This is consistent with the plants design bases.
    d. Editorial Changes
    The proposed revisions to TS 6.2.F.7 reformat the requirement to 
establish consistency with the remainder of TS 6.2.F. There are no 
technical changes being proposed.
    Based upon the above, the proposed amendment will 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 because:
    a. Drywell and Suppression Chamber Purge System
    No new plant equipment is being installed, and use of currently 
installed plant equipment is not affected by this proposed change. 
The proposed revision to TS 3.6.1.8 provides additional limitations 
on the opening of the drywell and suppression chamber purge system 
isolation valves.
    b. Ventilation Filter Testing Program
    No new plant equipment is being installed, and use of currently 
installed plant equipment is not affected by this proposed change. 
These proposed revisions will demonstrate operability of the Control 
Room and AEER Emergency Filtration System using the latest industry 
standards.
    c. Other Control Room and Auxiliary Electric Equipment Room 
Emergency Filtration System Changes
    No new plant equipment is being installed, and use of currently 
installed plant equipment is not affected by this proposed change. 
These proposed revisions will demonstrate habitability of the AEER 
by imposing operability requirements on the AEER recirculation 
filter units.
    d. Editorial Changes
    The proposed revisions to TS 6.2.F.7 reformat the requirement to 
establish consistency with the remainder of TS 6.2.F. There are no 
technical changes being proposed.
    Based upon the above, the proposed change will not create the 
possibility of a new or different kind of accident or transient 
previously evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    a. Drywell and Suppression Chamber Purge System
    The current TS 3.6.1.8 requirements are non-conservative with 
respect to the assumptions used when evaluating steam bypass of the 
suppression chamber; specifically, a maximum allowable leakage area 
of 0.03 square feet with the only credible leakage path was assumed 
to be suppression chamber vacuum breaker valve seat leakage. This 
proposed revision to TS 3.6.1.8 will make the TS requirements 
consistent with those assumptions.
    b. Ventilation Filter Testing Program
    These proposed revisions will ensure operability of the Control 
Room and Auxiliary Electric Equipment Room (AEER) Emergency 
Filtration system using the latest industry standards. Filter 
testing in accordance with the proposed program will ensure that GDC 
19 and 10 CFR 100 limits are not exceeded.
    c. Other Control Room and Auxiliary Electric Equipment Room 
Emergency Filtration System Changes
    These proposed revisions will ensure operably of the control 
room and AEER Emergency Filtration System by demonstrating system 
performance with the control room and AEER recirculation filter 
units to ensure GDC 19 limits are not exceeded.
    d. Editorial Changes
    The proposed revisions to TS 6.2.F.7 reformat the requirement to 
establish consistency with the remainder of TS 6.2.F. There are no 
technical changes being proposed.

[[Page 61841]]

    Based on the above, the proposed TS 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:Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348
    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. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: October 15, 1997
    Description of amendment request: The proposed amendments would 
revise Technical Specification Table 4.3.7.5-1, Accident Monitoring 
Instrumentation Surveillance Requirements, by deleting a footnote that 
provides details concerning the calibration requirements for the 
drywell hydrogen concentration analyzer and monitor.
    Basis for proposed no significant hazards 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) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The drywell hydrogen concentration analyzer and monitors are 
required to be operable by TS 3/4.7.5, Accident Monitoring 
Instrumentation. Table 4.3.7.5-1, Accident Monitoring 
Instrumentation Surveillance Requirements, includes a footnote 
providing unnecessary details related to the calibration of this 
specific analyzer and monitors. The footnote provides information 
that was determined to put the hydrogen analyzers and monitors 
outside of the design basis by limiting the range of the indication 
to 0% to 4% hydrogen in the drywell. The calibration method is being 
corrected to provide the correct range of 0% to 10%, and requires 
this note in the TS to be changed or deleted. The footnote is 
proposed to be deleted from the TS, because it provides unnecessary 
detail.
    Deletion of the footnote will not cause an increase in the 
probability of an accident, because this instrumentation is only for 
accident monitoring instrumentation and thus does not affect 
accident initiators or assumptions.
    Deletion of the footnote will not change the consequences of an 
accident previously evaluated, because this detail in the TS does 
not change the requirement of performing a channel calibration at 
the specified frequency. In addition, the ability to monitor 
hydrogen during an accident will not be affected by deletion of the 
footnote.
    Therefore, this change does not involve an increase in the 
probability or consequences of an accident previously evaluated.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    This is monitoring instrumentation only. Deletion of the 
footnote concerning specifics on how to calibrate this 
instrumentation will not affect the reliability or failure modes of 
the drywell hydrogen concentration analyzer and monitors. Therefore, 
this change will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    This is monitoring instrumentation only. Deletion of the 
footnote concerning specifics on how to calibrate this 
instrumentation will not change the requirement to perform Channel 
Calibrations at the frequency specified in the TS. The details of 
how to perform a Channel Calibration on the drywell hydrogen 
concentration analyzer and monitors are located in plant procedures 
and are in accordance with vendor recommendations. The TS 
requirements for redundancy of the instrumentation and the actions 
to be taken for inoperable instrumentation are also not affected by 
the deletion of this footnote.
    This change to the level of information regarding this 
calibration is consistent with the detail for this and other 
instrumentation in NUREG-1434, Revision 1, Standard Technical 
Specifications, General Electric Plants, BWR/6.
    Therefore, deletion of footnote * from TS Table 4.3.7.5-1 will 
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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location:Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: October 6, 1997
    Description of amendment request: The proposed amendments would 
delete all references to the steam line low pressure safety injection 
function.
    Basis for proposed no significant hazards 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Answer
    Probability
    Accident initiators can affect the probability of a previously 
evaluated accident. The addition of a new device or piece of 
equipment to the plant may introduce a new accident initiator. No 
new equipment is added to the plant as a result of this change. The 
proposed removal of the low steam line steam pressure will involve 
removing the steam line pressure safety injection function. This 
results in a reduction in the likelihood of spurious safety 
injections. Spurious safety injections can result in inadvertent 
ECCS [emergency core cooling system] actuations. Inadvertent ECCS 
Actuation is a UFSAR [updated Final Safety Analysis Report] accident 
(UFSAR 15.5.1). Therefore, this change will result in a reduction in 
the probability of an accident previously evaluated.
    Routine plant operating practices and conditions will not be 
altered by the removal of the safety injection function. Therefore, 
there is no operating practice or condition change that could 
increase the probability of occurrence of a previously evaluated 
accident.
    There is no significant increase in the probability of an 
accident previously evaluated.
    Consequences
    Accidents previously evaluated that could be adversely affected 
are the steam line break and the feedwater line break. These 
accidents will result in secondary side depressurization with 
pressure reaching the current actuation setpoint. The review of 
these accidents found that the consequences of the previous accident 
analysis acceptance criteria remain satisfied. The specifics of the 
accident analysis is discussed below.
    The steam line break accident was analyzed to demonstrate short 
term cooling capability. A spectrum of break sizes were evaluated to 
determine the limiting break size. For smaller breaks (including the 
limiting break size), the safety injection actuation on low 
pressurizer pressure occurs prior to low steam line pressure safety 
injection. However, for larger steam line breaks the setpoint for 
low steam line pressure safety injection is reached prior to low 
pressurizer pressure safety injection. The larger spectrum of breaks 
were analyzed without credit for the low steam line pressure safety 
injection. The results of this analysis found that there would be a 
slight increase in time required for safety injection to actuate. 
The low pressurizer safety injection would actuate in these 
accidents due to the cooldown and depressurization of the reactor 
coolant system in response to the secondary side energy removal. The 
Departure from

[[Page 61842]]

