[Federal Register Volume 64, Number 66 (Wednesday, April 7, 1999)]
[Notices]
[Pages 17021-17040]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-8503]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission 
(the Commission or NRC staff) is publishing this regular biweekly 
notice. Pub. L. 97-

[[Page 17022]]

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 March 13, 1999, through March 26, 1999. The 
last biweekly notice was published on March 24, 1999 (64 FR 14278).

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. 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 April 23, 1999, 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

[[Page 17023]]

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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos. 
1, 2, and 3, Maricopa County, Arizona

    Date of amendments request: February 26, 1999.
    Description of amendments request: The proposed amendment would 
revise Technical Specification (TS) 3.5.3, ``Emergency Core Cooling 
System--Operating,'' to extend the completion time for one inoperable 
low pressure safety injection subsystem from 72 hours to 7 days.
    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 amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed amendment will extend the Completion Time 
for one inoperable low pressure safety injection (LPSI) subsystem in 
Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems 
(ECCE)[S]--Operating, from 72 hours to 7 days. The LPSI subsystem is 
part of the ECCS train and part of the shutdown cooling subsystem. 
The LPSI components are not accident initiators in any accident 
previously evaluated. Therefore, this change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The LPSI system is primarily designed to mitigate the 
consequences of a large break loss of coolant accident (LOCA). These 
proposed changes do not affect any of the assumptions used in the 
deterministic LOCA analysis.
    In order to evaluate the LPSI Completion Time extension with 
respect to the ECCS, probabilistic safety analysis (PSA) methods 
were utilized. The results of these analyses show no significant 
increase in the core damage frequency. As a result, there would be 
no significant increase in the consequences of an accident 
previously evaluated. These analyses are detailed in CE NPSD-995, 
Combustion Engineering Owners Group ``Joint Applications Report for 
Low Pressure Safety Injection System AOT Extension,'' May 1995, as 
supplemented by updated PVNGS data provided in the attachment to 
this enclosure.
    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 create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed amendment will extend the Completion Time for one 
inoperable low pressure safety injection (LPSI) subsystem in 
Technical Specification (TS) 3.5.3, Emergency Core Cooling Systems 
(ECCE)[S]--Operating, from 72 hours to 7 days. The proposed change 
does not change the design, configuration, or method of operation of 
the plant. Therefore, this change does not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety. The proposed amendment will extend the 
Completion Time for one inoperable low pressure safety injection 
(LPSI) subsystem in Technical Specification (TS) 3.5.3, Emergency 
Core Cooling Systems (ECCE)[S]--Operating, from 72 hours to 7 days. 
The proposed change does not affect the limiting conditions for 
operation or their bases used in the deterministic analyses to 
establish the margin of safety. PSA evaluations were used to 
evaluate these changes. These evaluations demonstrate that the 
changes will be risk neutral or risk beneficial for PVNGS. These 
evaluations are detailed in CE NPSD-995, as supplemented by updated 
data provided in the attachment to this enclosure.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Project Director: William H. Bateman.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: January 22, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) Sections 3.7.D.1.g, 6.2.2.h and 
6.3.1. Specifically, (1) Section 3.7.D.1.g would be revised to correct 
an editorial error; (2) Section 6.2.2.h would be revised to change the 
senior reactor operator license requirement for the Operations Manager; 
and (3) Section 6.3.1 would modify the qualification requirement for 
the Operations Manager.
    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 Section 3.7.D.1.g] is administrative in 
nature. It involves making an editorial change to provide the 
correct functional description of the breakers. This change does not 
affect possible initiating events for accidents previously evaluated 
or alter the configurations or operation of the facility. The 
Limiting Safety Systems Settings and Safety Limits specified in the 
current Technical Specifications

