[Federal Register Volume 63, Number 87 (Wednesday, May 6, 1998)]
[Notices]
[Pages 25101-25129]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-11911]


-----------------------------------------------------------------------

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. Pub. L. 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 April 10 through April 24, 1998. The last 
biweekly notice was published on April 22, 1998 (63 FR 19964).

Notice of Consideration of Issuance of Amendments to Facility 
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 
expiration of the 30-day notice period.

[[Page 25102]]

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 Administration 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 June 5, 1998, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room 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 25103]]

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendment request: November 1, 1996, as supplemented by 
letters dated October 13, 1997, February 26, 1998, and March 13, 1998.
    Description of amendment request: Associated with a Carolina Power 
& Light Company (the licensee) application to convert from the Current 
Technical Specifications (CTS) for the Brunswick Steam Electric Plant, 
Units 1 and 2, to Improved Technical Specifications (ITS), as contained 
in Revision 1 of NUREG-1433, ``Standard Technical Specification General 
Electric Plants, BWR/4,'' the licensee proposed removing a restriction 
on a surveillance test described below.
    CTS 4.8.1.1.1.b requires that the offsite electrical power circuits 
be demonstrated OPERABLE, at least once per 18 months during shut down, 
by manually transferring the unit power supply from the normal circuit 
to the alternate circuit. As proposed, ITS SR 3.8.1.8.b will not 
contain the restriction to perform the Surveillance ``during 
shutdown.'' Currently, this test is performed by momentarily 
paralleling the 230 kV offsite alternating current (AC) power sources. 
The licensee has stated that paralleling offsite AC power sources is a 
controlled evolution and the increased risk associated with the 
performance of this test while the unit is at power is not significant 
for the following reasons: (1) the frequency and voltages are verified 
to be within the required range prior to paralleling the two offsite AC 
power sources; (2) breaker interlocks ensure that the alternate circuit 
is connected to the load prior to opening the preferred circuit; (3) 
the test does not result in de-energization of any 4.16 kV emergency 
bus and the potential for electrical perturbations on the grid system 
is the same whether performing the transfer while the unit is at power 
or while shutdown; and (4) operating history indicates that 
transferring offsite AC power sources while the units were in 
Operational Conditions 1 (power operation) or 2 (startup) has been 
performed satisfactorily without electrical distribution system 
perturbations. The licensee has further pointed out that Generic Letter 
91-04, ``Changes in Technical Specifications to Accommodate a 24-Month 
Fuel Cycle,'' states that licensees may omit the Technical 
Specification qualification that a refueling interval surveillance is 
to be performed ``during shutdown.'' Therefore, consistent with the 
guidance provided in Generic Letter 91-04, the licensee proposed 
deletion of the requirement to perform this Surveillance ``during 
shutdown'' as part of the conversion from CTS 4.8.1.1.1.b to ITS SR 
3.8.1.8.b.
    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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change would remove a specific restriction to perform the 
verification of the manual transfer of the unit power supply from 
the normal circuit to the alternate circuit ``during shutdown.'' The 
transfer of the unit power supply from the normal circuit to the 
alternate circuit is not an initiator of any previously analyzed 
accident. Therefore, this change does not significantly increase the 
frequency of such accidents. Currently, this test is performed by 
momentarily paralleling the 230 kV offsite AC power sources. 
Paralleling offsite AC power sources is a controlled evolution and 
the increased risk associated with the performance of this test 
while the unit is at power is not significant for the following 
reasons: (1) The frequency and voltages are verified to be within 
the required range prior to paralleling the two offsite AC power 
sources; (2) breaker interlocks ensure that the alternate circuit is 
connected to the load prior to opening the preferred circuit; (3) 
the test does not result in de-energization of any 4.16 kV emergency 
bus and the potential for electrical perturbations on the grid 
system is the same whether performing the transfer while the unit is 
at power or while shutdown; and (4) operating history indicates that 
transferring offsite AC power sources while the units were in MODE 
(Operational Condition) 1 or 2 has been performed satisfactorily 
without electrical distribution system perturbations. The 
appropriate plant conditions for performance of the Surveillance 
will continue to be controlled to assure the potential consequences 
are not significantly increased. This control method has been 
previously determined to be acceptable as indicated in Generic 
Letter 91-04. Therefore, this change does not significantly increase 
the consequences of any previously analyzed accident.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This change removes a specific restriction on the plant 
conditions for performing a Surveillance, but does not change the 
method of performance. The appropriate plant conditions for 
performance of the Surveillance will continue to be controlled to 
assure the possibility for a new or different kind of accident are 
not created. This control method has been previously determined to 
be acceptable as indicated in Generic Letter 91-04. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any previously analyzed accident.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The margin of safety considered in determining the appropriate 
plant conditions for performing the Surveillance will continue to be 
controlled to assure that there is no significant reduction. This 
control method has been previously determined to be acceptable as 
indicated in Generic Letter 91-04. Therefore, the 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 
amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina at 
Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297
Attorney for licensee: William D. Johnson, Vice President and Senior 
Counsel, Carolina Power & Light Company, Post Office Box 1551, Raleigh, 
North Carolina 27602
NRC Project Director: Pao-Tsin Kuo

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of amendment request: April 3, 1998.
    Description of amendment request: The Carolina Power & Light 
Company, licensee for the Brunswick Steam Electric Plant (BSEP), Unit 
Nos. 1 and 2, proposed amendments to the Technical Specifications (TS) 
to change the specified total volume of the condensate storage tank 
(CST) from 150,000 gallons to 228,200 gallons. During a recent review 
of industry operating experience, the licensee determined that 
information contained in TS 3.5.3.1, Core Spray System (CSS), and the 
associated bases regarding water inventory in the CST was incorrect. 
Specifically, the minimum CST volume requirement contained in TS 
3.5.3.1 would not assure the availability of 50,000 gallons of water 
for the CSS, as indicated in TS Bases section 3/4.5.3.1 for the CSS.
    The licensee has concluded that the proposed license amendments do 
not involve a Significant Hazards Consideration. In support of this 
determination, an evaluation of each of the three standards set forth 
in 10 CFR 50.92 is provided below.

[[Page 25104]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed TS change revises the minimum CST [Condensate 
Storage Tank] water volume required for OPERABILITY of the Core 
Spray system (CSS) in OPERATIONAL CONDITIONS 4 AND 5 when the 
suppression pool is inoperable. The proposed change does not alter 
the operation of any plant system or component; does not involve a 
physical modification to any structure, system, or component; and 
does not affect an initiator to any accident previously evaluated. 
The minimum CST water level is being increased to assure the 
availability of 50,000 gallons of water for use by the CSS. 
Therefore, the proposed license amendments do not involve an 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. This proposed TS change revises the minimum 
CST water volume required for OPERABILITY of the CSS in OPERATIONAL 
CONDITIONS 4 and 5 when the suppression pool is inoperable. The 
proposed change does not alter the operation of any plant system or 
component; does not involve a physical modification to any 
structure, system, or component; and does not affect an initiator to 
any accident previously evaluated. The proposed change does not add 
or modify equipment or components related to the CSS and will, 
therefore, not create new failure modes or common failure modes. The 
minimum CST water level is being increased to assure the 
availability of 50,000 gallons of water for use by the CSS. 
Therefore, the proposed license amendments do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety. The proposed license amendments 
increase the minimum CST water level to assure the availability of 
50,000 gallons of water for use by the CSS. These volumes ensure the 
validity of existing analyses, and ensure that the existing TS Bases 
are satisfied. The proposed change does not involve a physical 
modification to any structure, system, or component, and does not 
modify the operation of any existing equipment. Therefore, the 
proposed license amendments do not involve a reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina at 
Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297
Attorney for licensee: William D. Johnson, Vice President and Senior 
Counsel, Carolina Power & Light Company, Post Office Box 1551, Raleigh, 
North Carolina 27602
NRC Project Director: Pao-Tsin Kuo (Acting)

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities 
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

    Date of amendment request: March 31, 1998.
    Description of amendment request: Unreviewed Safety Question 
involving use of Station Blackout (SBO) diesel generators (DGs) and use 
of a mobile safe shutdown (SSD) battery cart in the 10 CFR part 50, 
appendix R, Safe Shutdown Safety Analysis.
    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:
    The licensee has provided a separate no significant hazards 
consideration determination for the SBO DGs and the battery cart under 
this amendment request. The following is the determination for the SBO 
DGs:

    (1) No significant increase in the probability or consequences 
of an accident previously evaluated is involved because of the 
following:
    Two types of previously evaluated accidents are relevant to this 
criterion: (1) A fire; (2) other accident evaluated in the UFSAR. 
For these previously evaluated accidents, the change would not 
result in an increase in either their probabilities of occurrence or 
the consequences of their occurrence, for the following reasons.
    The use of the SBO DGs in lieu of the [Emergency Diesel 
Generators] EDGs does not change the probability or consequences of 
a fire. The likelihood of a fire is unchanged. Use of the SBO DGs 
does not significantly change the fire loading nor introduce 
significant new ignition sources. The consequences of a fire are 
unchanged because use of the SBO DGs continues to support the 
station's ability to achieve and maintain shutdown in the event of a 
fire.
    Use of the SBO DGs for non-fire purposes is unchanged by use of 
the SBO DGs for post-fire safe shutdown in the event of a fire in 
areas requiring alternate shutdown capability. Accordingly there is 
no change in the probability or consequences of a previously 
evaluated accident involving the SBO DGs. Similarly, there is no 
change to the probability or consequences of other accidents that 
have been previously evaluated because they are independent of this 
change in use of the SBO DGs.
    (2) The possibility of a new or different kind of accident from 
any accident previously evaluated is not created because:
    The proposed change does not create the possibility of a new or 
different kind of accident from that previously evaluated for Quad 
Station. Although the SBO DGs will be used for a new function, there 
is no significant change in the operation of the SBOs for a non-fire 
event. Moreover, the overall use of the SBO DGs as an AC power 
source is not significantly different from the use of the EDGs. The 
SBO DGs buses provide power to the same buses that are powered from 
the EDGs. No new modes of operation are introduced by the proposed 
changes. The use of the SBO DGs provides a slightly different but 
effective method for achieving and maintaining post-fire safe 
shutdown for areas requiring alternate shutdown capability. As such, 
the proposed change does not create the possibility of a new or 
different kind of accident.
    (3) No significant reduction in the margin of safety is involved 
because:
    A change in the fire protection program does not result in a 
significant reduction in the margin of safety if the change does not 
result in a significant adverse impact on the plant's ability to 
achieve and maintain safe shutdown in the event of a fire. The 
proposed use of the SBO DGs instead of the EDGs to achieve and 
maintain safe shutdown within 72 hours change does not significantly 
affect the capability or reliability of the equipment assumed to 
operate in the safety analysis.
    The demonstrated capability and reliability of the SBO and EDGs 
are not significantly different. Indeed, the SBO DGs represent a 
safety improvement due to their physical separation from the 
postulated fire areas, and the operational benefits provided by 
their greater capacity. Any narrow reduction in margin associated 
with the need to manually start the SBO DGs is offset by the 
reduction in manual actions necessary to reduce electrical loads 
powered from the EDGs. The lack of Class 1E qualification for the 
SBO DGs is not significant from a safety perspective because the 
demonstrated reliability of the SBO DGs is comparable to the 
reliability of the EDGs. The lack of seismic qualification and 
single failure protection do not constitute a significant reduction 
in margin since neither of these attributes is required by Appendix 
R. Accordingly, the Commission has already determined that these 
attributes are not part of the Appendix R acceptance criterion. Any 
reduction in margin associated with the greater fuel consumption 
rate of the SBO DGs is partially offset by the increased flexibility 
in powering equipment to achieve and maintain post fire safe 
shutdown. Additionally, onsite fuel storage and manual transfer 
capabilities provide for at least 72 hours of SBO DG operation. 
Within 72 hours, deliveries of diesel fuel from offsite supplies is 
expected. Therefore, the use of the SBO DGs as an onsite AC power 
source for

[[Page 25105]]

equipment necessary to achieve and maintain post-fire safe shutdown 
in areas requiring alternate capabilities does not involve a 
significant reduction in margin.

The licensee has evaluated the use of the mobile SSD battery cart to 
provide the power source for the Automatic Depressurization System 
(ADS) valves under certain scenarios where the valves are needed to 
achieve cold shutdown and determined that it does not involve a 
significant hazards consideration for the reasons discussed below.

    (1) No significant increase in the probability or consequences 
of an accident previously evaluated is involved.
    The accident previously evaluated is the postulated fire 
requiring alternate shutdown capability. The probability of a 
previously evaluated fire is not increased significantly because the 
mobile SSD batteries do not create significant new ignition sources 
or any other fire initiators. The consequences of a previously 
evaluated fire are not increased significantly because the mobile 
SSD batteries do not significantly increase the fire loading in the 
plant, do not interfere with the plant's ability to extinguish a 
fire, and are fully capable of fulfilling the designed safety 
function.
    The associated systems related to this proposed change are not 
affected in a way that could impact the initiation of any accident 
sequence for the Quad Cities Station. No modes of operation are 
introduced by the proposed change such that adverse consequences 
result.
    The probability of an accident involving the use of the mobile 
SSD batteries would not be increased significantly by this proposed 
use because the use is not significantly different from the 
alternative manual attachment of a power source to the ADS valves.
    The consequences of an accident involving the use of the mobile 
SSD batteries are not increased because the only significant 
consequences would be a delay in achieving cold shutdown and that 
would have no different consequences than would a delay due to an 
accident related to the currently used manual power source.
    (2) The possibility of a new or different kind of accident from 
any accident previously evaluated is not created.
    The proposed change for the Quad Cities Station does not create 
the possibility of a new or different kind of accident from that 
previously evaluated. Because the mobile SSD batteries simply 
provide a different form of manually connecting a source of power to 
the ADS valves, the use of the mobile SSD batteries does not present 
new or different kinds of accidents related to such manual actions. 
Finally, because no new modes of operation are introduced by the 
proposed change, the change does not create the possibility of a new 
or different kind of accident that could be related to new modes of 
operation.
    (3) No significant reduction in the margin of safety is 
involved.
    The analytic framework for determining the extent to which a 
proposed change affects the margin of safety has been discussed 
above and, so will not be repeated here. In this case, a review of 
the proposed changes shows that they will not have an adverse impact 
on the ability to achieve and maintain safe shutdown. Several 
features associated with the use of the mobile SSD batteries show, 
as discussed above, that it provides an effective method for 
achieving and maintaining safe shutdown following a fire. In 
particular, use of the mobile SSD batteries reduces the overall 
complexity of the cold shutdown repairs required to supply power to 
the ADS valves and is familiar to plant personnel from their 
training on its use for other purposes.
    Design calculations regarding capabilities of the mobile SSD 
batteries show they will be capable in fulfilling their intended 
safety function for their design basis Appendix R scenario. 
Reliability of the mobile SSD batteries will be maintained by 
augmented quality standards. This will entail the conduct of 
appropriate maintenance and surveillance which is designed to ensure 
that the mobile batteries will function as intended. Reliability of 
this power source is further enhanced by the circumstance that there 
are two mobile SSD batteries, thus permitting one to act as a backup 
to the other.
    Under these circumstances, the margin of safety for achieving 
cold shutdown using the ADS valves is not reduced significantly, if 
at all, by the use of non-safety related mobile SSD batteries to 
power the ADS valves. Although safety-related station batteries had 
previously been used in this function, the method for attaching 
those batteries was more prone to human error than the method which 
has been developed for the mobile SSD batteries. Moreover, 
substantial steps have been taken to provide a high level of 
reliability for the mobile SSD batteries. Overall, therefore, the 
ability to achieve and maintain safe shutdown in the event of a fire 
has not been reduced by this change in the source of power to the 
ADS valves.