Nucleate Boiling Ratios (DNBRs) were analyzed with this time delay 
in safety injection. The DNBRs for these cases were found to be less 
limiting than those calculated for the limiting break size. 
Therefore, the removal of steam line low pressure safety injection 
does not adversely affect the DNBR, fuel failure or dose 
consequences of the main steam line break accident. Other acceptance 
criteria would not be expected to be affected by the small change in 
timing of the safety injection signal.
    In addition, to the Chapter 15 accident analysis, the Chapter 6 
containment response to mass and energy releases was evaluated 
without credit for steam line low pressure safety injection. The 
evaluation demonstrated that for steam line breaks inside of 
containment, the high containment pressure safety injection set 
point is reached prior to the pressure associated with steam line 
low pressure safety injection. Therefore the existing containment 
response evaluation is not adversely affected by the removal of the 
low steam pressure safety injection. This also assures that the 
existing environmental qualification envelope for McGuire is not 
affected by this change. For steam line breaks outside of 
containment the maximum required breaksize is 1.0 ft2, which results 
in transients with safety injection caused by low pressurizer 
pressure prior to low steam line pressure safety injection.
    The feedwater line break accidents were analyzed to demonstrate 
long term core cooling capability. During a feedwater line break, 
the secondary system will depressurize if the break occurs between 
the main feedwater check valve and the steam generator. However, 
breaks are only required to occur at the terminal ends of feedwater 
piping (i.e., at the feedwater pump or at the steam generator). For 
a feedwater line break at the main feedwater pump, the main feed 
check valve will prevent depressurization of the steam generator. 
For a feedwater line break at the steam generator, a safety 
injection on high containment pressure will occur prior to safety 
injection on steam pressure. Therefore, the elimination of the steam 
line low pressure safety injection does not adversely impact the 
feedwater line break accident.
    In summary, a review was conducted of all design basis accidents 
to identify those which result in a low steam pressure safety 
injection. These accidents were then evaluated to verify that the 
accident analysis were within acceptance criteria. This review 
revealed that all accident analysis results were within current 
analysis acceptance criteria.
    Therefore, there is no significant increase in the consequences 
of a previously evaluated accident.
    Conclusion
    Elimination of the low steam line pressure safety injection 
results in no 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[?]
    Answer
    There is no introduction of new equipment or operating practices 
that could result in a new operating condition. The plant will 
continue to operate in the same method with the same complement of 
equipment with the exception of the actuation logic associated with 
the steam line low pressure safety injection. Therefore, there is no 
new operating condition that would be expected to generate a new 
sequence of events which could generate a new or different accident. 
There is no new equipment that could interact with other plant 
structures, systems or components.
    The low pressure safety injection equipment is the only plant 
equipment affected by this change. There are no new equipment 
failure modes which might result in a new or different accident.
    Affected accidents were evaluated to validate that the accident 
sequence would not deviate in a fashion which would create a new or 
different accident. The analysis of the feedwater line break and 
steam line break did not reveal any new or different type of 
accident.
    Removal of the low steam line pressure safety injection will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated;
    (OR)
    3. Involve a significant reduction in the margin of safety?
    Answer
    The margin of safety relevant to this change is represented by 
the margin of physical protection provided by fuel cladding and the 
reactor containment. Effects of this change on the safety analysis 
was described under question 1 above. The results of the analysis 
demonstrate that DNBR, fuel clad integrity and containment response 
were not significantly affected by the removal of low steam line 
pressure safety injection. Therefore, the physical protection 
provide[d] by the fuel cladding and reactor containment were not 
affected by this change. Accident acceptance criteria continued to 
be met without credit for the safety function. The radiological 
consequences of accidents was not affected by the change.
    The removal of the low steam line pressure safety injection did 
not significantly reduce 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: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina

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: October 15, 1997
    Description of amendment request: The proposed amendment would 
affect nominal trip setpoints and allowable values for Reactor Trip 
System (RTS) Instrumentation Trip Setpoints Table 2.2-1, and Engineered 
Safety Features Actuation System (ESFAS) Instrumentation Trip Setpoints 
Table 3.3-4. In addition, the proposed amendment would (1) decrease the 
reactor trip setpoint for the reactor coolant pump (RCP) low shaft 
speed (underspeed trip setpoint) from 95.8 percent to 92.4 percent of 
rated speed, (2) make editorial changes, and (3) change the Bases to 
reflect the new methodology.
    Basis for proposed no significant hazards 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 revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed revision does not involve [an] SHC because 
the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes to Tables 2.2-1 and 3.3-4 involve changes 
from a five column format to a two column format. The RTS trip 
setpoints and ESFAS trip setpoints remain unchanged with the 
exception of the RCP low shaft speed trip setpoint discussed below. 
Detailed operability criteria will be moved to surveillance 
procedures and analysis has demonstrated that an adequate margin for 
normal trip setpoints exist and safety analysis limits are preserved 
in all RTS/ESFAS functions.
    Changing the RCP low shaft speed trip setpoint will not change 
the probability of occurrence of the event. The existing accident 
analysis (Millstone Unit No. 3 FSAR [final safety analysis report] 
section 15.3.2) of the complete loss of forced reactor coolant flow 
remains valid for the proposed change. Therefore, the change to the 
RCP low shaft speed trip setpoint does not increase the probability 
or consequences of any previously analyzed accident.
    In addition, the proposed changes to Tables 2.2-1 and 3.3-4 do 
not alter the intent or method by which the surveillances are 
conducted. Therefore, the scope of evaluation performed gives 
reasonable assurance that there will not be an adverse impact on the 
consequences or the probability of any previously analyzed accident.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.

[[Page 61843]]

    The existing design basis adequately covers the plant response 
with the proposed change to the RCP low shaft speed trip setpoint. 
The change does not introduce new failure modes.
    The proposed changes to Tables 2.2-1 and 3.3-4 do not modify the 
design or operation of any plant system. The proposed changes do not 
alter the intent or method by which the surveillances are conducted, 
other than adjusting the allowable values to reflect historical 
instrument performance data. Therefore, the proposed revision does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to Tables 2.2-1 and 3.3-4 modify the 
existing five column format to a two column format to show the RTS 
and ESFAS nominal trip setpoints and the process rack bistable 
allowable values for individual functions. Detailed operability 
criteria will be moved to the surveillance procedures. With the 
exception of the low shaft speed trip discussed below, the RTS and 
ESFAS setpoints remain unchanged and analysis has demonstrated that 
an adequate margin for normal trip setpoints exist and safety 
analysis limits are preserved in all RTS/ESFAS functions.
    Since the safety limits of the design are still met, the 
proposed change to the RCP low shaft speed trip setpoint does not 
reduce the margin of safety.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve an SHC.
    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, 
Connecticut NRC Deputy Director: Phillip F. McKee