[[Page 17024]]

remain unchanged. Therefore, the proposed change to the subject 
Technical Specification would not increase the probability or 
consequences of an accident previously evaluated.
    The proposed change [to Section 6.2.2.h] is administrative in 
nature. The individual who provides the day to day direction of the 
activities of the operating shift will still possess an SRO [Senior 
Reactor Operator] license and this proposed change is consistent 
with the statement in NUREG-1431, Section 5.2.2.f. This change does 
not affect possible initiating events for accidents previously 
evaluated or alter the configuration or operation of the facility. 
The Limiting Safety Systems Settings and Safety Limits specified in 
the current Technical Specifications remain unchanged. Therefore, 
the proposed change to the subject Technical Specification would not 
increase the probability or consequences of an accident previously 
evaluated.
    The proposed change [to Section 6.3.1] is administrative in 
nature. The individual who provides the day to day direction of the 
activities of the operating shift will still possess an SRO license 
and this proposed change is consistent with the statement in NUREG-
1431, Section 5.2.2.f. This change does not affect possible 
initiating events for accidents previously evaluated or alter the 
configuration or operation of the facility. The Limiting Safety 
Systems Settings and Safety Limits specified in the current 
Technical Specifications remain unchanged. Therefore, the proposed 
change to the subject Technical Specification would not increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    As stated above, the proposed change [to Section 3.7.D.1.g] is 
administrative in nature. The safety analysis of the facility 
remains complete and accurate. There are no physical changes to the 
facility and the plant conditions for which the design basis 
accidents have been evaluated are still valid. The operating 
procedures and emergency procedures are unaffected. Consequently, no 
new failure modes are introduced as a result of the proposed change. 
Therefore, the proposed change will not initiate any new or 
different kind of accident.
    The proposed change [to Section 6.2.2.h] is administrative in 
nature. The safety analysis of the facility remains complete and 
accurate. There are no physical changes to the facility and the 
plant conditions for which the design basis accidents have been 
evaluated are still valid. The operating procedures and emergency 
procedures are unaffected. Consequently, no new failure modes are 
introduced as a result of the proposed changes. Therefore, the 
proposed change will not initiate any new or different kind of 
accident.
    The proposed change [to Section 6.3.1] is administrative in 
nature. The safety analysis of the facility remains complete and 
accurate. There are no physical changes to the facility and the 
plant conditions for which the design basis accidents have been 
evaluated are still valid. The operating procedures and emergency 
procedures are unaffected. Consequently, no new failure modes are 
introduced as a result of the proposed changes. Therefore, the 
proposed change will not initiate any new or different kind of 
accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change [to Section 3.7.D.1.g] is administrative in 
nature. Since there are no changes to the operation of the facility 
or physical design the Updated Final Safety Analysis Report (UFSAR) 
design basis, accident assumptions, or Technical Specification Bases 
are not affected. Therefore, the proposed changes will not result in 
a reduction in the margin of safety.
    The proposed change [to Section 6.2.2.h] is administrative in 
nature. Since there are no changes to the operation of the facility 
or physical design the Updated Final Safety Analysis Report (UFSAR) 
design basis, accident assumptions, or Technical Specification Bases 
are not affected. Therefore, the proposed changes will not result in 
a reduction in the margin of safety.
    The proposed change [to Section 6.3.1] is administrative in 
nature. Since there are no changes to the operation of the facility 
or physical design the Updated Final Safety Analysis Report (UFSAR) 
design basis, accident assumptions, or Technical Specification Bases 
are not affected. Therefore, the proposed changes will not result in 
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 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: S. Singh Bajwa, Director.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: January 22, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) Section 4.3. Specifically, the 
revision would permit the reactor coolant system (RCS) leak test to be 
performed at normal operating pressure after it has been closed 
following normal opening in lieu of a hydrostatic test being performed 
at 2335 psig.
    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 license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The change proposes a system leakage test for 
the RCS that is comparable to the hydrostatic test that it replaces, 
as acknowledged by the NRC approval of ASME Code Case N-498, 
``Alternative Rules for 10-Year Hydrostatic Pressure Testing for 
Class 1 and 2 Systems Section XI, Division 1,'' and the ASME 
[American Society for Mechanical Engineers] Boiler and Pressure 
Vessel Code, Section XI. [. . .] The proposed change to substitute a 
system leak test at normal operating pressure in lieu of the 
hydrostatic test at 2335 psig will minimize challenge to plant 
safety and demonstrate leak tightness of the RCS. Therefore, the 
proposed change would 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 license amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed changes do not involve the addition of any 
new or different type of equipment, nor do they involve the 
operation of equipment required for safe operation of the facility 
in a manner different from those addressed in the Updated Final 
Safety Analysis Report. [. . .] Based on industry experience, it is 
expected that any leaks would be discovered by the leak test at 
normal operating pressure.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed license amendment does not involve a significant 
reduction in a margin of safety. The proposed changes do not 
adversely affect performance of any safety related system or 
component, instrument operation, or safety system setpoints and do 
not result in increased severity of any of the accidents considered 
in the safety analysis. Although the current basis states that if 
the system does not leak at 2335 psig (operating pressure + 100 
psig) it will be leak tight during normal operation, industry 
experience demonstrates that leaks are not discovered as a result of 
hydrostatic test pressure propagating a preexisting flaw through 
wall. In most cases, leaks are discovered when the system is at 
normal operating pressure. Also, testing will continue to be 
performed as required by the ASME Boiler and Pressure Vessel Code 
Section XI.