    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 
requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221 Hennepin 
Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and Austin, 
One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Stuart A. Richards

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: March 30, 1998.
    Description of amendment request: The proposed amendments would 
restore the Zion Custom Technical Specifications (CTS) that had been 
replaced with Improved Technical Specification by a previous amendment 
and would reinstate License Conditions that were deleted by that 
previous amendment. The proposed amendment would also modify the CTS to 
allow the use of Certified Fuel Handlers to satisfy shift staffing 
requirements and would change management titles and responsibilities to 
reflect the permanently shutdown organization.
    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:
    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    With a plant permanently shutdown and defueled the spectrum of 
accidents and events that remain credible is significantly reduced. 
As discussed below the proposed changes do not affect the 
probability or consequences of any accidents that do remain 
credible.
    The restoration of the CTS which were replaced with the ITS by 
Amendments 178/165 cannot increase the probability or consequences 
of any event or accident because the amendment was never 
implemented. The CTS have been maintained as the legally binding 
Technical Specifications in effect at Zion Station. The 
reinstatement of the five License Conditions deleted by Amendments 
178/165 is an administrative change in that the requirements 
contained in the License Conditions had been relocated elsewhere and 
are now being restored exactly as they were before the amendment was 
issued. Since the actual requirements have not changed there can be 
no change in the probability or consequences of any accident or 
event.
    The changes in management titles and responsibilities will not 
increase the probability or consequences of any accident or event 
because these changes are administrative and will not result in any 
decrease in the quality of management applied to Zion Station. The 
changes are commensurate with the significant reduction in site 
activities, site staffing, and risk to public health and safety that 
occurs when an operational nuclear power plant transitions to a 
permanently shutdown and defueled plant. Responsible individuals 
will have the authority to commit the personnel and resources 
necessary to fulfill their obligations for safe storage and handling 
of nuclear fuel. The change of position designations will have no 
effect on the frequency of occurrence of accident or event 
initiators, or on their consequences.
    The changes to allow use of Certified Fuel Handlers in lieu of 
personnel licensed in accordance with 10 CFR part 55 will not 
increase the probability or consequences of an accident or event 
because the Certified Fuel Handler Training and Retraining program 
(which will be approved by the

[[Page 25106]]

NRC) has been developed using a Systems Approach to Training as 
defined in 10 CFR 55.4. This approach provides assurance that the 
Certified Fuel Handlers have the knowledge, skills, and abilities 
that are commensurate with the tasks to be performed (i.e., the 
proper monitoring, handling, storage, and cooling of nuclear fuel). 
Therefore the frequency of occurrence of accident or event 
initiators is not increased and the consequences of the accidents or 
events are unaffected.
    The changes in shift staffing numbers and crew composition will 
not increase the probability or consequences of an accident or 
event. These staffing changes are commensurate with the quantity, 
complexity, and hazard level of the activities required for storage 
and handling of nuclear fuel. The elimination of the Shift Control 
Room Engineer does not affect any accident or event initiator or 
consequence since the previous specification would not have required 
that the position be manned with both units shut down. The 
elimination of the requirement for a Radiation Protection Person on 
shift will have no effect on the frequency of occurrence of 
accidents or events, nor on the consequences of the accident or 
event.
    The changes in verbiage to eliminate any implication that units 
are operational will not increase the probability or consequences of 
an accident or event because they are largely editorial changes and 
do not increase the frequency of occurrence of [or] event 
initiators, nor do they increase the consequences.
    Therefore this proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The changes proposed by this amendment do not involve new 
structures, systems, or components, or the use of existing 
structures, systems, or components in a new manner. Consequently no 
new failure mechanisms are introduced. The design and operation of 
structures, systems, or components is unaffected by:
    The restoration of CTS,
    The reinstatement of the five License Conditions deleted by 
Amendments 178/165,
    The changes in management titles and responsibilities,
    The changes to allow use of Certified Fuel Handlers in lieu of 
10 CFR [Part] 55 licensed personnel,
    The changes in shift staffing numbers and crew composition, or
    The changes in verbiage to eliminate any implication that units 
are operational.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    One of the License Conditions that would be reinstated by this 
amendment establishes limits that help ensure that the assumptions 
of the fuel handling accident analysis remain valid. License 
Condition 2.C.(7).b limits the weight of loads carried over fuel 
stored in the spent fuel pool to the weight of a single fuel 
assembly plus the tool for moving that assembly. This weight limit 
ensures that the number of fuel rods broken in a fuel handling 
accident does not exceed the maximum number of fuel rods assumed to 
break in the accident analysis. Consequently, this change continues 
to provide assurance that the margin of safety involving the number 
of fuel rods broken in the accident will not be reduced.
    Therefore, these 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(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.

Local Public Document Room location: Waukegan Public Library, 128 N. 
County Street, Waukegan, Illinois 60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and Austin, 
One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Stuart A. Richards

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: May 27, 1997, as supplemented by a 
letter dated April 20, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) of each unit to conform with 
NUREG-1431, Revision 1, ``Standard Technical Specifications--
Westinghouse Plants.'' The Commission had previously issued a Notice of 
Consideration of Issuance of Amendments in the Federal Register on July 
14, 1997 (62 FR 37628) covering all the proposed changes that were 
indeed within the scope of NUREG-1431. In DEC's May 27, 1997, 
submittal, there are proposed changes that are beyond the scope of 
NUREG-1431, which were thus not covered by the staff's July 14, 1997, 
notice. The following descriptions and no significant hazard analyses 
cover only those beyond-scope changes. Associated with each change are 
administrative/editorial changes such that the new or revised 
requirements would fit into the format of NUREG-1431.
    1. This proposed change affects the surveillance requirement 
currently contained in Sections 4.6.6.1 and 4.6.6.2, regarding the 
containment valve injection water system. The requirement to assure 
adequate capacity to maintain system pressure for at least 30 days 
would be deleted, the required system pressure of 16.2 pounds per 
square inch gauge (psig) would be replaced with a surge tank pressure 
of 36.4 psig, and the system would be tested at lower pressures and 
more restrictive leak rates.
    2. Section 3.9.2.1, regarding the boron dilution mitigating system, 
currently requires both trains to be operable in Mode 6 (refueling). 
DEC proposed to add a note stating that the system may be blocked 
during core reloading until two assemblies are loaded into the core. 
Adequate shutdown margin will continue to be controlled and verified by 
other specifications. This blocking would prevent inadvertent actuation 
of the system, which could distract the operating personnel, but would 
not diminish the monitoring function of the system.
    3. DEC proposed to change the definition of `dose equivalent 
iodine-131.' Subsequently, this proposed change was withdrawn by letter 
dated April 20, 1998.
    4. DEC proposed to change Section 3.3.3.6 regarding accident 
monitoring instrumentation. Specifically, the change would (a) increase 
the time allowed to return the required number of channels to operable; 
and (b) permit continued operation if one channel is inoperable given 
certain conditions are met, instead of requiring shutdown.
    5. DEC proposed to change Section 4.6.4.1 regarding surveillance 
requirements for the hydrogen monitors (combustible gas control). 
Specifically, this would eliminate the channel operational test, and 
extend the channel check frequency from once per 12 hours to once per 
31 days.
    6. DEC proposed to change Section 3.4.6.1 regarding reactor coolant 
leakage detection systems; a system comprising diverse instruments such 
as gaseous radioactivity monitoring, containment floor and equipment 
sump monitoring, etc. In addition to the instruments specified by this 
section, the plant has other installed instruments such as monitors for 
humidity, temperature, etc., which can provide indication for reactor 
coolant leakage. Currently, this specification allows operation up to 
30 days if the containment floor and equipment sump monitoring system 
is inoperable. The change would impose a requirement to perform a 
precision water balance of the reactor coolant system every 24 hours 
during this period. The change would also reduce the number of monitors 
required operable provided compensatory measures are performed or 
diverse instruments continue to be available.

[[Page 25107]]

    7. DEC proposed to change Section 4.5.4.b, which currently requires 
verification of the refueling water storage tank temperature to be 
within the allowed range once per 24 hours if the outside air 
temperature is less than 70 degrees or greater than 100 degrees 
Fahrenheit. The proposed change would simply require that the tank 
temperature be verified within range every 24 hours regardless of 
outside air temperature.
    8. DEC proposed to revise Table 3.7-1, which imposes limits on the 
maximum allowable power range neutron flux high setpoint for various 
numbers of inoperable safety valves on any operating steam generator. 
The revision would reduce the setpoints, making them more conservative.
    9. Section 3.7.6, regarding the condensate storage system, 
currently only exists in the Unit 2 TS. DEC proposed to impose these 
requirements also on Unit 1.
    10. Several electrical busses and inverters currently covered by 
Section 3.8.3.1 are qualified by a footnote, which specifies the 
conditions under which the inverter may be disconnected from its direct 
current source. DEC proposed to delete this footnote because it is not 
needed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analyses of the issue of no significant hazards 
consideration for each of the above proposed changes. The NRC staff has 
reviewed the licensee's analyses against the standards of 10 CFR 
50.92(c). The NRC staff's analysis is presented below.
    1. Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.'' 
The proposed changes will not affect the safety function of the subject 
systems. There will be no direct effect on the design or operation of 
any plant structures, systems, or components. No previously analyzed 
accidents were initiated by the functions of these systems, and the 
systems were not factors in the consequences of previously analyzed 
accidents. Therefore, the proposed changes will have no impact on the 
consequences or probabilities of any previously evaluated accidents.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.'' 
The proposed changes would not lead to any hardware or operating 
procedure change. Hence, no new equipment failure modes or accidents 
from those previously evaluated will be created.
    3. Will the change involve a significant reduction in a margin of 
safety?
    For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.'' 
Margin of safety is associated with confidence in the design and 
operation of the plant. The proposed changes to the TS do not involve 
any change to plant design, operation, or analysis. Thus, the margin of 
safety previously analyzed and evaluated is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for each of the proposed changes. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.

Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina
Attorney for licensee: Mr. Paul R. Newton, Legal Department (PB05E), 
Duke Energy Corporation, 422 South Church Street, Charlotte, North 
Carolina
NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414, 
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: April 8, 1998.
    Description of amendment request: The proposed amendments would 
revise Section 3.6.5.1 and 4.6.5.1 of the Technical Specifications (TS) 
of each unit to relax ice condenser stored ice weight requirements by 
approximately 6 percent. The proposed change is based mainly on DEC's 
gathered data showing lower sublimation rate than originally 
anticipated.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analyses of the issue of no significant hazards 
consideration for the proposed changes. The NRC staff has reviewed the 
licensee's analyses against the standards of 10 CFR 50.92(c). The NRC 
staff's analysis is presented below.
    1. Will the changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes will not affect the safety function of the 
ice condenser in that there will be no changes to the design or 
operation of any plant structures, systems, or components. No 
previously analyzed accidents were initiated by the functions of the 
ice condenser, and the ice condenser will remain fully capable of 
performing its design accident mitigation function. Therefore, the 
proposed changes will have no impact on the consequences or 
probabilities of any previously evaluated accidents.
    2. Will the changes create the possibility of a new or difference 
kind of accident from any accident previously evaluated?
    No. The proposed changes would not lead to any hardware or 
operating procedure change. Reducing the required ice weight will not 
have any impact on other plant systems that were assumed to be accident 
initiators. Hence, no new equipment failure modes or accidents from 
those previously evaluated will be created.
    3. Will the changes involve a significant reduction in a margin of 
safety? No. Margin of safety is associated with confidence in the 
design and operation of the plant; specifically, the ability of the 
fission product barriers to perform their design functions during and 
following an accident. The proposed changes regarding required ice 
weight do not involve any change to plant design, operation, or 
analysis. Thus, the margin of safety previously analyzed and evaluated 
is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for the proposed changes. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.

Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina
Attorney for licensee: Mr. Paul R. Newton, Legal Department (PB05E), 
Duke Energy Corporation, 422 South Church Street, Charlotte, North 
Carolina
NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: May 27, 1997.
    Description of amendment request: The proposed changes would lower 
the minimum required diesel generator (DG) air start receiver pressure 
from 220 per square inch gauge (psig) to 210 psig with a monthly 
verification, and would include an allowed outage time of 48 hours for 
a degraded air receiver provided the redundant air receiver is 
maintained at equal to or greater than 210 psig. These proposed changes 
are associated with DEC's application to convert to the Improved 
Technical

[[Page 25108]]

Specifications. Also, they are considered less restrictive requirements 
because of the lower required minimum pressure and the allowance of 
continued operation with a degraded starting air system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration for each change, which is presented below:

    1. (Do the changes) involve a significant increase in the 
probability or consequence of an accident previously evaluated?
    The proposed changes provide Actions for degraded capabilities 
of the diesel starting air subsystems for the DG. The proposed 
Actions establish limits for the DG starting air subsystems of 210 
psig, (are) allowed to decrease below the required value for 48 
hours(, and are verified every 31 days.) The Completion Times are 
based on the amount of capability remaining, and the time needed to 
correct any deficient condition. If the Completion Times are 
exceeded, the specification requires the associated DG to be 
declared inoperable immediately, consistent with the current TS 
(technical specifications). Since the new Actions continue to assure 
that the associated DG remains capable of performing its design 
safety function, the proposed (changes do) not significantly affect 
the probability or consequences of an accident previously evaluated.
    2. (Do the changes) create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed (changes do) not permit operation in a new or 
different mode, or permit the installation of a new or different 
type of equipment. The proposed changes provide Actions for degraded 
capabilities of the DG starting air subsystems. The proposed Actions 
establish Conditions, Required Actions, and Completion Times to be 
entered when in a degraded condition. The DG remains capable of 
performing its design safety function. Therefore, the proposed 
(changes do) not create the possibility of a new or different kind 
of accident from those previously evaluated.
    3. (Do these changes) involve a significant reduction in a 
margin of safety?
    The proposed (changes do) not significantly increase the 
probability or consequences of an accident previously evaluated. The 
changes provide assurance that timely action will be initiated to 
restore DG starting air subsystem when inoperabilities exist, 
without unnecessarily forcing plant shutdown. Based on the limit for 
the starting air subsystem for the DG, the limited time allowed is 
acceptable to restore the parameter to within the requirements 
without unnecessary plant shutdown. Therefore, (these 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: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422 
South Church Street, Charlotte, North Carolina
NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: May 27, 1997.
    Description of amendment request: The two proposed changes are 
associated with DEC's application to convert to the Improved Technical 
Specifications and are considered as administrative changes. The first 
change would delete a current requirement to only verify the refueling 
water storage tank temperature once every 24 hours if the outside air 
temperature is less than 70 degrees or greater than 100 degrees 
Fahrenheit, and would require that the tank temperature be verified 
within range every 24 hours regardless of the outside air temperature 
value. The second change would delete the current requirement that 32 
of 33 hydrogen igniters be operable on each train, and would require 
that 34 igniters per train to be operable. The actual design contains 
35 igniters per train. This change would correct an inadvertent error 
in the current Technical Specifications (TS). The number of igniters 
was increased to 35 after the first refueling outage of each unit. This 
change would correct the TS to reflect the requirements stated in 
Safety Evaluation Report Supplement 7.
    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 for each of the above proposed changes. The NRC staff has 
reviewed the licensee's analyses against the standards of 10 CFR 
50.92(c). The NRC staff's analysis is presented below:
    1. Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes will not affect the safety function of the 
subject systems. There will be no direct effect on the design or 
operation of any plant structures, systems, or components. No 
previously analyzed accidents were initiated by the functions of these 
systems, and the systems were not factors in the consequences of 
previously analyzed accidents. Therefore, the proposed changes will 
have no impact on the consequences or probabilities of any previously 
evaluated accidents.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes would not lead to any hardware or operating 
procedure change. Hence, no new equipment failure modes or accidents 
from those previously evaluated will be created.
    3. Will the change involve a significant reduction in a margin of 
safety?
    Margin of safety is associated with confidence in the design and 
operation of the plant. The proposed changes to the TS do not involve 
any change to plant design, operation, or analysis. Thus, the margin of 
safety previously analyzed and evaluated is maintained.
    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied for each of the proposed changes. 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
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422 
South Church Street, Charlotte, North Carolina
NRC Project Director: Herbert N. Berkow

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: May 27, 1997.
    Description of amendment request: The proposed change would allow 
two charging pumps or safety injection pumps capable of injecting into 
the Reactor Coolant System (RCS) when the RCS is depressurized and an 
RCS vent of at least 4.5 square inches is established. This proposed 
change is associated with the licensee's application to convert to the 
Improved Technical Specifications and results in a requirement less 
restrictive than the current requirement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 25109]]

issue of no significant hazards consideration for each change, which is 
presented below:

    1. Does the change involve a significant increase in the 
probability or consequence of an accident previously evaluated?
    The proposed change will provide an additional alternative for 
low temperature (overpressure) relief capacity when two charging 
pumps or safety injection pumps are capable of injecting into the 
RCS. The low temperature (overpressure) protection is not considered 
to be an initiator of any analyzed event, therefore, the proposed 
change does not increase the probability of a previously analyzed 
event.
    The proposed change provides an equivalent vent size to the 
existing two open PORVs (power-operated relief valves). Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    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 necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in the manner in which the plant is operated. 
The proposed change adds an additional alternative to overpressure 
protection equivalent to the current requirements. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident than any previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    As described above, the proposed change adds an additional 
alternative to overpressure protection equivalent to the current 
requirements. The inclusion of additional alternatives provides the 
operating staff with additional flexibility in meeting low 
temperature overpressure protection requirements. 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: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422 
South Church Street, Charlotte, North Carolina
NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: March 25, 1998
    Description of amendment request: Revise Technical Specification 
(TS) 3.9.8.1, ``Shutdown Coolant and Coolant Circulation High Water 
Level,'' and TS 3.9.8.2, ``Shutdown Cooling and Coolant Circulation Low 
Water Level,'' to change the minimum water level above the fuel 
assemblies seated in the reactor vessel at which the Shutdown Cooling 
(SDC) System is required to be maintained operable, or be in operation. 
In addition, TS 3.8.1.2, ``Electric Power Systems, A.C. Sources, 
Shutdown,'' and Technical Specification Bases 3/4.9.8, ``Shutdown 
Cooling and Coolant Circulation,'' have been changed to make the 
wording consistent with TS 3.9.8.1 and TS 3.9.8.2.
    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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequence of any accident?
    Response: No.
    The operation of the facility in accordance with this change 
does not involve an increase in the probability of any accident.
    Changing the water level at which the Shutdown Cooling (SDC) 
System is required to be maintained operable or be in operation will 
not increase the probability or consequences of an accident. The 
design, operation, or configuration of the SDC system will not be 
changed.
    At least one shutdown cooling train will be in operation to 
ensure sufficient cooling capacity is available to remove decay heat 
and maintain the water in the reactor pressure vessel below 140 
degree F as required during the refueling mode.
    At least one shutdown cooling train will be in operation to 
ensure sufficient coolant circulation is maintained through the 
reactor core to minimize the effects of a boron dilution incident 
and prevent boron stratification. Technical Specification 3.9.10.1, 
``Refueling Operations Water Level--Reactor Vessel Fuel 
Assemblies,'' will be complied with, and therefore, the assumptions 
related to iodine removal and the fuel handling accident will be 
preserved.
    Sufficient time, approximately 1.00 hours, will be available to 
the operators to initiate compensatory measures to preclude the 
initiation of core boiling in the unlikely event SDC should be loss 
[lost].
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: No.
    The operation of the facility in accordance with this proposed 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed change will not affect the design, configuration, 
or operation of the SDC system, and therefore there are no new modes 
of failure introduced.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    Operation of the facility in accordance with this proposed 
change will not involve a significant reduction in a margin of 
safety.
    The calculation of the time to the initiation of boiling based 
on 23 feet above the top of the fuel seated in the reactor vessel, 
at four days after shutdown, demonstrates there is significant time 
available, approximately 1.00 hour, to the operators within which to 
take compensatory measures to preclude the initiation of boiling. 
The calculation shows that based on 23 feet of water above the 
reactor flange there is 2.04 hours to the initiation of boiling. 
Although there is a reduction in the time to the initiation of 
boiling, compensatory measures could be taken within a few minutes 
to restore SDC, and thus, there is still a significant margin 
available to the operators within which to preclude the initiation 
of boiling. Thus, the margin of safety is not significantly reduced.
    The time to core uncovery was determined to be 27.74 hours based 
on four days after shutdown and water level twenty-three (23) feet 
above the fuel assemblies seated in the reactor vessel.

    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: University of New Orleans Library, 
Louisiana Collection, Lakefront, New Orleans, LA 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 L 
Street N.W., Washington DC 20005-3502
NRC Project Director: John N. Hannon

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

    Date of amendment request: March 20, 1998.
    Description of amendment request: The proposed amendment requests 
editorial changes to the Improved Technical Specifications (ITS) Safety 
Limits and Administrative Controls to replace the titles of the Senior 
Vice President, Nuclear Operations (SVPNO) and the Vice President, 
Nuclear Production (VPNP) with the position of Chief Nuclear Officer 
(CNO). The CNO combines the duties of the SVPNO and VPNP as currently 
described in ITS and is required to be an officer of the company. The 
proposed change is

[[Page 25110]]

intended to allow upgrading the position of the corporate officer 
responsible for overall nuclear operations without limiting the title.
    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.

    Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the deletion and updating of individual titles 
does not affect plant operation. No design basis accidents are 
affected by the proposed administrative and editorial changes and, 
as such, there are no physical changes to the facility or its 
operation.
    Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed ITS changes are administrative and editorial in 
nature. No changes to the facility structures, systems and 
components or their operation will result. The design and design 
basis of the facility remain unchanged. The plant safety analyses 
remain current and accurate. No new or different failure mechanisms 
are introduced. Therefore, the possibility of a new or different 
kind of accident from any accident previously evaluated is not 
introduced.
    Does not involve a significant reduction in the margin of 
safety.
    The proposed ITS changes are administrative and editorial in 
nature. The proposed safety margins established through the design 
and facility license including the Improved Technical Specifications 
remain unchanged. In addition, the proposed amendment ensures 
continued emphasis and assignment of responsibility for overall 
nuclear safety. Therefore, all margins of safety are maintained.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of Sec. 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: Coastal Region Library, 8619 W. 
Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida 
33733-4042
NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: March 20, 1998.
    Description of amendment request: The proposed amendment would 
change the Inservice Inspection Program described in Improved Technical 
Specification (ITS) 5.6.2.8.c. This ITS currently states that the 
reactor coolant pump (RCP) motor flywheels will be inspected during the 
``Spring 1998 refueling outage,'' which would have been refueling 
outage 11. Due to a recent 17-month extended outage, refueling outage 
11 has been deferred until Fall 1999. The proposed change is intended 
to accurately reflect the new refueling outage 11 schedule.
    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.

    The proposed change will not significantly increase the 
probability or consequences of an accident previously evaluated.
    The safety function of the RCP flywheels is to provide a 
coastdown period during which the RCPs would continue to provide 
reactor coolant flow to the reactor after loss of power to the RCPs. 
The maximum loading on the RCP motor flywheel results from overspeed 
following a large loss of coolant accident (LOCA). The estimated 
maximum obtainable speed in the event of a Reactor Coolant System 
piping break was established conservatively. The proposed one-time 
editorial change to remove the words ``Spring 1998 refueling 
outage'' and replace them with ``to coincide with Refueling Outage 
11R'' does not affect that analysis. The proposed change in dates is 
editorial in that it merely reflects the new date for cycle 11. The 
usage time for the flywheels is bounded by the original estimates. 
The proposed editorial change does not affect the amount of 
radioactive material available for release or modify any systems 
used for mitigation of such releases during accident conditions. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed editorial change will not change the design, 
configuration, or method of operation of the plant. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed change will not involve a significant reduction to 
any margin of safety.
    The proposed Amendment is an editorial change to reflect that 
CR-3's operating cycle is not ending in spring 1998, but in fall 
1999. The proposed change does not affect the methods of inspection 
or its acceptance criteria. Therefore, the margins of safety defined 
in RG [Regulatory Guide] 1.14 are not changed.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of Sec. 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: Coastal Region Library, 8619 W. 
Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida 
33733-4042
NRC Project Director: Frederick J. Hebdon

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: April 15, 1998.
    Description of amendment request: The proposed amendment would 
update the existing pressure-temperature curves with new curves with 
values from 18 to 32 effective full power years based on the testing 
and analysis of reactor pressure vessel surveillance materials.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The pressure-temperature limits are not 
derived from Design Basis Accident (DBA) analyses. They are 
prescribed by the ASME B&PV Code and 10 CFR part 50 appendices G and 
H as restrictions on normal operation to avoid encountering 
pressure, temperature, and temperature rate of change conditions 
that might cause undetected flaws to propagate and cause nonductile 
failure of the reactor coolant pressure boundary.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The amendment will merely update the pressure-temperature 
curves (and associated SRs and Bases) already existing in the plant 
Improved Technical Specifications to provide limits from 18 to 32 
EFPY of operation, which are based upon evaluation and analysis of 
actual in-vessel material specimens, per 10 CFR part

[[Page 25111]]

50, appendices G and H. The pressure-temperature curves are 
established to the requirements of 10 CFR part 50, appendix G to 
assure that brittle fracture of the reactor vessel is prevented.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety. 10 CFR part 50, appendix G 
specifies fracture toughness requirements to provide adequate 
margins of safety during operation over the service lifetime. The 
values of adjusted reference temperature and upper shelf energy 
determined as a result of the 10 CFR part 50, appendices G and H 
analysis are expected to remain within the limits of Regulatory 
Guide 1.99, Revision 2 and appendix G of 10 CFR part 50 (less than 
200 deg. F and greater than 50 ft-lbs respectively) for at least 32 
EFPY of operation.

    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: Cedar Rapids Public Library, 500 
First Street, SE., Cedar Rapids, IA 52401
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869
Acting NRC Project Director: Richard P. Savio

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 27, 1997.
    Description of amendment request: The proposed amendment, included 
as part of the proposed conversion from the current Technical 
Specifications (TS) to improved TS, would establish Allowable Values 
for the instrumentation included in Section 3.3, as a result of the 
plant-specific application of the General Electric Instrument Setpoint 
Methodology to the Cooper Nuclear Station (CNS).
    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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change in selected Allowable Values for the 
instrumentation included in proposed Section 3.3 of the Technical 
Specifications is the result of application of the CNS 
instrumentation setpoint methodology. This methodology incorporates 
the guidance of ISA Recommended Practice ISA-RP67.04, Part II, 
``Methodologies for the Determination of Setpoints for Nuclear 
Safety-Related Instrumentation,'' September 1994. Application of 
this methodology results in instrumentation selected Allowable 
Values which more accurately reflect total instrumentation loop 
accuracy as well as that of test equipment and setpoint drift 
between Surveillances. The proposed change will not result in any 
hardware changes. The instrumentation included in proposed Section 
3.3 of the Technical Specifications is not assumed to be an 
initiator of any analyzed event. Existing operating margin between 
plant conditions and actual plant setpoints is not significantly 
reduced due to this change. As a result, the proposed change will 
not result in unnecessary plant transients.
    The role of the proposed Section 3.3 instrumentation is in 
mitigating and thereby limiting the consequences of accidents. The 
Allowable Values have been developed to ensure that the design and 
safety analysis limits will be satisfied. The methodology used for 
the development of the Allowable Values ensures the affected 
instrumentation remains capable of mitigating design basis events as 
described in the safety analyses and that the results and 
consequences described in the safety analyses remain bounding. 
Additionally, the proposed change does not alter the plant's ability 
to detect and mitigate events. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change is the result of application of the CNS 
instrumentation setpoint methodology and do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. This is based on the fact that the method and 
manner of plant operation is unchanged. The use of the proposed 
Allowable Values does not impact safe operation of CNS in that the 
safety analysis limits will be maintained. The proposed Allowable 
Values involve no system additions or physical modifications to 
systems in the station.
    These Allowable Values were developed using a methodology to 
ensure the affected instrumentation remains capable of mitigating 
accidents and transients. Plant equipment will not be operated in a 
manner different from previous operation, except that setpoints may 
be changed. Since operational methods remain unchanged and the 
operating parameters have been evaluated to maintain the station 
within existing design basis criteria, no different type of failure 
or accident is created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change does not involve a reduction in a margin of 
safety. The proposed changes have been developed using a methodology 
to ensure safety analysis limits are not exceeded. As such, this 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 27, 1997.
    Description of amendment request: The proposed amendment, included 
as part of the proposed conversion from the current Technical 
Specifications (CTS) to the improved Technical Specifications (ITS), 
would add an additional action statement to a limiting condition for 
operation (LCO). The LCO is in the Improved Standard Technical 
Specifications (ISTS, NUREG-1433, Revision 1) 3.6.2.3 on the residual 
heat removal suppression pool cooling subsystems. The requirements in 
the proposed ITS 3.6.2.3 on the subsystems do not exist in the CTS. The 
Action B for ITS 3.6.2.3 would require that if the two such subsystems 
were inoperable, one subsystem would have to be restored to operability 
within 8 hours or the plant would be in ITS 3.0.3. ITS 3.0.3 governs 
plant operation if an LCO (i.e., ISTS 3.6.2.3) and the associated 
action statement are not met (i.e., Action B to ISTS 3.6.2.3).
    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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change provides more stringent requirements for 
operation of the facility. These more stringent requirements do not 
result in operation that will increase the probability of initiating 
an analyzed event and do not alter assumptions relative to (the) 
mitigation of an accident or transient event. The more restrictive 
requirements continue to ensure * * * systems, and components 
((i.e., the residual heat removal suppression pool cooling 
subsystems)) are maintained consistent with the safety analyses and 
licensing basis. Therefore, this (the proposed)