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of amendment requests: March 10, 1997, as supplemented by 
letter dated May 20, 1997
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 to revise TS 3/4.4.5 and 3.4.6.2, 
including associated Bases 3/4.4.5 and 3/4.4.6.2, to allow the 
implementation of steam generator (SG) tube alternate repair criteria 
for axial indications in the Westinghouse explosive tube expansion 
(WEXTEX) region below the top of the tubesheet and below the bottom of 
the WEXTEX transition that may exceed the current TS depth-based 
plugging limit. The allowed primary-to-secondary operational leakage 
from any one SG would be reduced from 500 gpd to 150 gpd.
    Basis for proposed no significant hazards 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.
    Probability
    Of the various accidents previously evaluated, the proposed 
changes only affect the steam generator tube rupture (SGTR) event 
evaluation and the postulated steam line break (SLB) accident 
evaluation. Loss-of-coolant accident (LOCA) conditions cause a 
compressive axial load to act on the tube. Therefore, since the LOCA 
tends to force the tube into the tubesheet rather than pull it out, 
it is not a factor in this amendment request. Another faulted load 
consideration is a safe shutdown earthquake (SSE); however, the 
seismic analysis of Series 51 steam generators has shown that axial 
loading of the tubes is negligible during an SSE.
    For the SGTR event, the required structural margins of the steam 
generator tubes will be maintained by the presence of the tubesheet. 
Tube rupture is precluded for cracks in the Westinghouse explosive 
tube expansion (WEXTEX) region due to the constraint provided by the 
tubesheet. Therefore, Regulatory Guide (RG) 1.121, ``Bases for 
Plugging Degraded PWR Steam Generator Tubes,'' margins against burst 
are maintained for both normal and postulated accident conditions.
    The W* length supplies the necessary resistive force to preclude 
pullout loads under both normal operating and accident conditions. 
The contact pressure results from the WEXTEX expansion process, 
thermal expansion mismatch between the tube and tubesheet and from 
the differential pressure between the primary and secondary side. 
Therefore, the proposed change results in no significant increase in 
the probability of the occurrence of an SGTR or SLB accident. 1
    The proposed changes do not affect other systems, structures, 
components or operational features. Therefore, based on the above 
evaluation, the proposed changes do not involve a significant 
increase in the probability of an accident previously evaluated.
    Consequences
    The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary 
leakage flow through a postulated broken tube is not affected by the 
proposed change since the tubesheet enhances the tube integrity in 
the region of the WEXTEX expansion by precluding tube deformation 
beyond its initial expanded outside diameter. The resistance to both 
tube rupture and collapse is strengthened by the tubesheet in that 
region. At normal operating pressures, leakage from primary water 
stress corrosion cracking (PWSCC) in the W* length is limited by 
both the tube-to-tubesheet crevice and the limited crack opening 
permitted by the tubesheet constraint. Consequently, negligible 
normal operating leakage is expected from cracks within the 
tubesheet region.
    SLB leakage is limited by leakage flow restrictions resulting 
from the crack and tube-to-tubesheet contact pressures that provide 
a restricted leakage path above the indications and also limit the 
degree of crack face opening compared to free span indications. The 
total leakage, that is, the combined leakage for all such tubes, 
plus the combined leakage developed by any other ARC, must be below 
the maximum allowable SLB leak rate limit, such that off-site doses 
are maintained less than 10 CFR 100 guideline values.
    Therefore, based on the above evaluation, the proposed changes 
do 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.
    The proposed changes do not introduce any changes or mechanisms 
that create the possibility of a new or different kind of accident. 
Tube bundle integrity is expected to be maintained for all plant 
conditions upon implementation of the proposed steam generator 
alternate tube plugging criteria.
    WCAP-14797, Revision 1, ``Generic W* Tube Plugging Criteria for 
51 Series Steam Generator Tubesheet Region WEXTEX Expansions,'' 
requires that any tubes with indications identified using the bobbin 
coil probe during the bobbin sampling plan also be inspected with 
the RPC coil throughout the W* length of the tubes. The use of the 
RPC will: (a) identify any new or non-expected degradation mode that 
may not be identified using the bobbin coil probe, and (b) confirm 
and characterize the bobbin coil indication.
    These changes do not introduce any new equipment or any change 
to existing equipment. No new effects on existing equipment are 
created nor are any new malfunctions introduced.
    Therefore, based on the above evaluation, the proposed changes 
do 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 maintain the required structural margins of 
the steam

[[Page 61844]]

generator tubes for both normal and accident conditions. RG 1.121 is 
used as the basis in the development of the W* alternate tube 
plugging criteria for determining that steam generator tube 
integrity considerations are maintained within acceptable limits. RG 
1.121 describes a method acceptable to the NRC staff for meeting 
General Design Criteria 14, 15, 31, and 32 by reducing the 
probability and consequences of an SGTR. RG 1.121 concludes that by 
determining the limiting safe conditions of tube wall degradation 
beyond which tubes with unacceptable cracking, as established by 
inservice inspection, should be removed from service or repaired, 
the probability and consequences of an SGTR are reduced. This RG 
uses safety factors on loads for tube burst that are consistent with 
the requirements of Section III of the ASME Code.
    For primarily axially oriented cracking located within the 
tubesheet, tube burst is precluded due to the presence of the 
tubesheet. WCAP-14797 defines a length, W*, of degradation free 
expanded tubing that provides the necessary resistance to tube 
pullout due to the pressure induced forces (with applicable safety 
factors applied). Application of the W* criteria will preclude 
unacceptable primary-to-secondary leakage during all plant 
conditions. The methodology for determining leakage provides for 
large margins between calculated and actual leakage values in the W* 
criteria.
    Plugging of the steam generator tubes reduces the reactor 
coolant flow margin for core cooling. Implementation of the proposed 
changes are expected to result in plugging of fewer tubes than with 
the current criteria. Thus, implementation of the proposed changes 
will maintain the margin of flow that may have otherwise been 
reduced by tube plugging.
    Based on the above, it is concluded that the proposed changes do 
not result in a significant reduction of margin with respect to 
plant safety as defined in the FSAR Update or bases of the plant 
Technical Specifications.
    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 requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

Philadelphia Electric Company, Docket No. 50-352, Limerick 
Generating Station, Unit 1, Montgomery County, Pennsylvania

    Date of amendment request: October 24, 1997
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would revise TS Section 3/4.1.3.6 to exempt 
control rod 50-27 from the coupling test for the remainder of Cycle 7.
    Basis for proposed no significant hazards 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 Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. The probability of occurrence of 
the analyzed Control Rod Drop Accident (CRDA) is not increased by 
operating with the subject control blade in a condition not known to 
be coupled since the compensatory measures will assure that the 
blade will remain fully inserted below 10% rated thermal power where 
the CRDA is a concern. Monitoring of nuclear instrumentation 
responses in the vicinity of the blade when the drive is withdrawn 
above 10% power will assure the blade is tracking with the drive 
with no potential to stick and then drop. Scram impact forces from 
an uncoupled control rod are of insufficient energy to dislodge the 
fuel support (or fuel) or to cause a threat to the pressure boundary 
integrity. No reduction of system or equipment redundancy is 
involved.
    The CRDA analyzed in the Safety Analysis Report (SAR) remains 
the limiting rod drop accident, and its consequences are unaffected 
by operation of the subject blade in the proposed manner. Operation 
of the control blade as described, i.e., withdrawn no further than 
the 46 position and in a condition not known to be coupled, has no 
adverse effect on scram performance in response to any other 
postulated accident. The scram insert motion of the rod is not 
affected by the potentially uncoupled condition, and since the rod 
is already partially inserted at position 46, it should have a 
slightly better negative reactivity insertion characteristic. 
Therefore, no potential to increase onsite or offsite radiological 
consequences beyond those previously analyzed in the SAR is created.
    Operating the subject control blade in a condition not known to 
be coupled does not result in any onsite or offsite radiological 
consequences different from those previously analyzed in the SAR. 
The subject control blade will be fully inserted below 10% thermal 
power where the CRDA is a concern and will be monitored during drive 
withdrawal above 10% thermal power to assure it is tracking with the 
drive. Scram performance is not adversely affected by operation from 
the near full-out position of 46. Hence, no new failure modes are 
created and consequences of any postulated failures are not 
increased.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The Safety Analysis Report (SAR) analyzed Control Rod Drop 
Accident (CRDA) remains the only type of accident initiated (or 
contributed to) by the control rod drive/control blade interface. 
The compensatory actions to be taken when operating the subject 
blade in a condition not verified to be coupled assure that no new 
types of accidents can occur. The subject control blade will be 
fully inserted below 10% thermal power where the CRDA is a concern 
and will be monitored during drive withdrawal above 10% thermal 
power to assure it is tracking with the drive. Scram performance is 
not adversely affected by operation from the near full-out position 
of 46. Since no adverse effect on insertion or scram performance is 
expected, the previously analyzed accidents encompass any potential 
consequence of operating with an uncoupled control blade.
    The compensatory actions to be taken when operating the subject 
blade in a condition not verified to be coupled assure that no new 
failure modes are created, and, therefore, no new type of equipment 
malfunction is introduced by operating the subject control blade in 
the proposed manner.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    Operation with the subject control blade in a condition not 
known to be coupled for the remainder of Cycle 7 at LGS [Limerick 
Generating Station] Unit 1, but with the compensatory actions 
described below, does not reduce the existing margin of safety 
determined by the analysis of the Control Rod Drop Accident (CRDA). 
The CRDA analyzed in the Safety Analysis Report (SAR) remains 
bounding in that the subject rod will be fully inserted below 10% 
rated thermal power where the CRDA is a concern. Above 10% power, 
when the associated drive is withdrawn, the nuclear instrumentation 
in the vicinity of the blade will be monitored to assure the blade 
tracks with the drive, providing assurance that the position of the 
blade can be ascertained by the drive position. If the control blade 
can not be verified to have followed the drive, then the rod shall 
be completely inserted and the control rod directional valves 
disarmed in accordance with existing TS requirements. To minimize 
any scram impact loadings, the blade will be operated at the near 
full-out position of 46 except for intermediate positions 
temporarily occupied during standard rod withdrawal sequences. 
Operating the subject control blade in the proposed manner will have 
no adverse effect on insertion or scram performance of the blade and 
will preserve the margin of safety.