    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.

[[Page 17025]]

Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: March 8, 1999.
    Description of amendment request: The proposed amendments would 
delete certain requirements from Technical Specification (TS) Section 
6.0 ``Administrative Controls'' that are adequately controlled by 
existing regulations, other than 10 CFR 50.36 and the TS. The 
amendments also relocate selected requirements from TS Section 6.0 to 
the licensee's controlled documents such as the Turkey Point Units 3 
and 4 Updated Final Safety Analysis Report (UFSAR). The amendments also 
clarify certain provisions of TS Section 6.0. The proposed changes are 
to relocate, revise, delete, or clarify the following provisions of the 
TS:

------------------------------------------------------------------------
  Existing TS section            Subject              Proposed change
------------------------------------------------------------------------
6.2.2.f................  Administrative Controls  Partly delete, partly
                          on Working Hours of      relocate within TS.
                          Plant Staff.
Table 6.2-1............  Minimum Shift Crew       Clarify.
                          Composition.
6.2.3..................  Shift Technical Advisor  Clarify.
6.4....................  Training...............  Delete.
6.5....................  Review and Audit.......  Relocate to UFSAR.
6.6....................  Reportable Event Action  Partly delete, partly
                                                   relocate to UFSAR.
6.8.2..................  Review and Approval of   Relocate to UFSAR.
                          Procedures.
6.8.3..................  Temporary Changes to     Relocate to UFSAR.
                          Procedures.
6.8.4.b................  In-Plant Radiation       Relocate to UFSAR.
                          Monitoring.
6.8.4.g................  Radiological             Relocate to UFSAR.
                          Environmental
                          Monitoring Program.
6.10...................  Record Retention.......  Relocate to UFSAR.
6.11...................  Radiation Protection     Relocate to UFSAR.
                          Program.
6.12...................  High Radiation Area....  Clarify.
6.13...................  Process Control Program  Relocate to UFSAR.
                          (PCP).
6.14...................  Offsite Dose             Revise to reflect
                          Calculation Manual       changes to 6.5 &
                          (ODCM).                  6.10.
------------------------------------------------------------------------

    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. Operation of the plant in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed changes are administrative in nature. These 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because they do not affect assumptions contained in plant safety 
analyses, the physical design and/or operation of the plant, nor do 
they affect Technical Specifications that preserve safety analysis 
assumptions. None of the proposed changes involve a physical 
modification to the plant, a new mode of operation or a change to 
the UFSAR transient analyses. No Limiting Condition for Operation, 
ACTION statement or Surveillance Requirement is affected by any of 
the proposed changes. Also, these proposed changes, in themselves, 
do not reduce the level of qualification or training such that 
personnel requirements would be decreased. Further, the Proposed 
changes do not alter the design, function, or operation of any plant 
component. Therefore, the proposed changes do not affect the 
probability or consequences of accidents previously evaluated.
    2. Operation of the plant in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The changes being proposed are administrative in nature and do not 
affect assumptions contained in plant safety analyses, the physical 
design and/or modes of plant operation defined in the plant 
operating license, or Technical Specifications that preserve safety 
analysis assumptions. The proposed changes do not introduce a new 
mode of plant operation or surveillance requirement, nor involve a 
physical modification to the plant. The proposed changes are 
administrative in nature. The changes propose to revise, delete, or 
relocate the stated administrative control provisions from the TS to 
the UFSAR whereby adequate control of information is maintained. 
Furthermore, the proposed changes do not alter the design, function, 
or operation of any plant components. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of the plant in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety because they are administrative in nature. The 
operating limits and functional capabilities of the affected 
systems, structures, and components are unchanged by the proposed 
amendments. None of the proposed changes involve a physical 
modification to the plant, a new mode of operation or a change to 
the UFSAR transient analyses. No Limiting Condition for Operation, 
ACTION statement, or Surveillance Requirement is affected. 
Additionally, the proposed changes do not alter the scope of 
equipment currently required to be OPERABLE or subject to 
surveillance testing, nor does the proposed change affect any 
instrument setpoints or equipment safety functions. 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 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: Florida International 
University, University Park, Miami, Florida 33199.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Cecil O. Thomas.