[[Page 25112]]

change does not involve a significant (an) increase in the 
probability or consequences of any accident previously evaluated.
    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 a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in the methods governing normal plant operation. The 
proposed change does impose different requirements. However, this 
change is consistent with the assumptions in the safety analyses and 
licensing basis. Thus, this change does not create 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 imposition of more restrictive requirements either has no 
impact on or increases the margin of plant safety. As provided in 
the discussion of the change, each change in this category (i.e., 
more restrictive requirements) is, by definition, providing 
additional restrictions to enhance plant safety. The change 
maintains requirements (systems and components) within the safety 
analyses and licensing basis. Therefore, this 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: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 27, 1997.
    Description of amendment request: The proposed amendment, included 
as part of the proposed conversion from the current Technical 
Specifications (CTS) to the improved Technical Specifications (ITS), 
would add an additional test (i.e., water and sediment content within 
limits) of diesel fuel oil that could be used in place of a current 
test (i.e., clear and bright appearance with proper color) in the 
diesel fuel oil testing program. The current tests are listed in CTS 
4.9.A.2.d/e. The testing program will be in the new ITS 5.5.9. The 
additional test is change number 25 to Section 5.0 of the Improved 
Standard Technical Specifications (NUREG-1433, Revision 1).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change provides more stringent requirements for 
operation of the facility. (This) more stringent (requirement) 
do(es) not result in operation that will increase the probability of 
initiating an analyzed event and do(es) not alter assumptions 
relative to (the) mitigation of an accident or transient event. The 
more restrictive (requirement) continue(s) to ensure * * * systems 
and components (i.e., the diesel generators) are maintained 
consistent with the safety analyses and licensing basis. Therefore, 
the proposed change does not involve an increase in the probability 
or consequences of any accident previously evaluated.
    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 a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in the methods governing normal plant operation. However, 
this change is consistent with the assumptions in the safety 
analyses and licensing basis. Thus, this change does not create 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 imposition of more restrictive requirements either has no 
impact on or increases the margin of plant safety. As provided in 
the discussion of the change, each change in this category (i.e., a 
more restrictive requirement) is, by definition, providing 
additional restrictions to enhance plant safety. The change 
maintains (systems and components) within the safety analyses and 
licensing basis. Therefore, this 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: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 27, 1997.
    Description of amendment request: The proposed amendment, included 
as part of the proposed conversion from the current Technical 
Specifications (TS) to improved TS for the Cooper Nuclear Station 
(CNS), would relocate the Trip Level Settings for the Rod Block Monitor 
from Table 3.2.C of the current TS to the Core Operating Limits Report. 
Also, details relating to the Alternate Shutdown system design and 
operation are proposed to be relocated from current TS 3.2.I and 4.2.I 
to the improved TS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the three criteria of 10 CFR 50.92(c), and has determined the 
following:
    The proposed changes relocate certain details from the Technical 
Specifications to the Bases and the Core Operating Limits Report 
(COLR). The Bases and the COLR containing the relocated information 
will be maintained in accordance with 10 CFR 50.59. In addition, the 
Bases and COLR are subject to the applicable change control provisions 
of Chapter 5.0, Administrative Controls'', of the proposed improved 
Technical Specifications. Since any changes to the Bases or the COLR 
will be evaluated per the requirements of 10 CFR 50.59 or other 
applicable change control provisions, no increase in the probability or 
consequences of an accident previously evaluated will result. 
Therefore, these changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not involve any physical alterations to the 
plant (no new or different type of equipment will be installed), or 
changes in the methods governing normal plant operation. The proposed 
changes will not impose or eliminate any requirements, and adequate 
control of the information will be maintained. Thus, these changes do 
not create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed changes will not reduce a margin of safety because 
they have no impact on any safety analysis assumptions. In addition, 
the details to be transposed from the TS to the Bases

[[Page 25113]]

and the COLR are unchanged. Since any future changes to these details 
in the Bases or the COLR will be evaluated per the requirements of 10 
CFR 50.59 or other applicable change control provisions, no reduction 
in a margin of safety will result. As such, these proposed changes do 
not involve a significant reduction in a margin of safety.
    Based on the above discussion, 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: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499
NRC Project Director: John N. Hannon

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: April 8, 1998.
    Description of amendment request: The proposed change would revise 
Technical Specifications (TSs) 4.4.5.3, Steam Generators--Inspection 
Frequencies, and 3.4.6.2.c, Reactor Coolant System (RCS) Leakage, and 
the associated bases to accommodate fuel cycles of up to 24 months with 
respect to the allowed time interval between steam generator inservice 
inspections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Extending Surveillance Requirement (SR) 4.4.5.3 to accommodate a 
24 month cycle for inspection of steam generator tubes structural 
integrity, as well as, imposing a more restrictive Limiting 
Condition for Operation (TS 3.4.6.2.c) for reactor coolant system 
leakage through Category C-2 steam generators, will neither 
exacerbate nor significantly increase the probability or 
consequences of an accident previously evaluated in the Seabrook 
Station [updated final safety analysis report] UFSAR.
    The proposed changes to SR 4.4.5.3 do not alter the intent or 
method by which the surveillances are conducted, do not involve 
physical changes to the plant, do not alter the way structures, 
systems or components (SSCs) function, and do not modify the manner 
in which the plant is operated.
    The proposed change to TS 3.4.6.2.c imposes more restrictive 
limits on plant operations due to RCS leakage through steam 
generators. The proposed change does not involve physical changes to 
the plant or alter the way a SSC functions.
    The proposed changes to SR 4.4.5.3 and TS 3.4.6.2.c, and their 
associated Bases, will not adversely affect the ability of the steam 
generators to perform their intended safety function. Furthermore, 
the proposed changes do not adversely affect the physical protective 
boundaries of the plant. The proposed changes do not affect accident 
initiators or precursors and do not alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. The proposed changes do not alter or prevent the 
ability of SSCs to perform their intended function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the Updated Final Safety Analysis Report (UFSAR). The 
proposed changes are administrative in nature and do not change the 
level of programmatic controls or the procedural details associated 
with aforementioned surveillance requirements. While the proposed 
changes will lengthen the interval between surveillances, the 
increase in interval has been evaluated; and based on the reviews of 
the steam generator tube eddy current test (ECT) inspections, it is 
concluded that the wear growth rate of the only active degradation 
mechanism (Anti-Vibration Bar (AVB) wear) identified to date at 
Seabrook Station is such that sufficient margin exists between the 
plugging criteria and structural limit such that no tubes are 
predicted to exceed the structural limit even with the longer 
surveillance interval.
    Since there are no changes to previous accident analyses, the 
radiological consequences associated with these analyses remain 
unchanged, therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. Therefore, the proposed changes will 
not significantly increase the probability or consequences of any 
previously analyzed accident.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed changes to TS 3.4.6.2 and SR 4.4.5.3, and 
associated Bases, do not alter the design assumptions, conditions, 
configuration of the facility or the manner in which the plant is 
operated. There are no changes to the source term, containment 
isolation or radiological release assumptions used in evaluating the 
radiological consequences in the Seabrook Station UFSAR. Existing 
system and component redundancy is not being changed by the proposed 
changes. The proposed changes have no impact on component or system 
interactions. The proposed changes are administrative in nature and 
do not change the level of programmatic controls and procedural 
details associated with the aforementioned surveillance 
requirements. Therefore, since there are no changes to the design 
assumptions, conditions, configuration of the facility, or the 
manner in which the plant is operated and surveilled, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed change ( ) to the surveillance intervals for SR 
4.4.5.3 is still consistent with the basis for the interval. The 
intent or method of performing the surveillances remains unchanged. 
The more restrictive limit for leakage through any one steam 
generator placed in Category C-2, as well as, the requirement to do 
an engineering assessment of steam generator tube integrity, 
provides additional margin of ensuring safe plant operation.
    In addition, there is no adverse affect on equipment design or 
operation and there are no changes being made to the Technical 
Specification required safety limits or safety system settings that 
would adversely affect plant safety. The proposed changes are 
administrative in nature and do not change the level of programmatic 
controls and procedural details associated with the aforementioned 
surveillance requirements. While the proposed changes will lengthen 
the interval between surveillances, the increase in interval has 
been evaluated; and based on the reviews of the steam generator tube 
ECT inspections, it is concluded that the wear growth rate of the 
only active degradation mechanism (AVB wear) identified to date at 
Seabrook Station is such that sufficient margin exists between the 
plugging criteria and structural limit such that no tubes are 
predicted to exceed the structural limit even with the longer 
surveillance interval. Therefore, extension of the current 
surveillance intervals to accommodate a 24 month cycle will not 
significantly degrade the ability, the availability or the 
reliability of the steam generators to perform their intended safety 
function, thus, it is concluded that there is no 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Local Public Document Room location: Exeter Public Library, Founders 
Park, Exeter, NH 03833
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear Counsel, 
Northeast Utilities Service Company, PO Box 270, Hartford, CT 06141-
0270
NRC Project Director: Cecil O. Thomas

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

    Date of amendment request: April 6, 1998.
    Description of amendment request: The proposed amendment will 
modify

[[Page 25114]]

the Technical Specifications (TSs) by (1) adding a surveillance 
requirement to verify pressurizer heater capacity to TS 3.4.4, 
``Reactor Coolant System--Pressurizer,'' (2) moving the identification 
of the location of the containment air temperature detectors from the 
surveillance requirements portion of TS 3.6.1.5, ``Containment 
Systems--Air Temperature,'' to the TS Bases for Containment Systems, 
Section 3/4.4.6.1.5, ``Air Temperature,'' and (3) modifying the action 
statements and surveillance requirements of TS 3.7.1.5, ``Plant 
Systems--Main Steam Isolation Valves.'' The TS Bases would also be 
updated to include the list of containment air temperature detectors 
and reflect the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to add a surveillance requirement (SR) 
4.4.4.2 to verify pressurizer heater capacity will help ensure the 
pressurizer will be able to function as designed to maintain Reactor 
Coolant System pressure. There will be no effect on any design basis 
accident previously evaluated or on any equipment important to 
safety. Therefore, the proposed change will not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes to modify the wording of SR 4.6.1.5 and to 
relocate the list of containment air temperature detectors from SR 
4.6.1.5 to the Bases will not affect the Technical Specification 
limit for containment temperature or the frequency of verification 
of this limit. The proposed changes do not alter the way any 
structure, system, or component functions. The initial assumption 
for containment temperature used in the design basis accident 
analysis will remain the same. There will be no affect on any design 
basis accident previously evaluated or on any equipment important to 
safety. Therefore, the proposed changes will not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes to the action statements and surveillance 
requirements of Technical Specification 3.7.1.5 will not affect the 
operability requirements of the main (steamline) isolation valves 
(MSIVs). There will be no effect on any design basis accident 
previously evaluated or on any equipment important to safety. 
Therefore, the proposed changes will not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes have no adverse effect on any of the design 
basis accidents previously evaluated or on any equipment important 
to safety. Therefore, the License Amendment Request does not impact 
the probability of an accident previously evaluated nor does it 
involve a significant increase in the consequences or an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will 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 will add SR 4.4.4.2 to verify pressurizer 
heater capacity, relocate the list of containment temperature 
detectors used to verify containment temperature from SR 4.6.1.5 to 
the associated Bases, and modify the action statements and 
surveillance requirements of Technical Specification 3.7.1.5.
    These changes will have no adverse effect on equipment important 
to safety. This equipment will continue to function as assumed in 
the design basis accident analysis. Therefore, there will be no 
significant reduction in the margin of safety as defined in the 
Bases for the technical Specifications affected by these proposed 
changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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

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

    Date of amendment request: April 13, 1998
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) by adding a new TS 3.5.5, 
``Emergency Core Cooling Systems--Trisodium Phosphate (TSP).'' Also, 
the surveillance requirements in TSs 4.5.2.c.3 and 4.5.2.c.4 would be 
relocated to new TS 3.5.5 as TS 4.5.5.1 and TS 4.5.5.2, respectively. 
The applicable TS Index page and Bases sections will be updated to 
reflect the proposed changes.
    Changes to the current requirements for the TSP are also proposed. 
The TSP requirements in TS 4.5.2.c.3 would become the limiting 
conditions for operation in the new TS; the amount of TSP required 
would increase from ``equal to or greater than 110 cubic feet'' to 
``equal to or greater than 282 cubic feet'' based on the new 
calculations; the applicability would be expanded to include all of 
Mode 3; the action statement would allow 48 hours to restore the TSP 
volume; and changes would also be made to the required tests and 
specific details would be relocated to the applicable TS Bases.
    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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to relocate the current trisodium phosphate 
(TSP) dodecahydrate Technical Specification requirements from the 
surveillance requirements for the Emergency Core Cooling System to a 
new TSP Technical Specification will not change the requirement to 
store TSP inside containment. The proposed changes will require a 
large quantity of TSP to be stored inside containment. This large 
quantity, based on a recently revised calculation, will ensure 
sufficient TSP is available for containment sump water pH control. 
These proposed changes do not alter the way any structure, system, 
or component functions. There will be no adverse effect on any 
design basis accident previously evaluated, on any equipment 
important to safety, or o n the radiological consequences of any 
design basis accident. Therefore, this License Amendment Request 
does not impact the probability of an accident previously evaluated 
nor does it involve a significant increase in the consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change to increase the TSP volume stored inside 
containment will require two of the wire mesh TSP baskets inside 
containment to be replaced by two new and larger wire mesh baskets. 
The design of the new baskets has been evaluated and it is 
consistent with the requirements for equipment installed in 
containment. The replacement of the two wire mesh baskets

[[Page 25115]]

will not result in any significant change in plant configuration and 
will not require any new or unusual operator actions. It will alter 
the way any structure, system, or component functions and does not 
alter the manner in which the plant is operated. It will not 
introduce any new failure modes. Therefore, the proposed changes 
will 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 will relocate the current Technical 
Specification requirements for TSP to a new Technical Specification. 
The minimum required volume will be increased to reflect the results 
of a new calculation performed to support the current requirement to 
raise containment sump pH [equal to or greater than] 7.0. These 
changes will have no adverse effect on equipment important to 
safety. This equipment will continue to function as assumed in the 
design basis accident analysis. Therefore, there will be no 
significant reduction of the margin of safety as defined in the 
Bases for the Technical Specifications affected by these proposed 
changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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