[[Page 61845]]

    Therefore, the proposed TS change does 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:Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-272, Salem 
Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

    Date of amendment request: October 6, 1997
    Description of amendment request: The amendment to the Technical 
Specifications would increase the allowable band for control and 
shutdown rod demanded position versus indicated position from plus or 
minus 12 steps to plus or minus 18 steps when the power level is not 
greater than 85% rated thermal power. The amendment is identical to 
Amendment 183 for Salem Unit 2, which was issued September 10, 1997, as 
an exigent amendment.
    Basis for proposed no significant hazards 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 change to the rod misalignment criteria of [plus or 
minus] 18 steps for core powers equal to or below 85% of RATED 
THERMAL POWER (RTP) does not increase the probability of previously 
evaluated accidents. Increasing the magnitude of the allowed control 
rod misalignment is not a contributor to the mechanistic cause of an 
accident evaluated in any accident analysis. The magnitude of 
control rod indicated misalignment is a parameter used to establish 
the initial conditions for accident evaluation.
    The proposed increase in the allowable rod misalignment from the 
current [plus or minus] 12 steps for reactor powers equal to or less 
than 85% RTP does not involve a significant increase in the 
consequence of any previously evaluated accident. Rod misalignment 
affects power distribution, shutdown margin and the ejected rod 
accident. An extension of the allowable rod misalignment above and 
below 85% RTP has been analyzed in Westinghouse WCAP-14672. As 
provided in WCAP-14672, above 85% the allowable misalignment is 
governed by the available peaking factor margins as determined by 
flux maps.
    [Public Service Electric & Gas] PSE&G is simplifying the 
proposed change by keeping the currently allowed [plus or minus] 12 
step misalignment in Technical Specifications 3.1.3.1 and 3.1.3.2.1 
for reactor power greater than 85% RTP.
    The PSE&G proposed change is to allow [plus or minus] 18 steps 
misalignments in Technical Specifications 3.1.3.1 and 3.1.3.2.1 for 
reactor power less than or equal to 85% RTP. As demonstrated in 
WCAP-14672, for reactor powers less than 85% RTP, the available 
peaking factor margin increases faster than any penalty associated 
with a [plus or minus] 18 step misalignment.
    As described in Section 4.0 of the Westinghouse WCAP, a 
conservative penalty factor has been applied to the rod insertion 
allowance (RIA) of the shutdown margin calculation to account for 
rods misaligned an additional [plus or minus] 6 steps (for a total 
of [plus or minus] 18 steps). This conservative penalty factor is 
applied as part of the reload analysis in order to satisfy Technical 
Specification 3.1.1.1.
    In addition to the normal, or Condition 1, operational 
transients, the impacts of increased rod misalignment on Condition 
II, III and IV accident analysis have also been evaluated. The 
proposed increase in rod misalignment does not have a significant 
effect on any moderator or Doppler reactivity coefficients or 
defects, boron worth or reactor kinetics parameters.
    To account for the potential increase in ejected rod parameters, 
conservative penalty factors have been applied to the reload safety 
evaluation to cover the additional [plus or minus] 6 step 
misalignment. Margin is available in the reload safety analysis to 
accommodate this impact.
    Therefore, the proposed amendment does not increase the 
probability or consequences of any 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.
    0No new accident scenarios, failure mechanisms or limiting 
single failures are introduced as a result of the proposed change to 
the rod misalignment criteria of [plus or minus] 18 steps below 85% 
RTP. The implementation of the proposed rod misalignment criteria 
will have no adverse effect on the performance of any other safety 
related system. Therefore, the proposed amendment does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety. The Technical Specifications allowed increase in peaking 
factors as power is reduced accommodates the peaking factor penalty 
associated with the additional [plus or minus] 6 step misalignment 
for core powers equal to or less than 85% RTP. Therefore, there is 
no change to the peaking factors assumed in the safety analysis. In 
addition to peaking factors, there is no change in any other current 
limit input into the safety analysis. As the input, or initial 
conditions, of the safety analysis have not changed, there is no 
reduction in the margin to 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

Public Service Electric & Gas Company, Docket No. 50-272, Salem 
Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

    Date of amendment request: October 14, 1997
    Description of amendment request: The proposed amendment will 
modify the Salem Unit 1 Technical Specification (TS) 3.4.6.3, ``Primary 
Coolant System Pressure Isolation Valves Limiting Condition for 
Operation,'' to be consistent with Salem Unit 2 TSs.
    Basis for proposed no significant hazards 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 majority of the proposed changes, as described above, are 
editorial in nature. Rewording, and reformatting the Limiting 
Condition for Operation, including the surveillance requirements do 
not involve a significant increase to the probability or 
consequences of an accident.
    Those substantive changes involving the addition of (1) new 
reactor coolant system pressure isolation valves, (2) providing for 
a shorter test frequency upon entry into Mode 4, and (3) adding a 
new surveillance test requirement, do not increase the probability 
or consequences of an accident. These changes ensure that the system 
and components needed to prevent and minimize the effects of inter-
system loss of coolant are properly identified in the Technical 
Specifications.
    Although pressure isolation valves are being added to the 
Technical Specification table, these valves were already included in 
the IST [inservice testing] program as pressure isolation valves and 
were being tested as such. The proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

[[Page 61846]]

    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change, as described above, does not physically 
alter the facility or the operation of the facility. The majority of 
the changes are editorial in nature and provide for improvement in 
the human factors of the Technical Specifications, while properly 
identifying all the pressure isolation valves in the Technical 
Specifications. The addition of valves into the Technical 
Specification is an administrative change that improves the quality 
of the LCO [limiting condition for operation], but does not add 
components to the facility.
    The proposed changes do 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 margin of safety, as defined in the bases for any technical 
specifications, depend upon proper identification of equipment and 
performance of the proper surveillance requirements to demonstrate 
equipment operability. The proposed change will ensure that the 
proper valves are identified and tested in accordance with the 
Technical Specification requirements.
    The proposed changes 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: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038
    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: October 21, 1997
    Description of amendment request: The proposed amendment revises 
Technical Specification Tables 3.3-1 and 4.3-1 to require that 
Functional Unit, <gr-thn-eq>2. Power Range, Neutron Flux,'' be operable 
in Mode 3, as well as in Modes 1 and 2. The change is being proposed 
because the licensee has determined that the power range nuclear 
instrumentation should be operable in Mode 3 whenever the reactor trip 
system breakers are in the closed position and the control rods are 
capable of being withdrawn.
    Basis for proposed no significant hazards 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 requirement for operability of a trip and the surveillance 
requirements to ensure the functionality of the trip are independent 
of the probability of an accident previously evaluated. The accident 
that this trip is intended to mitigate is the Rod Withdrawal from 
Subcriticality event. The surveillance procedure and the requirement 
for the trip to be operational when the Control Rod Drive System is 
capable of rod movement mitigate the consequences of this event, and 
do not increase the probability of a rod withdrawal from 
subcritical.
    Therefore, the probability 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, do not introduce any new operating configurations or new 
modes of plant operation, or change the safety analyses. The 
proposed change is intended to ensure that the trip function is 
available and will perform as designed in the event of a previously 
evaluated event.
    The proposed 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 change does not reduce the margin of safety, 
because assurance of the operability of the trip function is 
increased by the proposed change.
    Based on the above, PSE&G [Public Service Electric & Gas 
Company] has determined that the proposed changes do not involve a 
significant hazards consideration.
    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
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038
    NRC Project Director: John F. Stolz

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 1, Lake County, Ohio

    Date of amendment request: August 28, 1997
    Description of amendment request: The proposed change would revise 
Technical Specification 3.4.11, ``Reactor Coolant System (RCS) Pressure 
and Temperature (P/T) Limits,'' to incorporate the new P/T curves, 
which were provided by General Electric Nuclear Energy in report number 
GE-NE-B1301793-01, ``Perry Unit 1 RPV Surveillance Materials Testing 
and Analysis.''
    Basis for proposed no significant hazards 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 change will provide for approved P/T limit curves 
which are valid through 9 effective full-power years (EFPY) and 18 
EFPY. This change will not affect any Safety Limits, Power 
Distribution Limits, or Limiting Conditions for Operation. The 
proposed changes incorporate operating limits which provide margin 
to brittle failure of the reactor vessel based on testing of the 
irradiated reactor vessel materials (base metal, weld material, and 
heat affected zone material). The limits ensure that adequate safety 
margins against nonductile or rapidly propagating failure exist 
during normal operation, anticipated operational occurrences, and 
system hydrostatic tests. The specimens have been tested and 
analyzed in accordance with 10 CFR 50, Appendices G and H, using the 
methods described in Generic Letter 88-11 and Regulatory Guide 1.99 
Revision 2. The predicted lowest upper shelf energy at 32 EFPY was 
greater than the minimum required by 10 CFR 50, Appendix G. The 
adjusted reference temperature for the limiting material was lower 
than the 200 degree Fahrenheit limit required by Regulatory Guide 
1.99 Revision 2. As such, the integrity of the reactor pressure 
coolant boundary is maintained. The changes will result in 
equivalent or more conservative limits on reactor vessel pressure as 
a function of temperature for all operational conditions 
(hydrostatic and leak testing, non-nuclear heatup/cooldown, and core 
critical operations). The methodology used to derive these values 
produces limits which continue to ensure that sufficient margin is 
maintained to meet the criteria of GDC 31, ``Fracture Prevention of 
Reactor Coolant Pressure Boundary.'' There are no plant 
modifications associated with this change and no new or revised 
system interfaces. The proposed

[[Page 61847]]

changes do not increase the probability of occurrence or 
consequences previously evaluated because the temperature shifts are 
well within equipment operating ranges. As such, there is no 
increase in the probability of occurrence or 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.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed changes do not involve any new modes of 
operation. The only change will be operation of the plant within 
operating pressure limits which are determined in a more 
conservative manner. Therefore, no new failure mode or accident 
sequence is introduced by this change.
    The testing and analysis meets 10 CFR 50, Appendices G and H, 
requirements; therefore, no new accident types, such as brittle 
fracture of a reactor pressure coolant boundary component is 
postulated. The adjusted reference temperature and upper shelf 
energy predicted at 32 EFPY are well within the limits of 10 CFR 50, 
Appendices G and H. Therefore, the possibility of an accident of a 
new or different type than any previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The P/T limits are established to provide acceptable margins for 
the operation of the reactor coolant system during heat up and cool 
down, criticality, and hydrotest conditions. Technical Specification 
3.4.11 limits the rates of change of temperature and pressure to 
values consistent with the fracture toughness requirements of 10 CFR 
50, Appendices G and H, and ASME Boiler and Pressure Vessel Code 
Section III Appendix G. The bases section for Technical 
Specification 3.4.11 refers to 10 CFR 50, Appendices G and H, and 
ASME Code Section III Appendix G. Changes in these limits are 
necessary because the fracture toughness properties of ferritic 
materials in the reactor vessel change as a function of reactor 
operating time. The specific requirements for fracture toughness and 
reactor vessel material surveillance that must be considered in 
developing the P/T limits are defined by 10 CFR 50, Appendices G and 
H. The specific limits defined by 10 CFR 50, Appendices G and H, set 
the margin of safety for the reactor pressure vessel coolant 
boundary. Since the testing and analysis of the vessel specimens 
meet the requirements and limits defined in 10 CFR 50, Appendices G 
and H, the margin of safety as defined in the basis for Technical 
Specification 3.4.11 is not reduced. The revised curves are based on 
the latest NRC guidelines along with actual neutron fluence data for 
Perry. The new limits conservatively account for irradiation 
embrittlement effects, thereby maintaining 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: Perry Public Library, 3753 
Main Street, Perry, OH 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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 1, Lake County, Ohio

    Date of amendment request: September 8, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.2.2.e, ``Organization - Unit 
Staff,'' by removing the reference to the NRC Policy Statement on 
working hours. Administrative procedures will be developed to limit the 
working hours of unit staff who perform safety-related functions.
    Basis for proposed no significant hazards 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 change to TS 5.2.2.e only alters the administrative 
location of and the regulatory controls applicable to unit staff 
specific overtime limits and working hours. Overtime will remain 
controlled by plant administrative procedures. Changes to the 
relocated overtime limits and working hours will be subject to 
review and evaluation under 10 CFR 50.59, ``Changes, Tests and 
Experiments.'' There is not an increase in the probability of an 
accident previously evaluated because no change is being made to any 
accident initiator. No previously analyzed accident scenario is 
changed, and initiating conditions and assumptions remain as 
previously analyzed.
    There is not an increase in the radiological consequences of an 
accident previously evaluated because the proposed change does not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The proposed change does 
not alter the source term, containment isolation, or allowable 
radiological releases. Therefore, there is no increase in the 
radiological consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility or a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed change does not change the way the plant is 
operated, and no new or different failure modes have been defined 
for any plant system or component important to safety, nor has any 
limiting single failure been identified as a result of the proposed 
change. No new or different types of failures or accident initiators 
are introduced by the proposed change.
    The proposed change to TS 5.2.2.e only alters the administrative 
location of and the regulatory controls applicable to unit staff 
specific overtime limits and working hours. Therefore, there is no 
possibility created for a new or different kind of accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not involve a reduction in a margin of 
safety because unit staff overtime is not an input in the 
calculation of a safety margin with regard to Technical 
Specification Safety Limits, Limiting Safety System Settings, other 
Technical Specification Limiting Conditions for Operation, the 
Operational Requirements Manual, or other previously defined margins 
for any structure, system, or component important to safety. The 
proposed change to TS 5.2.2.e only alters the administrative 
location of and the regulatory controls applicable to unit staff 
specific overtime limits and working hours.
    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, OH 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Units 1 and 2, Somervell County, Texas