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

    Date of amendment request: April 11, 1997 (supersedes July 26, 
1996, application)
    Description of amendment request: The proposed amendment would 
modify the Monticello Technical Specifications (TS) sections 3.6.C, 
Coolant Chemistry, and 3/4.17.B, Control Room Emergency Filtration 
System. The changes were proposed to establish TS requirements 
consistent with modified analysis inputs used for the evaluation of the 
radiological consequences of the main steam line break accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    A limit is established in the plant Technical Specifications for 
steady state radioiodine concentration in the reactor coolant to 
ensure that in the event of a release of radioactive material to the 
environment due to a postulated high energy line break up to and 
including a design basis Main Steam Line Break Accident, radiation 
doses are maintained within the guidelines of 10 CFR part 100. The 
steady state radioiodine concentration in the reactor coolant is an 
input for analysis of the radiological consequences of an accident 
due to a Main Steam Line Break outside of containment and postulated 
high energy line breaks. In addition, requirements are established 
in the Technical Specifications for control room habitability. 
During an accident, the control room emergency filtration system 
provides filtered air to pressurize the Control Room to minimize the 
activity, and therefore the radiological dose, inside the control 
room.
    A change is proposed for the steady state radioiodine 
concentration. This value is conservative with respect to the value 
used in the Main Steam Line Break dose consequences analysis and is 
consistent with the dose consequences evaluation of a postulated 
Reactor Water Cleanup (RWCU) line break. Changes are proposed to the 
limiting conditions for operation and surveillance requirements for 
the Control Room Emergency Filtration Train iodine removal 
efficiency. These changes are consistent with the inputs used in the 
analysis of the radiological consequences of the postulated RWCU 
line break and the Main Steam Line Break Accident. These proposed 
requirements maintain operating restrictions for analytical inputs 
used in the analysis of the Main Steam Line Break Accident. 
Evaluation of these events has demonstrated that the postulated 
radiological consequences will remain within the licensing basis 
established in the AEC [Atomic Energy Commission] Provisional 
Operating License Safety Evaluation Report, dated March 18, 1970, 
thus the proposed changes do not result in an increase in the 
consequences of previously evaluated accidents.
    The analysis of the Main Steam Line Break Accident performed 
using a reactor coolant radioiodine concentration of 2 
(microcuries)/gm dose equivalent Iodine-131 and a control room 
ventilation filter efficiency consistent with the proposed Technical 
Specifications changes demonstrated that radiological consequences 
of the Main Steam Line Break are not changed significantly. The 
radiological consequences of the Main Steam Line Break Accident 
remain within the exposure guidelines of 10 CFR part 100 and 10 CFR 
part 50 appendix A, General Design Criterion 19. The offsite dose 
consequences remain bounded by the licensing basis provided in the 
AEC Provisional Operating License Safety Evaluation Report, dated 
March 18, 1970. The control room doses calculated for the hot 
standby Main Steam Line Break Accident using the TID-14844 dose 
conversion factors remain bounded by the dose consequences of the 
comparable design basis loss of coolant accident.
    The evaluation of the postulated RWCU line break, performed 
using a reactor coolant radioiodine concentration of 0.25 
(microcurie)/gm dose equivalent Iodine-131 and a control room 
ventilation filter efficiency consistent with the proposed Technical 
Specifications changes, demonstrated that the radiological 
consequences of this event remain within the exposure guidelines of 
10 CFR part 100 and 10 CFR part 50 Appendix A, General Design 
Criterion 19. The offsite dose consequences remain bounded by the 
Main Steam Line Break as established in the licensing basis provided 
in the AEC Provisional Operating License Safety Evaluation Report, 
dated March 18, 1970.
    The proposed Technical Specification changes do not introduce 
new equipment operating modes, nor do the proposed changes alter 
existing system inter-relationships. The proposed changes do not 
introduce new failure modes. The system improvements to reduce 
bypass leakage during postulated accidents do not have an adverse 
effect on control room habitability. Therefore, this amendment will 
not cause a significant increase in the probability of an accident 
previously evaluated for the Monticello plant.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed Technical Specification changes do not introduce 
new equipment operating modes, nor do the proposed changes alter 
existing system inter-relationships. Operator action to mitigate the 
consequences of the postulated RWCU line break is conservative based 
on the very limited action required by the operator to close the 
containment isolation valves and the availability of control room 
indications to alert the operator to the postulated break. The use 
of a ten (10) minute operator response time to take manual actions 
in response to postulated events is consistent with Monticello's 
licensing basis for similar events. The use of operator actions and 
all available equipment is consistent with current regulatory 
guidance for mitigating the consequences of postulated line breaks.
    The proposed change to the specification for reactor coolant 
dose equivalent radioiodine is conservative with respect to the re-
evaluation of the Main Steam Line Break Accident for the more 
conservative hot standby initial condition for the postulated 
accident. The proposed change to the specification for reactor 
coolant dose equivalent radioiodine is consistent with the 
postulated high energy line break of a Reactor Water Cleanup line. 
The proposed changes to the limiting conditions for operation and

[[Page 25116]]

surveillance requirements for the control room emergency filtration 
train iodine removal efficiency are consistent with the inputs used 
in the evaluation of the radiological consequences of the postulated 
RWCU line break and the Main Steam Line Break Accident. The system 
improvements to reduce bypass leakage during postulated accidents do 
not have an adverse effect on control room habitability. Therefore, 
the proposed amendment will not create the possibility of a new or 
different kind of accident.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Surveillance data has demonstrated the proposed requirements are 
within the current capability of the facility. The proposed changes 
maintain margins of safety. These proposed requirements maintain 
operating restrictions for analytical inputs used in the analysis of 
the bounding postulated high energy line break of a Reactor Water 
Cleanup line and the Main Steam Line Break Accident. The proposed 
change to the specification for reactor coolant dose equivalent 
radioiodine is conservative with respect to the re-evaluation of the 
Main Steam Line Break Accident for the more conservative hot standby 
initial condition for the postulated accident. The proposed change 
to the specification for reactor coolant dose equivalent radioiodine 
is consistent with the postulated high energy line break of a 
Reactor Water Cleanup line. The evaluation of these postulated 
events determined that the radiological consequences remain within 
the exposure guidelines of 10 CFR part 100 and of 10 CFR part 50 
Appendix A, General Design Criterion 19. The proposed changes to the 
limiting conditions for operation and surveillance requirements for 
the control room emergency filtration train iodine removal 
efficiency provide assurance that the system will perform at the 
filter efficiency as used in the evaluation of the radiological 
consequences of the postulated events. Therefore, the proposed 
amendment will not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Cynthia A. Carpenter

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 request: April 10, 1998.
    Description of amendment request: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 to revise TS 6.2.2.g and 6.3 to change 
the name of the Operations Manager to Operations Director and to change 
the requirement for the Operations Director to hold a senior reactor 
operator (SRO) license.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to revise the title of the Operations 
Manager to Operations Director is an administrative change that 
clarifies the Technical Specification (TS) to reflect current 
position titles.
    The proposed change provides assurance that the Operations 
Director will continue to have knowledge of pressurized water 
reactor (PWR) operation and emergency event mitigation. The proposed 
change does not detract from the Operations Director's ability to 
perform his primary responsibilities. In this case, by having 
previously held a senior reactor operator (SRO) license, the 
Operations Director has achieved the necessary training, skills, and 
experience to fully understand the operation of plant equipment and 
the watch requirements for operators. In summary, the proposed 
change does not affect the ability of the Operations Director to 
provide the plant oversight required of his position.
    Additionally, another off-shift individual that holds an SRO 
license for Diablo Canyon Power Plant (DCPP) directs the licensed 
activities of licensed operators (an Operations middle manager) will 
have specific knowledge of operation and emergency event mitigation 
at DCPP. This will assure that the change in qualification of the 
Operations Director does not affect the probability of an operator 
initiating an accident or increasing the consequences of an accident 
due to improper direction from management. The training and 
qualification programs for operators on shift will not be affected 
by the proposed changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to revise the title of the Operations 
Manager to Operations Director is an administrative change that 
clarifies the TS to reflect current position titles.
    The proposed change to TS 6.2.2g. and 6.3 do not affect the 
design or function of any plant system, structure, or component, nor 
does it change the way plant systems are operated. It does not 
affect the performance of NRC licensed operators since the proposed 
changes do not impact the training or qualification of any operator 
on shift. Operation of the plant in conformance with TS and other 
license requirements will continue to be supervised by personnel who 
hold an SRO license. The proposed change to TS 6.2.2g and 6.3 
ensures that the Operations Director will be a knowledgeable and 
qualified individual by requiring the individual to have held an SRO 
license at a PWR.
    Therefore, 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 change to revise the title of the Operations 
Manager to Operations Director is an administrative change that 
clarifies the TS to reflect current position titles.
    The proposed change involves an administrative control that is 
not related to the margin of safety. The proposed change does not 
reduce the level of knowledge or experience required of an 
individual who fills the Operations Director position, nor does it 
affect the conservative manner in which the plant is operated. The 
on-shift licensed operators will continue to be supervised by 
personnel who hold an SRO license in accordance with 10 CFR 
50.54(l).
    Therefore, neither of the proposed changes 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 Sec. 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 & 
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman

[[Page 25117]]

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: March 26, 1998.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3/4.8.2.1, ``AC Distribution--
Operating,'' to add operability conditions and action statements for 
the 115-volt vital instrument bus (VIB) D and inverter. The proposed 
amendments complete the recommended action from NRC Generic Letter 91-
11, Resolution of Generic Issues 48, ``LCOs for Class 1E Vital 
Instrument Buses,'' and 49, ``Interlocks and LCOs for Class 1E Tie 
Breakers'' pursuant to 10 CFR 50.54(f), dated July 18, 1991.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change, as described above, does not make any 
physical changes to the plant or components, nor changes the manner 
in which the plant or components are operated as a result of the 
addition of the Note and the D VIB and Inverter to the TS. The 
proposed change incorporates the operating requirements of the 
Technical Specification Interpretation (TSI) developed in response 
to GL 91-11 into the Salem Unit 1 and 2 Technical Specifications. 
Incorporating this interpretation into the Technical Specifications 
eliminates the need for the TSI.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not introduce any design or physical 
configuration change to the plants, change the function of the 115 
Volt D VIBs and inverters, or the manner in which they are 
maintained or tested.
    Therefore, the proposed amendment will 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.
    The proposed Action Times associated with the incorporation of 
the D VIB into the Technical Specifications are consistent with the 
current Action Times for the A, B, and C VIBs for a loss of an AC 
bus. Adding the note to the Salem Unit 1 Technical Specification 
brings consistency between Salem Units 1 and 2, and is also 
consistent with NUREG 1431, Vol. 1, Rev 1 ``Standard Technical 
Specifications Westinghouse Plants.''
    The outage duration limit of 72 hours for the D inverter is 
acceptable based on the following: (1) the proposed 72 hours Action 
Time to restore the inoperable inverter to operable is supported by 
a PSA [probabilistic safety assessment] assessment. NRC Draft SRP 
[Standard Review Plan] Chapter 16.1, Revision 13, ``Risk-Informed 
Decision making: Technical Specifications'' notes that an 
incremental conditional core damage probability (ICCDP) of 5.0 E-7 
is considered very small. The proposed 72 hour allowable outage time 
was calculated utilizing the NRC incremental conditional core damage 
probability (ICCDP), and (2) the inoperability of the D VIB Inverter 
will not affect the operation of any Safeguard Equipment Cabinet 
(SEC) or Emergency Diesel Generator (EDG).
    Therefore, the proposed amendment will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Robert A. Capra.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: April 18, 1997, as supplemented by 
letters dated October 10, 1997, and February 27, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3/4.7.6, ``Plant Systems--
Control Room Emergency Ventilation System.'' Additional Limiting 
Conditions for Operation would be added related to the availability of 
the station vent normal range radiation monitoring instrumentation. The 
associated TS bases would also be modified consistent with these 
changes. The staff's proposed no significant hazards consideration 
determination for the requested change was published on June 4, 1997 
(62 FR 30646).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power Station 
(DBNPS), Unit No. 1, in accordance with this change would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions, or assumptions are affected by the proposed changes.
    The proposed change to Limiting Condition for Operation (LCO) 
3.7.6.1 would include new required Action statements in the event 
that one or both channels of Station Vent Normal Range Radiation 
Monitoring instrumentation become inoperable. Under the proposed 
Action statements for inoperable Station Vent Normal Range Radiation 
Monitoring instrumentation, should the control room normal 
ventilation system be isolated and at least one train of the control 
room emergency ventilation system be placed in operation, these 
systems would be in a state equivalent to that which they would be 
in following an actual high radiation condition. These proposed 
changes have no bearing on the probability of an accident.
    The proposed change to the terminology utilized in Surveillance 
Requirement (SR) 4.7.6.1.e is an administrative change made to make 
the terminology consistent with the proposed new Action statements. 
The proposed changes to Bases 3/4.7.6 are administrative changes 
consistent with the proposed changes to LCO 3.7.6.1. These changes 
have no bearing on the probability of an accident.
    Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
change the source term, containment isolation, or allowable 
releases.
    As described above, under the proposed Action statements for 
inoperable Station Vent Normal Range Radiation Monitoring 
instrumentation, should the control room normal ventilation system 
be isolated and at least one train of the control room emergency 
ventilation system be placed in operation, these systems would be in 
a state equivalent to that which they would be in following an 
actual high radiation condition. Therefore, in the unlikely event of 
an accident requiring control room isolation while in this 
condition, the dose consequences to control room operators would be 
unchanged.
    The proposed change to the terminology utilized in Surveillance 
Requirement (SR) 4.7.6.1.e is an administrative change made to make 
the terminology consistent with the proposed new Action statements. 
The proposed changes to Bases 3/4.7.6 are administrative changes 
consistent with the proposed changes to LCO 3.7.6.1. These changes 
have no bearing on the consequences of an accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident

[[Page 25118]]

previously evaluated because no new accident initiators or 
assumptions are introduced by the proposed changes.
    As described above, under the proposed Action statements for 
inoperable Station Vent Normal Range Radiation Monitoring 
instrumentation, should the control room normal ventilation system 
be isolated and at least one train of the control room emergency 
ventilation system be placed in operation, these systems would be in 
a state equivalent to that which they would be in following an 
actual high radiation condition. Operation of the equipment and 
components in this manner would not introduce the possibility of any 
new or different kinds of accidents.
    The proposed change to the terminology utilized in Surveillance 
Requirement (SR) 4.7.6.1.e is an administrative change made to make 
the terminology consistent with the proposed new Action statements. 
The proposed changes to Bases 3/4.7.6 are administrative changes 
consistent with the proposed changes to LCO 3.7.6.1. These changes 
would not introduce the possibility of any new or different kinds of 
accidents.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes to the Action under LCO 3.7.6.1 ensure 
that control room isolation capability is maintained in the event a 
station vent radiation monitor is inoperable. The proposed allowable 
outage time of seven days for one inoperable channel is consistent 
with the presently allowable outage time for one inoperable CREVS. 
The proposed Action to place at least one CREVS train in operation 
within one hour, in the event both channels of radiation monitoring 
become inoperable, is more conservative than the present Action 
which requires that a plant shutdown commence within one hour, but 
does not require the CREVS be placed in operation.
    The proposed change to the terminology utilized in Surveillance 
Requirement (SR) 4.7.6.1.e is an administrative change made to make 
the terminology consistent with the proposed new Action statements. 
The proposed changes to Bases 3/4.7.6 are administrative changes 
consistent with the proposed changes to LCO 3.7.6.1. These changes 
would not affect 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 Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Acting Project Director: Richard P. Savio