    Date of amendment request: October 24, 1997
    Brief description of amendments: Change to the core safety limit 
curves and overtemperature N-16 reactor trip function setpoints to 
support operation with Unit 1, cycle 7 core configuration.
    Basis for proposed no significant hazards 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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    A. Revision to the Unit 1 Core Safety Limits

[[Page 61848]]

    Analyses of reactor core safety limits are required as part of 
reload calculations for each cycle. TU Electric has performed the 
analyses of the Unit 1, Cycle 7 core configuration to determine the 
reactor core safety limits. The methodologies and safety analysis 
values result in new operating curves which, in general, permit 
plant operation over a similar range of acceptable conditions. This 
change means that if a transient were to occur with the plant 
operating at the limits of the new curve, a different temperature 
and power level might be attained than if the plant were operating 
within the bounds of the old curves.
    However, since the new curves were developed using NRC approved 
methodologies which are wholly consistent with and do not represent 
a change in the Technical Specification BASES for safety limits, all 
applicable postulated transients will continue to be properly 
mitigated. As a result, there will be no significant increase in the 
consequences, as determined by accident analyses, of any accident 
previously evaluated.
    B. Revision to Unit 1 Overtemperature N-16 Reactor Trip 
Setpoints
    As a result of changes discussed, the Overtemperature reactor 
trip setpoint has been recalculated. These trip setpoints help 
ensure that the core safety limits are protected and that all 
applicable limits of the safety analysis are met.
    Based on the calculations performed, no significant changes to 
the safety analysis values for Overtemperature reactor trip setpoint 
were required. The f(deltaI) trip reset function was revised due to 
more top-skewed axial power distributions predicted for this cycle. 
The analyses performed show that, using the TU Electric 
methodologies, all applicable limits of the safety analysis are met. 
This setpoint provides a trip function which allows the mitigation 
of postulated accidents and has no impact on accident initiation. 
Therefore, the changes in safety analysis values do not involve an 
increase in the probability of an accident and, based on satisfying 
all applicable safety analysis limits, there is no significant 
increase in the consequences of any accident previously evaluated.
    In addition, sufficient operating margin has been maintained in 
the overtemperature setpoint such that the risk of turbine runbacks 
or reactor trips due to upper plenum flow anomalies or other 
operational transients will be minimized, thereby, reducing 
potential challenges to the plant safety systems.
    SUMMARY
    The changes in the amendment request applies NRC approved 
methodologies to changes in safety analysis values, new core safety 
limits and new N-16 setpoint and parameter values to assure that all 
applicable safety analysis limits have been met. The potential for 
an operational transient to occur has not been affected and there 
has been no significant impact on the consequences of any accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve the calculation of new reactor core 
safety limits and overtemperature reactor trip setpoint resets. As 
such, the changes play an important role in the analysis of 
postulated accidents but none of the changes effect plant hardware 
or the operation of plant systems in a way that could initiate an 
accident. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    In reviewing and approving the methods used for safety analyses 
and calculations, the NRC has approved the safety analysis limits 
which establish the margin of safety to be maintained. While the 
actual impact on safety is discussed in response to question 1, the 
impact on margin of safety is discussed below:
    A. Revision to the Unit 1 Reactor Core Safety Limits
    The NRC-approved TU Electric reload analysis methods have been 
used to determine new reactor core safety limits. All applicable 
safety analysis limits have been met. The methods used are wholly 
consistent with Technical Specification BASES 2.1 which is the bases 
for the safety limits. In particular, the curves assure that for 
Unit 1, Cycle 7, the calculated DNBR is no less than the safety 
analysis limit and the average enthalpy at the vessel exit is less 
than the enthalpy of saturated liquid. The acceptance criteria 
remains valid and continues to be satisfied; therefore, no change in 
a margin of safety occurs.
    B. Revision to Unit 1 Overtemperature N-16 Reactor Trip 
Setpoints
    Because the reactor core safety limits for CPSES Unit 1, Cycle 7 
are recalculated, the Reactor Trip System instrumentation setpoint 
values for the Overtemperature N-16 reactor trip setpoint which 
protect the reactor core safety limits must also be recalculated. 
The Overtemperature N-16 reactor trip setpoint helps prevent the 
core and Reactor Coolant System from exceeding their safety limits 
during normal operation and design basis anticipated operational 
occurrences. The most relevant design basis analysis in Chapter 15 
of the CPSES Final Safety Analysis Report (FSAR) which is affected 
by the Overtemperature reactor trip setpoint is the Uncontrolled Rod 
Cluster Control Assembly Bank Withdrawal at Power (FSAR Section 
15.4.2). This event has been analyzed with the new safety analysis 
value for the Overtemperature reactor trip setpoint to demonstrate 
compliance with event specific acceptance criteria. Because all 
event acceptance criteria are satisfied, there is no degradation in 
a margin of safety.
    The nominal Reactor Trip System instrumentation setpoints values 
for the Overtemperature N-16 reactor trip setpoint (Technical 
Specification Table 2.2-1) are determined based on a statistical 
combination of all of the uncertainties in the channels to arrive at 
a total uncertainty. The total uncertainty plus additional margin is 
applied in a conservative direction to the safety analysis trip 
setpoint value to arrive at the nominal and allowable values 
presented in Technical Specification Table 2.2-1. Meeting the 
requirements of Technical Specification Table 2.2-1 assures that the 
Overtemperature reactor trip setpoint assumed in the safety analyses 
remains valid. The CPSES Unit 1, Cycle 7 Overtemperature reactor 
trip setpoint is not significantly different from the previous 
cycle, and thus provides operational flexibility to withstand mild 
transients without initiating automatic protective actions. Although 
the value of the f(deltaI) trip reset function setpoint is 
different, the Reactor Trip System instrumentation setpoint values 
for the Overtemperature N-16 reactor trip setpoint are consistent 
with the safety analysis assumptions which have been analytically 
demonstrated to be adequate to meet the applicable event acceptance 
criteria. Thus, there is no reduction in a margin of safety.
    Using the NRC approved TU Electric methods, the reactor core 
safety limits are determined such that all applicable limits of the 
safety analyses are met. Because the applicable event acceptance 
criteria continue to be met, there is no 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 Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: James W. Clifford, Acting

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.