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application request: March 9, 1998.
    Description of amendment request: The proposed amendment 
application would revise Technical Specification 3/4.5.2b.1 and its 
associated Bases to add clarification in regard to venting the 
emergency core cooling system (ECCS) pump casings and accessible 
discharge piping high points. Technical Specification 3/4.5.2b.1 
requires verification that the ECCS piping is full of water at least 
once per 31 days by venting the ECCS pump casings, i.e., the safety 
injection pump, residual heat removal pump, and centrifugal charging 
pump casings and accessible discharge piping high points. The 
centrifugal charging pump (CCP) casings do not have installed casing 
vents. Instead of a casing vent, the suction and discharge piping is 
installed as vertical runs attached to the top-mounted suction and 
discharge nozzles of each CCP pump. Information provided by the pump 
manufacturer indicates that the vertical configuration of the piping is 
sufficient to prevent the accumulation of noncondensible gases that 
could cause gas binding. Therefore the CCP casings are effectively 
vented by vents on the CCP discharge lines. The proposed amendment 
application would revise Technical Specification 3/4.5.2b.1 and 
associated Bases to require the residual heat removal and safety 
injection pump casings and accessible ECCS discharge piping high points 
be vented to ensure the ECCS piping is full of water.
    Basis for proposed no significant hazards consideration 
determination:
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will align the surveillance requirements 
with the installed system design and normal operating conditions. 
The performance of surveillances required by Technical 
Specifications is not postulated to initiate an accident. The intent 
of the surveillance ensures OPERABILITY of the ECCS by verifying 
that the ECCS piping is full of water and not subjected to gas 
binding or water hammer. The design of the CCPs is such that 
significant noncondensible gases do not collect in the pumps, 
whether they are running or not. Therefore, it is unnecessary to 
require periodic pump casing venting to ensure the CCPs will remain 
OPERABLE. In addition, operating experience has shown that no 
significant voiding has occurred in the affected piping which will 
continue to be vented at a high point every 31 days per Surveillance 
Requirement 4.5.2b.1). Therefore, no increase in the probability or 
consequences of an accident will occur as a result of this change.
    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 will not result in new failure modes because 
there are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The design of the CCPs is such that significant 
noncondensible gases do not collect in the pumps, whether they are 
running or not. Therefore, it is not necessary to require periodic 
pump casing venting to ensure the equipment will remain OPERABLE. 
Manual venting operations will be performed to minimize the 
potential for voids in system piping. Accordingly, this change will 
not create the possibility of 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 affect the acceptance criteria for 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protective functions. There will be no impact 
on any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Missouri-Columbia, 
Elmer Ellis Library, Columbia, Missouri 65201-5149
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: William H. Bateman

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: December 18, 1997.
    Description of amendment request: The proposed changes revise the 
Technical Specifications (TS) to clarify the terminology used to 
describe equipment surveillances performed with a refueling interval 
frequency. Currently the TS are somewhat ambiguous in the wording in 
this regard, and the proposed changes would adhere to the improved 
Standard TS

[[Page 25119]]

and make it clear whether the reactor must be shutdown when performing 
the test, or whether a ``refueling interval'' frequency (e.g., 18 
months) is intended. All of the clarifications are in Section 4 of the 
TS. In addition, minor typographical errors are being corrected, and an 
obsolete reference is proposed to be deleted.
    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:

    Criterion 1--Operation of Surry Units 1 and 2 in accordance with 
the proposed Technical Specifications change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The probability of an accident is not increased as a result of 
the proposed Technical Specification change since surveillance 
intervals are being clarified, not changed, and will continue to 
validate system/component availability, operability and performance 
during the appropriate unit mode. The proposed change is 
administrative in nature, therefore, station operations are not 
being affected. The consequences of an accident previously evaluated 
are not increased since station operations are not being changed, 
and no physical modifications are being made to plant systems or 
components.
    Criterion 2--The proposed Technical Specifications change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    As noted above, the proposed change is administrative in nature. 
A new or different type of accident is not being created since no 
new accident precursors are being introduced and equipment 
surveillances will continue to be performed as required to ensure 
proper system/component operation. Plant systems are not being 
modified, system operations are not being affected, and equipment 
surveillance intervals are not being increased. Consequently, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Criterion 3--The proposed Technical Specifications change does 
not involve a significant reduction in a margin of safety.
    This is an administrative change. Clarification of refueling 
surveillance interval terminology to ensure consistency in 
application does not affect plant equipment performance. 
Surveillance intervals are not being increased, and equipment 
surveillance tests performed on a refueling interval frequency (i.e. 
once per 18 months) will continue to ensure system/component 
performance as assumed in the existing safety analyses. Therefore, 
the proposed Technical Specification change does 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 Sec. 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: Swem Library, College of William 
and Mary, Williamsburg, Virginia 23185
Attorney for licensee: Michael W. Maupin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219
NRC Project Director: P.T. Kuo, Acting

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: March 25, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) Sections 6.1.A; 6.1.A.2; 
6.1.C.1.a and b; 6.1.C.1.f.1,4 and 8; 6.1.C.1.g.1 and 3; 6.8.A.2; and 
6.8.B.2 for Units 1 and 2, changing the title of Station Manager to 
Site Vice President, and the titles of the Assistant Station Managers 
to Manager-Station Operations and Maintenance and Manager-Station 
Safety and Licensing.
    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:

    Virginia Electric and Power Company has reviewed the proposed 
Technical Specifications changes against the criteria of 10 CFR 
50.92 and has concluded that the changes do not pose a significant 
hazards consideration. Specifically, station operations in 
accordance with the proposed Technical Specifications changes will 
not:
    a. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are administrative in nature. The overall 
responsibility for safe operation and review of plant operations is 
not being changed. There are no changes to the operation of any 
plant system or its design as a result of these changes. Therefore, 
neither the probability of occurrence nor the consequences of an 
accident or malfunction of equipment important to safety previously 
evaluated in the safety analysis report are increased.
    b. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes are administrative in nature. The overall 
responsibility for safe operation and review of plant operations is 
not being changed. There are no changes to the operation of any 
plant system or its design that could create any new modes of 
operation or accident precursors. Therefore, it is concluded that no 
new or different kind of accident or malfunction from any previously 
evaluated has been created.
    c. The proposed changes do not result in a significant reduction 
in margin of safety as defined in the basis for any Technical 
Specifications.
    The proposed changes are administrative in nature. The overall 
responsibility for safe operation and review is not being changed. 
There are no changes to the operation of any plant system or its 
design as a result of these changes. Safety systems are maintained 
operable as required by Technical Specifications. Therefore, the 
margin of safety is not changed.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

Local Public Document Room location: Swem Library, College of William 
and Mary, Williamsburg, Virginia 23185
Attorney for licensee: Michael W. Maupin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219
NRC Project Director: P.T. Kuo, Acting

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: April 8, 1998.
    Description of amendment request: The change would reduce allowable 
reactor coolant system (RCS) specific activity from 1.0 microcurie/gram 
to 0.35 microcurie/gram dose equivalent I-131.
    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:

    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    (1) Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The change implements a more restrictive RCS activity limit. 
Specific RCS activity is an initial plant condition and, therefore, 
is not an accident initiator and can not cause the occurrence of or 
increase the probability of an accident. The change also lowers the 
curve of Figure TS 3.1-3 which restricts operation with high 
specific activity. The new value for specific activity is justified 
by

[[Page 25120]]

the Westinghouse calculation which demonstrates acceptable offsite 
and control room doses following a (main steamline break) MSLB with 
a maximum allowable primary to secondary leak rate. By lowering the 
RCS specific activity and maintaining leakage within the projected 
maximum allowable, 10 CFR 100 and GDC 19 criteria are satisfied. 
Therefore, the change does not involve a significant 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 previously evaluated.
    The proposed change to the RCS specific activity limit will not 
significantly effect operation of the plant nor will it alter the 
configuration of the plant. There will be no additional challenges 
to the main steam system or the reactor coolant system pressure 
boundary and no new failure modes are introduced. Therefore, the 
proposed 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.
    Reduction of the RCS specific activity limit allows an increase 
in the MSLB allowable primary to secondary leakage. The net effect 
is no reduction in the margin of safety provided by 10 CFR part 100 
and GDC 19 criteria. The maximum allowable leakage is the leakage 
limit for projected SG leakage following SG tube inspection and 
repair. Reducing specific activity to increase projected leak rate 
follows guidance given by GL 95-05 and effectively takes margin 
available in the specific activity limits and applies it to the 
projected SG leak rate. This has been determined to be an acceptable 
means for accepting higher projected leak rates while still meeting 
the applicable limits of 10 CFR part 100 and GDC 19 criteria with 
respect to offsite and control room doses. Additionally, monitoring 
of the specific activity and compliance with the required actions 
remains unchanged. Therefore, the proposed change does not involve a 
significant reduction in the margin of safety.
    For consistency, the value of secondary coolant activity in 
Table TS 4.1.2 is being corrected from 1.0 microcurie/gram to 0.1 
microcurie/gram. This is consistent with a previously submitted and 
approved amendment, therefore, no significant hazards exist for this 
change.

    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 Wisconsin, Cofrin 
Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497
NRC Project Director: Richard P. Savio

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: April 15, 1998.
    Description of amendment request: The revisions in the proposed 
Technical Specification amendment are part of the licensee's fuel and 
reload change plan for Cycle 23. The revisions implement changes 
associated with a new fuel design and also reflect changing plant 
conditions due to steam generator tube plugging and repair. The 
Technical Specifications (TS) would be modified as follows:
    (1) Figure 2.1-1 would be revised to reflect the recently approved 
High Thermal Performance (HTP) Critical Heat Flux (CHF) correlation and 
corresponding Departure from Nucleate Boiling Ratio (DNBR) limit of 
1.14. The figure would also reflect changes in peak rod power and 
minimum reactor coolant flow.
    (2) TS 3.10.b--new hot channel factors would be incorporated for 
the new fuel design and the corresponding increase in peaking factors. 
The limits for Height Dependent Nuclear flux Hot Channel Factor are 
specified in TS 3.10.b.1 and the limits for Nuclear Enthalpy Rise Hot 
Channel Factor are specified in 3.10.b.2.
    (3) TS 3.10.k--the specification for the maximum Reactor Coolant 
System (RCS) Inlet Temperature would be replaced with a specification 
for the maximum Reactor Coolant System (RCS) Average Temperature.
    (4) TS 3.10.l--the statement ``During 100% steady-state power 
operation'' would be revised in the specification for minimum Reactor 
Coolant System (RCS) pressure and replaced with ``During steady-state 
power operation.''
    (5) TS 3.10.m--the minimum Reactor Coolant Flow is being decreased 
to 85,500 gallons per minute per loop.
    (6) TS 3.10.n--would be revised to reflect the new Minimum DNBR 
limit.
    (7) Figure TS 3.10-1--the Required Shutdown Reactivity vs. Boron 
Concentration would be revised to reflect the change to an 18 month 
fuel cycle.
    (8) Figure TS 3.10-2, the Hot Channel Factor Normalized Operating 
Envelope would be revised to reflect the values used in the new safety 
analyses.
    (9) The Table of Contents and the Basis sections would be revised 
to accommodate the above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Figure TS 2.1-1: The proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The safety limits curves are not accident initiators. Therefore, 
the change will not increase the probability of an accident 
previously evaluated. The proposed changes to the safety limits 
curves do not alter the plant configuration, operating set points, 
or overall plant performance. The safety limits curves reflect the 
changes to the DNBR limit, CHF correlation, RCS flow peaking factors 
and fuel design. The significant hazards determinations for these 
parameters are evaluated later in this submittal. Therefore, the 
change will not increase the consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes in the safety limits curves do not alter 
the plant configuration, operating set points, or overall plant 
performance. Therefore, it does not create the possibility of a new 
or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    Operation in the acceptable regions (i.e., below and to the left 
of the safety limit curves) in combination with the reactor 
protection and engineered safety systems designed into the plant 
will ensure that the safety limits are not exceeded during normal 
operation or during anticipated design basis operational transients. 
The core will be operated in the nucleate boiling heat transfer 
regime. Departure from nucleate boiling (DNB) will not occur and 
therefore fuel cladding integrity will be assured.
    The revised safety limit curves have been developed using 
operating parameters at their bounding values (e.g., rod powers at 
the peaking factor limits, reactor coolant flow at the minimum 
operating limit). The revised curves will bound plant operation with 
Siemens Power Corporation standard or heavy fuel. Therefore, this 
change will not involve a significant reduction in safety margin.
    TS 3.10.b: The proposed changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Peaking factor limits are input assumptions to the safety 
analyses and are not accident initiators. Therefore, this change 
would not increase the probability of occurrence of an accident 
previously evaluated.
    The safety analyses input assumptions are designed to bound 
actual plant operation. Changing the safety analysis input 
assumption for the increased peaking factor limits does not change 
the underlying progression of design basis accidents evaluated in 
the safety analyses. All safety analysis acceptance criteria are 
satisfied in the increased peaking factor limit conditions. 
Additionally, the radiological consequences