[[Page 61849]]

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: October 13, 1997
    Description of amendment request: The proposed amendments would 
support replacement of the three safety-related wide range level 
instruments. The engineered safety features trip setpoint for the 
refueling water automatic switchover to recirculation would be revised 
to account for the difference in instrument uncertainty associated with 
wide range level instruments and provide additional response time 
margin.
    Date of publication of individual notice in Federal Register: 
October 22, 1997 (62 FR 54859)
    Expiration date of individual notice: November 21, 1997
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina

Duke Energy Corporation, 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: October 20, 1997
    Description of amendment request: The proposed amendments would 
allow use of a rerolling process as an additional repair method for 
tube degradation found in the tubesheet region. The rerolling method is 
designed to ensure that the area of degradation will not serve as a 
pressure boundary once the repair roll is installed, thus permitting 
the tube to remain in service.
    Date of publication of individual notice in Federal Register: 
October 28, 1997 (62 FR 55835)
    Expiration date of individual notice: Comment period ends November 
12, 1997; Notice period ends November 28, 1997
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina

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 Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of application for amendments: October 4, 1997
    Brief description of amendments: These amendments revise the 
surveillance requirements in Technical Specifications (TSs) 4.1.2.3.1, 
4.1.2.4.1, 4.5.2.b, and 4.6.2.1.b and associated Bases. The subject 
surveillance requirements are applicable to the charging/high-head 
safety injection pumps, low-head safety injection pumps, and the 
containment quench spray pumps. The proposed changes replace the 
current specific test acceptance criteria contained in these 
surveillance requirements with requirements to verify pump performance 
in accordance with the inservice testing program, the emergency core 
cooling system flow analysis, or the containment integrity safety 
analysis, as applicable. The proposed changes also make minor editorial 
changes in these TSs and make conforming changes in the TS Index pages.
    Date of issuance: October 28, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 207, 86
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66706) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 28, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: May 27, 1997, supplemented by 
October 6, 1997
    Brief description of amendment: The amendment eliminated selected 
response time testing (RTT) surveillance requirements (SRs) from the 
Technical Specifications (TSs) for certain components of the following 
systems: reactor protection system (SR 3.3.1.1.15), primary containment 
and drywell isolation instrumentation (SR 3.3.6.1.8), and emergency 
core cooling system (SRs 3.5.1.8 and 3.5.2.7).
    Date of issuance: November 5, 1997
    Effective date: November 5, 1997
    Amendment No.: 133
    Facility Operating License No. NPF-29: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33122) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 5, 1997.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 17, 1996, as supplemented October 
14, 1997
    Brief description of amendment: The amendment revises

[[Page 61850]]

    Facility Operating License No. NPF-38 to reflect the name change 
from Louisiana Power & Light Company to Entergy Louisiana, Inc.
    Date of issuance: November 3, 1997
    Effective date: November 3, 1997, to be implemented within 60 days.
    Amendment No.: 134
    Facility Operating License No. NPF-38: Amendment revised
    Facility Operating License No. NPF-38.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
29749) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 3, 1997. The letter 
dated October 14, 1997, provided clarifying information which did not 
alter the initial no significant hazards determination. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: March 27, as supplemented April 
3, May 1, and August 20, 1997.
    Brief description of amendment: Change Technical Specifications 
(TS) to permanently establish a primary-to-secondary leak rate of 150 
gallons per day through any one steam generator and specify the steam 
generator tube inservice inspection requirements for pit-like 
intergranular attack degradation in the ``B'' Once-Through-Steam-
Generator.
    Date of issuance: October 28, 1997
    Effective date: October 28, 1997
    Amendment No.: 158
    Facility Operating License No. DPR-72: Amendment revised the TS.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30632) The August 20, 1997, letter provided clarifying information that 
did not affect the initial no significant hazards consideration. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated October 28, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: August 27, 1997
    Brief description of amendments: The admendments change the 
Administrative Section of the Technical Specifications (TS) to allow 
the use of 12-hour shifts.
    Date of issuance: October 27, 1997
    Effective date: October 27, 1997
    Amendment Nos: 194 and 188Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the TS.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50006) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 27, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: March 26, 1997.
    Brief description of amendments: The amendments modify surveillance 
4.7.5.1.e.2 which requires verification of the control room ventilation 
system autostart function.
    Date of issuance: October 28, 1997
    Effective date: October 28, 1997, with full implementation within 
45 days.
    Amendment Nos.: 218 and 202
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27796) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 28, 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

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone 
Nuclear Power Station, Unit 1, New London County, Connecticut

    Date of application for amendment: May 15, 1997
    Brief description of amendment: The amendment revises Technical 
Specification Sections 3.1 and 4.1, ``Reactor Protection System,'' and 
the associated Bases to remove run mode intermediate range monitor high 
flux/inoperative with the associated average power range monitor 
downscale scram trip function. The amendment also makes other editorial 
revisions.
    Date of issuance: October 27, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 103
    Facility Operating License No. DPR-21: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33127) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 27, 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 at the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut

    Date of application for amendment: September 2, 1997
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by modifying the maximum allowed primary 
containment internal pressure during normal operation from 2.1 pounds 
per square inch gauge (psig) to 1.0 psig. The TS Bases, Section 3/
4.6.1.4, is also updated to reflect the new maximum allowed primary 
containment internal pressure during normal operation.
    Date of issuance: October 27, 1997
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 209
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50007) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 27, 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, and Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, Connecticut

[[Page 61851]]

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of application for amendment: November 25, 1996, as 
supplemented December 12, 1996, April 23, May 8, July 1, August 21, and 
September 29, 1997
    Brief description of amendment: The amendment modifies the 
Technical Specification requirements associated with the Minimum 
Critical Power Ratio (MCPR) safety limits for Cycle 18 based on the 
cycle-specific analysis of the current mixed core of GE11/GE10 fuel 
parameters.
    Date of issuance: October 29, 1997
    Effective date: October 29, 1997
    Amendment No.: 99
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17238) The December 12, 1996, letter provided an affidavit for the 
original application dated November 25, 1996. The April 23, May 8, 
August 21, and September 29, 1997, letters provided clarifying 
information in response to the staff's request for additional 
information during a teleconference on March 18, 1997. The July 1, 
1997, letter provided a nonproprietary version of the April 23, 1997, 
submittal. This information was within the scope of the original 
application and did not change the staff's initial proposed no 
significant hazards considerations determination. Therefore, renoticing 
was not warranted. The Commission's related evaluation of the amendment 
is contained in a Safety Evaluation dated October 29, 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

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: November 27, 1996, as 
supplemented August 15, September 2, and October 3, 1997
    Brief description of amendments: The amendments incorporate 
Combustion Engineering steam generator tube sleeve designs and 
installation and examination techniques into the plant Technical 
Specifications (TS). Specifically, the amendments make changes to TS 
4.12, ``Steam Generator Tube Surveillance,'' and its associated Bases 
Section B.4.12, ``Steam Generator Tube Surveillance.''
    Date of issuance: November 4, 1997
    Effective date: November 4, 1997, with full implementation within 
30 days
    Amendment Nos.: 132 and 124
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43370) The August 15, September 2, and October 3, 1997, letters 
provided clarifying information and updated TS pages. This information 
was within the scope of the original application and did not change the 
staff's initial no significant hazards considerations determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated November 4, 1997.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

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: June 4, 1997
    Brief description of amendments: The proposed change revises the 
Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, Technical 
Specifications to eliminate an inconsistency between emergency core 
cooling system (ECCS) operability requirements and the auto-start and 
protective trip bypass of the emergency diesel generators on an ECCS 
initiation signal during certain plant configurations.
    Date of issuance: October 24, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 30.
    Amendments Nos.: 221 and 226
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43373) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 24, 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

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: May 19, 1997, as supplemented by 
letter dated August 25, 1997
    Brief description of amendment: This amendment changes the Hope 
Creek Technical Specification (TS) 3.7.1.3, ``Ultimate Heat Sink,'' to 
raise the minimum allowable ulimate heat sink (UHS) water level from 76 
feet to 80 feet, lower the maximum allowable UHS temperature from 
88.6 deg.F to 85 deg.F, and reflect that continued plant operation to a 
UHS temperature of 87 deg.F depends upon the association of UHS 
temperature and safety system availability. The associated Surveillance 
Requirement, TS 4.7.1.3, is changed to decrease the river water 
temperature, at which increasing temperature surveillance is required, 
from 85 deg.F to 82 deg.F. The requirements of TS 3.7.1.1, ``Safety 
Auxiliaries Cooling System (SACS),'' TS 3.7.1.2, ``Station Service 
Water System (SSWS),'' and TS 3.8.1.1, ``Electrical Power Systems,'' 
are revised to reflect the revised TS 3.7.1.3. In addition, the Bases 
for 3/4.7.1, ``Service Water Systems,'' are appropriately revised.
    Date of issuance: October 28, 1997
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 106
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33132) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 28, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: March 31, 1997, as supplemented 
by letters dated July 16, August 26, and October 3, 1997
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 2.1.2, ``THERMAL POWER, High Pressure and High 
Flow,''