[[Page 25121]]

are bounded by existing analysis at the increased peaking factor 
limits. Therefore, this change will not significantly increase the 
consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This change incorporates the safety analyses assumptions for 
core peaking factor limits for Siemens Power Corporation heavy fuel. 
The change does not alter plant equipment, set points or plant 
performance. Therefore, changing the peaking factor limits for 
analysis purposes will not create a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    Results of the safety analyses and of radiological consequences 
indicate that all acceptance criteria are satisfied. The peaking 
factor limits assumed in the safety analyses are consistent with the 
proposed revised limits and these revised limits are established to 
bound actual plant operation. Therefore, this change will not 
involve a significant reduction in the margin of safety.
    TS 3.10.k: The proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The RCS average temperature limit is not an accident initiator. 
Changing the technical specification limit consistent with the 
accident analyses will not increase the probability of an accident 
previously evaluated.
    The proposed change limits the maximum reactor coolant system 
average temperature to 568.8  deg.F. The design basis safety 
analyses, the Large and Small Break LOCA accidents and the non-LOCA 
accidents, have been analyzed and/or evaluated consistent with the 
revised RCS average temperature. The re-analysis and evaluation have 
demonstrated that all safety analysis acceptance criteria are 
satisfied at the specified temperature. Therefore, the change will 
not increase the consequences of an accident previously evaluated.
    The proposed technical specification limit for maximum allowed 
RCS average temperature was decreased below the analytical limit to 
account for instrument error.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not alter the plant configuration, 
operating set points, or overall plant performance. Therefore, it 
does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed change is consistent with the safety analyses. All 
safety analyses acceptance criteria are satisfied at the revised 
reactor coolant system average temperature. The TS limit will bound 
actual plant operation. Therefore, there is no significant reduction 
in the margin of safety.
    TS 3.10.l: The proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The RCS pressure limit is not an accident initiator. By removing 
the 100% value from the specification, the assumptions in the safety 
analyses are not changed. Changing the technical specification to 
remove the 100% power criteria will not increase the probability of 
an accident previously evaluated.
    The design basis safety analyses have been analyzed and/or 
evaluated at the specified RCS pressure. The analyses and 
evaluations have demonstrated that all safety analyses acceptance 
criteria are satisfied at this pressure. Therefore, the change would 
not increase the consequences of an accident previously evaluated.
    The proposed technical specification limit for minimum allowed 
RCS pressure was increased above the analytical limit to account for 
instrument error.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not alter the plant configuration, 
operating set points, or overall plant performance. Therefore, it 
does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed change is consistent with the safety analyses. All 
safety analyses acceptance criteria are satisfied at the reactor 
coolant system pressure. The limit will bound actual plant 
operation. Therefore, there is no significant reduction in the 
margin of safety.
    TS 3.10.m: The proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The RCS flow limit is not an accident initiator. Changing the 
technical specification limit consistent with the accident analysis 
will not increase the probability of an accident previously 
evaluated.
    The proposed change limits the minimum reactor coolant flow. The 
design basis safety analyses have been analyzed and/or evaluated at 
the revised RCS flow. The re-analysis and evaluation have 
demonstrated that all safety analysis acceptance criteria are 
satisfied at the specified flow. Therefore, the change will not 
significantly increase the consequences of an accident previously 
evaluated.
    The proposed technical specification limit for minimum allowed 
RCS flow was increased above the analytical limit to account for 
instrument error.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not alter the plant configuration or 
overall plant performance. Therefore, it does not create the 
possibility of a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed change is consistent with the safety analyses. All 
safety analyses acceptance criteria are satisfied at the revised 
reactor coolant system flow. The limit will bound actual plant 
operation.
    The change reduces the RCS flow rate limit. Re-analysis of LOCA 
and non-LOCA transients determined all safety requirements of KNPP 
accident analyses were still met at the reduced RCS flow rate limit. 
Therefore, this proposed change does not significantly reduce the 
margin of safety.
    TS 3.10.n: The proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The Departure from Nucleate Boiling Ratio (DNBR) is not an 
accident initiator. Therefore, the change in the DNBR will not 
increase the probability of an accident previously evaluated.
    The proposed change to the DNBR value does not change plant 
configuration, operating set points, or overall plant performance. 
Therefore, the change will not increase the consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not alter the plant configuration, 
operating set points, or overall plant performance. Therefore, it 
does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    All safety analyses acceptance criteria are satisfied using the 
HTP CHF correlation. The DNBR limits assumed in the safety analyses 
will bound actual plant operation and assures at 95/95 that DNBR 
will not occur. Therefore, there is no reduction in the margin of 
safety.
    TS Figure 3.10-1: The proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Required Shutdown Reactivity vs. Boron Concentration was revised 
to reflect the longer cycle length and the resulting increase in 
boron concentration. The Required Shutdown Reactivity vs. Boron 
Concentration is not an accident initiator. Extending the boron 
concentrations to account for longer fuel cycles will not increase 
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.
    The proposed change does not alter the plant configuration, 
operating set points, or overall plant performance. Therefore, it 
does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed change is consistent with the cycle length and core 
physics analyses for longer fuel cycles. Operation within the limits 
specified in the figure will assure all core safety evaluation 
acceptance criteria are satisfied. The limit will bound actual plant 
operation. Therefore, there is no reduction in the margin of safety.
    TS Figure 3.10-2: The proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

[[Page 25122]]

    The Hot Channel Factor Normalized Operating Envelope figure was 
revised to reflect the values used in the safety analyses.
    The Hot Channel Factor Normalized Operating Envelope figure is 
not an accident initiator. Changing the technical specification 
figure consistent with the assumptions of the accident analyses will 
not increase 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.
    The proposed change does not alter the plant configuration, 
operating set points, or overall plant performance. Therefore, it 
does not create the possibility of a new or different kind of 
accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed change is consistent with the safety analyses. 
Operation within the limits specified in the figure will assure all 
safety analyses acceptance criteria are satisfied. The limit will 
bound actual plant operation. Therefore, there is no 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 Wisconsin, Cofrin 
Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497
NRC Project Director: Richard P. Savio

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point 
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: May 2, 1995, October 12, 1995, March 26, 
1996, and December 15, 1997 (TSCR 172)
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) Table 15.4.1-1, ``Minimum 
Frequencies for Checks, Calibrations, and Tests of Instrument 
Channels,'' to change the test frequencies for radiation monitors as 
discussed in Generic Letter 93-05 (``Line-Item Technical Specifications 
Improvements To Reduce Surveillance Requirements For Testing During 
Power Operation''), remove the radiation monitoring system as item 36, 
revise note(s), and add those radiation monitors and their surveillance 
requirements that support current TS or meet the requirements of 10 CFR 
50.36. Additionally, several typographical and nomenclature errors 
would be corrected. This amendment request was initially noticed in the 
Federal Register on June 6, 1995 (60 FR 29890).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Operation of this facility under the proposed TS will not create 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The probabilities of accidents previously evaluated are based on 
the probability of initiating events for these accidents. Initiating 
events for accidents previously evaluated for the Point Beach Nuclear 
Plant (PBNP) include control rod withdrawal and drop, chemical volume 
control system malfunction (boron dilution), startup of an inactive 
reactor coolant loop, reduction in feedwater enthalpy, excessive load 
increase, losses of reactor coolant flow, loss of external electrical 
load, loss of normal feedwater, loss of all alternating current (ac) 
power to the auxiliaries, turbine overspeed, fuel handling accidents, 
accidental releases of waste liquid or gas, steam generator tube 
rupture, steam pipe rupture, control rod ejection, and primary coolant 
system ruptures.
    These proposed changes do not cause an increase in the 
probabilities of any accidents previously evaluated because these 
changes will not cause an increase in the probability of any initiating 
events for accidents previously evaluated. In particular, these changes 
affect the radiation monitoring system surveillance requirements and 
make administrative changes that will not result in changing accident 
initiators.
    The consequences of the accidents previously evaluated in the Final 
Safety Analysis Report (FSAR) are determined by the results of analyses 
that are based on initial conditions of the plant, the type of 
accident, transient response of the plant, and the operation and 
failure of equipment and systems.
    The proposed changes reduce the burden associated with radiation 
monitoring system required surveillance by establishing surveillances 
for only the necessary monitors (i.e., elimination of the testing 
requirement for monitors that do not perform a required function) and 
changing the testing frequency for these monitors from monthly to 
quarterly. The proposed changes do not increase the probability of 
failure of this equipment or its ability to operate as required for the 
accidents previously evaluated in the PBNP FSAR. The proposed changes 
to correct typographical errors and correct nomenclature are 
administrative only and do not increase the probability of an accident 
previously evaluated nor do they affect the consequences of any 
accident previously evaluated.
    Therefore, these proposed license amendments do not affect the 
consequences of any accident previously evaluated in the PBNP FSAR 
because the factors that are used to determine consequences of 
accidents are not being changed.
    2. Operation of this facility under the proposed TS change will not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    New or different kinds of accidents can only be created by new or 
different accident initiators or sequences. The changes proposed by 
this license amendment request do not create any new or different 
accident initiators or sequences because the revisions to TS Table 
15.4.1-1, ``Minimum Frequencies for Checks, Calibrations, and Tests of 
Instrument Channels,'' will not cause failures of equipment or accident 
sequences different than the accidents previously evaluated. The 
proposed changes to correct typographical errors and correct 
nomenclature are administrative only. Therefore, these proposed TS 
changes do not create the possibility of an accident of a different 
type than any previously evaluated in the Point Beach FSAR.
    3. Operation of this facility under the proposed TS change will not 
create a significant reduction in a margin of safety.
    The margins of safety for Point Beach are based on the design and 
operation of the reactor and containment and the safety systems that 
provide their protection. The changes proposed by this license 
amendment request provide the appropriate surveillance requirements for 
the radiation monitoring system. The revised surveillance requirements 
will continue to ensure that the required radiation monitors will 
operate as required. The design and operation of the reactor and 
containment are not affected by these proposed changes. The proposed 
changes to correct typographical errors and correct nomenclature are 
administrative only. Therefore, the margins of safety for Point Beach 
are not being reduced because the design and operation of the reactor 
and containment are not being changed.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff

[[Page 25123]]

proposes to determine that the amendment request involves no 
significant hazards considerations.

Local Public Document Room location: The Lester Public Library, 1001 
Adams Street, Two Rivers, Wisconsin 54241
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Cynthia A. Carpenter

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.

Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick 
Steam Electric Plant, Unit 1, Brunswick County, North Carolina

    Date of amendment request: February 23, 1998, as supplemented March 
27, 1998.
    Brief description of amendment: The proposed amendment would allow 
addition of a footnote to the Safety Limit Minimum Critical Power Ratio 
value in the Technical Specifications and the associated action 
statement.
    Date of publication of individual notice in the Federal Register: 
April 10, 1998 (63 FR 17900).
    Expiration date of individual notice: May 11, 1998.

Local Public Document Room location: University of North Carolina at 
Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: April 3, 1998, and related application 
dated November 22, 1995, as supplemented February 19, April 19, May 3, 
June 12, and December 4, 1996, and January 30 and August 7, 1997.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.8.1.1 to change the emergency diesel 
generator allowed outage time from 3 to 7 days. This would be a one-
time amendment, effective from the date of issuance until September 30, 
1998.
    Date of publication of individual notice in Federal Register: April 
13, 1998 (63 FR 18048).
    Expiration date of individual notice: May 13, 1998.

Local Public Document Room location: Monroe County Library System, 3700 
South Custer Road, Monroe, Michigan 48161

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: April 9, 1998, TXX-98107.
    Description of amendment request: The proposed amendment would 
allow on a one time basis, the verification of the proper operation of 
the Unit 2 load shed seal-in contacts and the diesel generator trip 
bypass contacts at power and crediting performance of Surveillance 
Requirements (SR) 4.8.1.1.2f.4(a) and 4.8.1.1.2f.6(a), at power as 
opposed to ``during shutdown'' as currently required by those SR. The 
proposed amendment would also allow on a one time basis the 
verification of the proper operation of the Unit 2 lockout relays and 
contacts to be deferred until the startup from 2RFO4 or earlier outage 
to at least MODE 3.
    Date of individual notice in the Federal Register: April 20, 1998.
    Expiration date of individual notice: May 5, 1998.

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

    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 Ch. 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.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of application for amendment: March 17, 1997, as supplemented 
April 13, 1998. The April 13, 1998, submittal contained clarifying 
information only, and did not change the proposed no significant 
hazards consideration.
    Brief description of amendment: The amendment revises Technical 
Specifications 4.1.2.2.c, 4.5.2.e, 4.6.2.1.c, 4.6.2.2.c, 4.6.3.2, 
4.7.1.2.1.b, 4.7.3.b, and 4.7.4.b to delete specific restrictions in 
the text of the surveillances that the tests must be done while the 
unit is shut down.
    Date of issuance: April 14, 1998.
    Effective date: April 14, 1998
    Amendment No.: 77.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19826)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 1998.

[[Page 25124]]

    No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional Library, 
1930 Clark Avenue, Raleigh, North Carolina 27605

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

    Date of application for amendments: December 12, 1997.
    Brief description of amendments: The amendments modify the bypass 
logic for Main Steam Line Isolation Valve Isolation Actuation 
Instrumentation on Condenser Low Vacuum as stated in Technical 
Specification Tables 3.3.2-1 and 4.3.2-1.
    Date of issuance: April 14, 1998.
    Effective date: Immediately, to be implemented prior to startup 
from L1F35 for Unit 1 and from L2R07 for Unit 2.
    Amendment Nos.: 124 and 109.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6982).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 14, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Jacobs Memorial Library, Illinois 
Valley Community College, Oglesby, Illinois 61348

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: December 18, 1997, as 
supplemented by letter dated January 26, 1998.
    Brief description of amendments: The amendments revise the 
operating license of Unit 1 and Unit 2 to (1) delete license conditions 
that have been fulfilled; (2) delete exemptions that have expired; (3) 
update information to reflect current plant status and regulatory 
requirements; and (4) make other corrections and editorial changes.
    Date of issuance: April 23, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-164; Unit 2-156.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6983).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 23, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: March 3, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications to change the qualification requirements for 
the members of the Safety Review Group.
    Date of issuance: April 27, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-165; Unit 2-157.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1998 (63 FR 
14486).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 27, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South 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 application of amendments: August 28, 1997. Supplement 
January 22, February 19, March 19, and April 6, 13, and 17, 1998.
    Brief description of amendments: The amendments incorporate new 
testing and operability requirements related to the installation of new 
systems and upgrades associated with the Emergency Condenser 
Circulating Water System. Review of the system for this amendment also 
includes a review of the new design features incorporated into the 
upgrade and its acceptability as a safety grade system.
    Date of Issuance: April 24, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-229; Unit 2-230; Unit 3-226
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications and Appendix C of the 
Operating Licenses.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50002).
    The January 22, 1998, February 19, March 19, and April 6, 13, and 
17, 1998, letters provided clarifying information that did not change 
the scope of the August 28, 1997, application and the initial proposed 
no significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated April 24, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501 West 
South Broad Street, Walhalla, South Carolina

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: March 16, 1998.
    Brief description of amendments: These amendments add a new 
Limiting Condition for Operation (LCO) 3.0.6 to TS Section 3/4.0, 
``APPLICABILITY.'' The new LCO 3.0.6 provides specific guidance for 
returning equipment to service under administrative control to perform 
testing required to demonstrate OPERABILITY.
    Date of issuance: April 15, 1998.
    Effective date: Both units, effective immediately, to be 
implemented within 30 days.
    Amendment Nos.: 213 and 90.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (63 FR 14142, March 24, 1998). That notice provided 
an opportunity to submit comments on the Commission's proposed no 
significant hazards consideration determination. No comments have been 
received. The notice also provided for an opportunity to request a 
hearing by April 23, 1998, but indicated that if the Commission makes a 
final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated April 
15, 1998.


[[Page 25125]]


Local Public Document Room location: B.F. Jones Memorial Library, 663 
Franklin Avenue, Aliquippa, PA 15001

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

    Date of amendment request: June 26, 1997, as supplemented by letter 
dated September 11, 1997.
    Brief description of amendment: The amendment changes the Appendix 
A TSs by modifying Tables 3.7-1 and 3.7-2. The revision to Table 3.7-1 
changes the Main Steam Safety Valves (MSSVs) orifice size from 26 
square inches to 28.27 square inches and relocates the orifice size 
from the TS Table to the TS Bases. The change to correct the orifice 
size is an editorial change to make the TS consistent with plant 
design. The changes to Table 3.7-2 delete the provisions that allows 
continued plant operation with three MSSVs inoperable. The proposed 
amendment will also revise TS Bases 3/4.7.1.1 to remove the equation 
used for determining the reduced maximum allowable linear power level-
high reactor trip settings of TS Table 3.7-2.
    Date of issuance: April 20, 1998.
    Effective date: April 20, 1998, to be implemented within 30 days.
    Amendment No.: 142.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38135).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 20, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans Library, 
Louisiana Collection, Lakefront, New Orleans, LA 70122

GPU Nuclear, Inc. and Saxton Nuclear Experimental Corporation (SNEC), 
Docket No. 50-146, Saxton Nuclear Experimental Facility (SNEF)

    Date of application for amendment: November 25, 1996, as 
supplemented on May 30, June 4 and 16, August 21 and September 16, 
1997, and February 3 and 9, 1998, and March 31, 1998. During the 
amendment request review, the staff also referred to the SNEF 
Decommissioning Environmental Report dated April 17, 1996, licensee 
responses to NRC questions about the environmental report dated July 
18, 1996, and March 3 and 31, 1998, the SNEC Facility Updated Safety 
Analysis Report, Revision 0, submitted on October 25, 1996, Revision 1, 
submitted on August 21, 1997, and Revision 2, submitted on February 3, 
1998, and the SNEC Facility Decommissioning Quality Assurance Plan 
submitted by letter dated November 8, 1996, as supplemented on May 30, 
1997, and February 3 and 9, 1998.
    Brief description of amendment: The amendment allows 
decommissioning of the SNEF. The changes to the license and Technical 
Specifications (TSs) (1) accommodate decommissioning activities at the 
SNEF, (2) establish specific TS controls over decommissioning 
activities, (3) establish limiting conditions for performing 
decommissioning activities, (4) extend exclusion area controls to 
include the SNEF Decommissioning Support Facility, (5) establish 
requirements for a Radiological Environmental Monitoring Program, and 
an Offsite Dose Calculation Manual, and (6) establish requirements for 
technical and independent safety reviews. In addition, the amendment 
authorizes other administrative and editorial changes to the TSs 
associated with the changes described above.
    Date of issuance: April 20, 1998.
    Effective date: April 20, 1998.
    Amendment No.: 15.
    Amended Facility License No. DPR-4: Amendment changed the Amended 
Facility License and TSs.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11494).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 20, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room Location: Saxton Community Library, Front 
Street, Saxton, Pennsylvania 16678

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: December 16, 1996, as 
supplemented September 11, 1997 and March 25, 1998.
    Brief description of amendment: The amendment (1) reflects the 
change in the legal name of the operator of TMI-1 from GPU Nuclear 
Corporation to GPU Nuclear, Inc., and (2) reflects in the TMI-1 
Facility Operating License the registered trade name of GPU Energy now 
used by the owners of the facility.
    Date of Issuance: April 24, 1998.
    Effective Date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 207.
    Facility Operating License No. NPF-50: Amendment revised the 
Facility Operating License and the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4350).
    The September 11, 1997 and March 25, 1998, submittals provided 
clarifying information and 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 April 24, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit No. 2, Oswego County, New York

    Date of application for amendment: October 7, 1997.
    Brief description of amendment: The amendment revised the Technical 
Specifications surveillance requirements to change setpoints for the 
refueling platform main hoist overload cutoff, loaded interlock, and 
redundant loaded interlock due to planned modifications to the 
refueling platform mast.
    Date of issuance: April 16, 1998.
    Effective date: As of the date of issuance to be implemented upon 
completion and acceptance of design modifications to the refueling 
platform mast.
    Amendment No.: 81.
    Facility Operating License No. NMF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68309).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 16, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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

    Date of application for amendment: March 13, 1998, as supplemented 
March 25, 1998.

[[Page 25126]]

    Brief description of amendment: The amendment modifies the 
Technical Specification requirements associated with the Minimum 
Critical Power Ratio (MCPR) safety limits for Cycle 19 based on the 
cycle-specific analysis of the current mixed core of GE [General 
Electric] 11, GE10, four GE12 lead use assemblies, and eight SPC 
[Siemens Power Corporation] ATRIUM-9B assemblies.
    Date of issuance: April 20, 1998.
    Effective date: April 20, 1998.
    Amendment No.: 100.
    Facility Operating License No. DPR-22: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 20, 1998 (63 FR 
13704).
    The March 25, 1998, letter provided clarifying information in 
response to the staff's request for additional information during a 
teleconference. 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 April 20, 1998.
    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

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

    Date of application for amendment: October 14, 1997, as 
supplemented on March 26, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.6.3, ``Primary Coolant System Pressure Isolation 
Valves Limiting Condition for Operation,'' to add additional pressure 
isolation valves, establish the operability and testing requirements 
for the pressure isolation valves, and make this section more 
consistent with Salem Unit 2 TSs.
    Date of issuance: April 20, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-70: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61845).
    The March 26, 1998, 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 April 20, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: January 26, 1998.
    Brief description of amendments: The proposed amendments would (1) 
modify the requirement to hold a Susquehanna Steam Electric Station 
(SSES) Senior Reactor Operator (SRO) license in Section 6.3.1 for the 
Manager-Nuclear Operations (MNO), (2) replace the position of MNO with 
Operations Supervisor--Nuclear in the Section 6.2.2g requirement to 
hold an SSES SRO license and (3) renumber existing TS Section 6.3.1 to 
include 6.3.1.1, 6.3.1.2, and 6.3.1.3.
    Date of issuance: April 10, 1998.
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 175 and 147.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1998 (63 
FR 9270).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 10, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library, Reference 
Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

Pennsylvania Power and Light Company, Docket No. 50-388, Susquehanna 
Steam Electric Station, Unit 2, Luzerne County, Pennsylvania

    Date of application for amendment: January 11, 1996, as 
supplemented March 16, 1998.
    Brief description of amendment: This amendment changes the TSs to 
preclude the need to enter into Limiting Condition for Operation 3.0.3 
to allow performance of certain emergency diesel generator testing.
    Date of issuance: April 10, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 148.
    Facility Operating License No. NPF-22: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 13, 1996 (61 FR 
10397).
    The February 15, 1996, letter corrected the no significant hazards 
(NSH) determination. The NSH determination was used in the March 13, 
1996 (61 FR 10397) notice. The March 24, 1998, 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 April 10, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library, Reference 
Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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

    Date of application for amendment: January 12, 1998.
    Brief description of amendment: This amendment revises TS Table 
4.4.6.1.3-1 to change the withdrawal schedule for the first capsule to 
be withdrawn from 10 Effective Full Power Years (EFPY) to 15 EFPY. In 
addition, TS Surveillance Requirement 4.4.6.1.4 will be revised to 
remove the references to flux wire removal and analysis that was 
originally required following the first cycle of operation and replaced 
with a new surveillance requirement. The new requirement refers to the 
flux wires that are located within the surveillance capsules, which 
will be removed and analyzed in accordance with the surveillance 
capsule removal schedule located in Table 4.4.6.1.3-1.
    Date of issuance: April 15, 1998.
    Effective date: As of the date of issuance.
    Amendment No.: 126.
    Facility Operating License No. NPF-39: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6988).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 1998.
    No significant hazards consideration comments received: No.


[[Page 25127]]


Local Public Document Room location: Pottstown Public Library, 500 High 
Street, Pottstown, PA 19464

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: February 27, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications by revising the pressure-temperature curves to extend 
heatup and cooldown limits from 11 to 13.3 effective full-power years, 
provides the corresponding overpressure protection system limits, and 
makes some minor changes to ensure specification clarity and 
conservatism.
    Date of issuance: April 10, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 179.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 9, 1998 (63 FR 
11456).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library, 100 
Martine Avenue, White Plains, New York 10610

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: February 26, 1998, as 
supplemented by letter dated March 20, 1998.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Section 3/4.4.5, ``Reactor Coolant System--Steam 
Generators,'' TS Section 3/4.4.6.2, ``Reactor Coolant System--
Operational Leakage,'' and the associated bases to allow use of the 
``repair roll'' steam generator tube repair process.
    Date of issuance: April 14, 1998.
    Effective date: April 14, 1998.
    Amendment No.: 220.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 9, 1998 (63 FR 
11460).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 1998.
    No significant hazards consideration comments received: No. The 
supplemental information submitted by the licensees did not affect the 
proposed no significant hazards consideration determination.
Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit 1, Ottawa County, Ohio

    Date of application for amendment: June 24, 1997.
    Brief description of amendment: This amendment revises TS Section 
3/4.3.2.1, ``Safety Features Actuation System Instrumentation,'' TS 
Section 3/4.6.1.7, ``Containment Ventilation System,'' TS Section 3/
4.6.3.1, ``Containment Isolation Valves,'' and TS Section 3/4.9.4, 
``Refueling Operations--Containment Penetrations,'' and the associated 
TS Bases. Valve position requirements have been added, and certain 
containment radiation monitor requirements, valve isolation 
verification requirements, and containment radiation monitor optional 
uses have been deleted. Administrative changes have also been made.
    Date of issuance: April 15, 1998.
    Effective date: April 15, 1998.
    Amendment No.: 221.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40858).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606

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 requests: February 25, 1998, (TXX-98050) as 
supplemented by letter dated March 9, 1998, (TXX-98066) for License 
Amendment Request (LAR) 98-002, March 12, 1998, (TXX-98076) for LAR 98-
003, and March 18, 1998, (TXX-98079) for LAR 98-004.
    Brief description of amendments: This amendment is the result of 
three Notice of Enforcement Discretions (NOEDs) dated February 24, 
March 13, and 17, 1998. These NOEDs although distinct actions changed 
the same page of the CPSES TS therefore the single amendment is being 
issued to cover the three parts of this amendment.
    The first part of the amendment would be a temporary change to the 
TSs to remove the requirement to demonstrate the load shedding feature 
of MCC XEB4-3 as part of Surveillance Requirements (SRs) 
4.8.1.1.2f.4)a) and 4.8.1.1.2f.6)a) until the plant startup subsequent 
to the next refueling outage for Unit or until an outage of 24 hour in 
duration.
    The second part of the amendment would provide a temporary 
Technical Specification change for SRs 4.8.1.1.2f.4)b) and 
4.8.1.1.2f.6)b) to allow the verification of the auto connected shut-
down loads through the load sequencer to be performed at power for fuel 
cycle 6 on Unit 1 and fuel cycle 4 on Unit 2.
    The third part of the amendment would allow on a one time basis, 
crediting performance of Surveillance Requirements (SR) 4.8.1.1.2f.4)a) 
and 4.8.1.1.2f.6)a), during POWER OPERATIONS as opposed to ``during 
shutdown.'' Note that the bus tie breaker for MCC XEB4-3 for Unit 2 was 
not tested during the last surveillance test and was the subject of 
part one of this amendment.
    Date of issuance: April 20, 1998.
    Effective date: April 20, 1998.
    Amendment Nos.: Unit 1--Amendment No. 58; Unit 2--Amendment No. 44.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 9, 1998 (63 FR 
11458), March 27, 1998 (63 FR 14974) and April 2, 1998 (63 FR 16287).
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated April 20, 
1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at Arlington 
Library, Government Publications/Maps, 702 College, PO Box 19497, 
Arlington, TX 76019

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: October 31, 1997, as 
supplemented by letter dated February 27, 1998.
    Brief description of amendment: The amendment revises the Callaway 
Plant,

[[Page 25128]]

Unit 1 Technical Specifications to change setpoint and allowable stress 
values of certain reactor trip system (RTS) and engineered safety 
features actuation system (ESFAS) functional units.
    Date of issuance: April 13, 1998.
    Effective date: April 13, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 125.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 FR 
2283).
    The February 27, 1998, supplemental letter provided additional 
clarifying information that did not change the staff's original no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 13, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: University of Missouri-Columbia, 
Elmer Ellis Library, Columbia, Missouri 65201-5149

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: December 11, 1997, as 
supplemented on March 3, 1998.
    Brief description of amendment: The amendment revises the values 
for the safety limit minimum critical power ratio for Cycle 20 
operation.
    Date of Issuance: April 10, 1998.
    Effective date: April 10, 1998, to be implemented within 30 days.
    Amendment No.: 159.
    Facility Operating License No.DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998, (63 
FR 7000).
    The March 3,1998 supplement did not change the original proposed no 
significant hazards consideration.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 10, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224 Main 
Street, Brattleboro, VT 05301

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont 
Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: September 11, 1996, as 
supplemented by letter dated December 8, 1997.
    Brief description of amendment: The amendment involves a change to 
the safety and relief valve setpoint tolerance and power operation with 
an inoperable safety relief valve.
    Date of Issuance: April 15, 1998.
    Effective date: April 15, 1998, to be implemented within 30 days.
    Amendment No.: 160.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17241).
    The information provided in the December 8, 1997, submittal did not 
change the original proposed no significant hazards determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 15, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224 Main 
Street, Brattleboro, VT 05301

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: November 26, 1996.
    Brief description of amendments: The proposed action would revise 
the Technical Specifications (TS) to eliminate the records retention 
requirements from Section 6.10 of the TS since these requirements have 
already been relocated to the Operational Quality Assurance program, 
Chapter 17, in revision 32 of the Updated Final Safety Analysis Report.
    Date of issuance: April 13, 1998.
    Effective date: April 13, 1998.
    Amendment Nos.: 208 and 189.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
132).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 13, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: February 3, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Surveillance Requirement Tables 3.3-1 and 
4.3-1 for both units, modifying the testing requirements for the 
reactor trip bypass breaker.
    Date of issuance: April 14, 1998.
    Effective date: April 14, 1998.
    Amendment Nos.: 209 and 190.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: March 11, 1998 (63 FR 
11925).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 14, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: November 18, 1997.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Surveillance Requirements 4.7.1.7.2.a.1 
and 4.7.1.7.2.a.2 for both units, modifying the testing frequency of 
the Turbine throttle and Governor valves.
    Date of issuance: April 16, 1998.
    Effective date: April 16, 1998.
    Amendment Nos.: 210 and 191.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66146)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 16, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498

[[Page 25129]]

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: February 3, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Surveillance Requirement 4.4.10.1.1, 
modifying the inspection requirements for the Reactor Coolant Pump 
(RCP) flywheels for both units and eliminating the examination 
requirements for the flow straighteners in each steam generator to the 
RCP elbow on Unit 1.
    Date of issuance: April 22, 1998.
    Effective date: April 22, 1998.
    Amendment Nos.: 211 and 192.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: March 11, 1998 (63 FR 
11924)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 22, 1998.
    No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498

    Dated at Rockville, Md., this 29th day of April 1998.

    For the Nuclear Regulatory Commission.
Stuart A. Richards,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-11911 Filed 5-5-98; 8:45 am]
BILLING CODE 7590-01-P