[[Page 61852]]

ACTION a.1.c for TS 3.4.1.1, ``Recirculation Loops,'' and the Bases for 
TS 2.1, ``Safety Limits.'' These changes are being made to implement an 
appropriately conservative Safety Limit Minimum Critical Power Ratio to 
include Cycle 8 specific analyses for all Hope Creek core and fuel 
designs.
    Date of issuance: November 4, 1997
    Effective date: The license amendment is effective as of its date 
of issuance and shall be implemented within 60 days.
    Amendment No.: 107
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43374) The August 26 and October 3, 1997, letters provided clarifying 
information that did 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 
November 4, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location:Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: September 29, 1997
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 3/4.11.1, ``Liquid Effluent - Concentration.'' The 
change adds a requirement to perform weekly sampling and monthly and 
quarterly composite analyses of the Station Service Water System when 
the Reactor Auxiliaries Cooling System is contaminated. The licensee 
has also proposed an editorial change to TS Table 4.11.1.1.1-1. In 
Liquid Release Type B, the licensee is proposing that the acronym for 
Station Service Water System be changed from GSW to SSWS. This proposed 
change will be addressed in a future license amendment.
    Date of issuance: November 6, 1997
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 108
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 6, 1997 (62 FR 
52161) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 6, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: September 24, 1997
    Brief description of amendment: This amendment adds a Surveillance 
Requirement to Technical Specification 3/4.5.1, ``Emergency Core 
Cooling Systems'', to perform a monthly valve position verification for 
the four residual heat removal cross-tie valves.
    Date of issuance: November 6, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 109
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 6, 1997 (62 FR 
52162) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 6, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: August 20, 1997
    Brief description of amendment: This amendment changes the 
Technical Specifications (TSs) to provide for: 1) the relocation of 
suppression chamber volume references in Limiting Condition for 
Operation (LCO) 3.5.3 to the Hope Creek (HC) Updated Final Safety 
Analysis Report (UFSAR) and TS Bases as appropriate; 2) the revision of 
the suppression chamber volume currently listed in LCO 3.5.3.b; 3) the 
relocation of the suppression chamber volume references in LCO 
3.6.2.1.a.1 to the UFSAR and TS Bases; and 4) the revision to the 
suppression chamber volume reference in TS 5.2.1 to reference the TS 
Bases section where this information will reside.
    Date of issuance: November 6, 1997
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 110
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50010) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 6, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: August 19, 1997, as supplemented 
September 29, 1997.
    Brief description of amendment: The proposed amendment revises the 
Ginna Station Improved Technical Specifications by adding a note to the 
Containment Spray (CS) Limiting Condition for Operation 3.6.6 which 
allows the CS pumps in MODE 4 to be placed in pull-stop, and motor-
operated valves 896A and 896B to have their DC control power restored 
with the valves placed in the closed position in order to perform 
interlock and valve testing of MOVs 857A, 857B, and 857C. A time limit 
of 2 hours is placed on this configuration for each test.
    Date of issuance: October 29, 1997
    Effective date: October 29, 1997
    Amendment No.: 68
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50011) The September 29, 1997, letter provided clarifying 
information that did 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 
October 29, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama

    Date of amendments request: September 17, 1997
    Brief description of amendments: The amendments change Technical 
Specification 3/4.4.9, ``Specific Activity,'' and the associated Bases 
to reduce the limit associated with dose equivalent iodine-131. The 
steady-state dose equivalent iodine-131 limit would

[[Page 61853]]

be reduced by 50 percent from 0.3 mu Curie/gram to 0.15 mu Curie/gram 
and the maximum instantaneous value would be reduced by 50 percent from 
18 mu Curie/gram to 9 mu Curie/gram.
    Date of issuance: October 29, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: Unit 1 - 132; Unit 2 - 124
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 49998) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 29, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear, 
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 
50-440 Perry Nuclear Power Plant, Unit 1, Lake County, Ohio

    Date of application for amendment: August 14, 1997, as supplemented 
September 26 and October 1, 1997.
    Brief description of amendment: This amendment changes the design 
basis as described in the Updated Safety Analysis Report by adding a 
description of the methodology utilized for determining the systems and 
components that are considered to require protection from tornado 
missiles.
    Date of issuance: November 4, 1997
    Effective date: November 4, 1997
    Amendment No.: 90
    Facility Operating License No. NPF-58: This amendment revised the 
license.
    Date of initial notice in Federal Register: September 16, 1997 (62 
FR 48674). The September 26 and October 1, 1997, submittals provided 
supplemental information that did 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 November 4, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: May 16, 1997 (TXX-97119)
    Brief description of amendments: The amendments revised core safety 
limit curves and Overtemperature N-16 reactor trip setpoints based on 
analyses of the core configuration for CPSES Unit 2, Cycle 4. These 
changes apply equally to CPSES Units 1 and 2 licenses since the 
Technical Specifications are combined.
    Date of issuance: October 30, 1997
    Effective date: October 30, 1997
    Amendment Nos.: Unit 1 - Amendment No. 55; Unit 2 - Amendment No. 
41
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38140) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 30, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019

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 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

[[Page 61854]]

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 December 19, 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.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to 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).

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of application for amendment: October 24, 1997
    Brief description of amendment: The amendment adds a footnote to 
Technical Specification 3.7.A.5, ``Primary Containment.'' The footnote 
provides a one time exception to the reverse flow testing requirement 
for containment isolation check valve 30-CK-432.
    Date of issuance: October 30, 1997
    Effective date: As of date of issuance and shall be implemented by 
November 2, 1997.
    Amendment No.: 174
    Facility Operating License No. DPR-35: This amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, consultation with the State of Massachusetts, 
and final no significant hazards consideration determination are 
contained in a Safety Evaluation dated October 30, 1997.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360
    NRC Project Director: Ronald B. Eaton, Acting Director

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: October 17, 1997

[[Page 61855]]

    Brief description of amendment: The amendment revised Technical 
Specification 4.5.2b and associated Bases to eliminate the requirement 
to vent the centrifugal charging pump casings.
    Date of issuance: November 3, 1997
    Effective date: November 3, 1997
    Amendment No.: 114
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications. Press release issued requesting comments as 
to proposed no significant hazards consideration: Yes. October 24, 
1997. Coffey County Today Newspaper (Kansas). Comments received: Yes. 
Comments were submitted by Mr. Dave Lochbaum of the Union of Concerned 
Scientists by letter dated October 29, 1997. Verbal comments were 
received from Larry Myers on October 28, 1997. The staff responded to 
these comments in the safety evaluation attached to the November 3, 
1997, amendment. The Commission's related evaluation of the amendment, 
finding of exigent circumstances, consultation with the State of Kansas 
and final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated November 3, 1997.
    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
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman
    Dated at Rockville, Maryland, this 12th day of November 1997.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV Office of 
Nuclear Reactor Regulation.
[FR Doc. 97-30217 Filed 11-18-97; 8:45 am]
BILLING CODE 7590-01-F