[[Page 17026]]

GPU Nuclear, Inc. etal., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: December 23, 1998.
    Description of amendment request: The proposed Technical 
Specification (TS) change request will change the surveillance 
frequency for verifying the operability of motor-operated isolation 
valves and condensate makeup valves in the Isolation Condenser TS 
4.8.A.1 and Bases page from once per month to once per 3 months.
    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 TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed surveillance interval change does not alter the 
actual surveillance requirements, nor does it alter the limits and 
restrictions on plant operations. The reliability of systems and 
components relied upon to prevent or mitigate the consequences of 
accidents previously evaluated is not degraded by the proposed 
change to the surveillance interval. Assurance of system and 
equipment availability is maintained. The proposed change does not 
alter any system or equipment configuration.
    Based on the above, the proposed change does not significantly 
increase the probability or consequences of a[n] accident previously 
evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed surveillance interval change does not alter the 
actual surveillance requirements, nor does it alter the limits and 
restrictions on plant operations. Assurance of system and equipment 
availability is maintained. The proposed change does not alter any 
system or equipment configuration nor does it introduce any new 
mechanisms which could contribute to the creation of a new or 
different kind of accident than previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed change extends the surveillance interval for 
verifying the operability of Isolation Condenser motor-operated 
isolation valves and condensate makeup valves from once per month to 
once per three months. The proposed change does not alter the actual 
surveillance requirements, the limits and restrictions on plant 
operations nor the design, function or manner of operation of any 
structures, systems or components. System availability and 
reliability are maintained. Accordingly, the proposed TS 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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Elinor G. Adensam.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of amendment request: February 12, 1999.
    Description of amendment request: The proposed Technical 
Specification (TS) change will delete the organizational chart and the 
related organizational references from the Appendix B Environmental TS 
and revise the appearance and format of the Environmental TS.
    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 because deletion of the organization charts and other 
organizational references in the [Environmental Technical 
Specifications] ETS does not affect plant operation. GPU Nuclear 
will continue to inform the NRC of organizational changes through 
other required controls.
    2. The proposed change does not create the possibility of a new 
or different type of accident than previously evaluated because the 
proposed change is administrative in nature, and no physical 
alteration of plant configuration, changes to setpoints or operating 
parameters are proposed.
    3. The proposed change does not involve a significant reduction 
in the margin of safety because it does not alter the design, 
function or manner of operation of any structures, systems or 
components. Organizational structure or its representation does not 
directly impact 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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Elinor G. Adensam.

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

    Date of amendment request: February 18, 1999.
    Description of amendment request: The proposed amendment would 
revise Duane Arnold Energy Center (DAEC) Technical Specification (TS) 
Table 3.3.6.1-1, ``Primary Containment Isolation Instrumentation,'' by 
deleting the manual initiation function of the high pressure coolant 
injection (HPCI) system and reactor core isolation cooling (RCIC) 
system isolation. A related condition as well as corresponding 
surveillance requirements and bases would also be deleted. Thus, the 
change would (1) revise Table 3.3.6.1-1 by removing items 3j. and 4.j.; 
(2) revise Note 2 to Surveillances to Licensing Condition for Operation 
(LCO) 3.3.6.1 by deleting information regarding items 3 j. and 4.j.; 
and (3) revise LCO 3.3.6.1 by removing Condition G and Surveillance 
Requirement 3.3.6.1.10.
    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:

    After reviewing this proposed amendment, we [the licensee] have 
concluded:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The Manual Initiation Function for HPCI and 
RCIC Isolation is not considered to be an initiator for any accident 
previously evaluated in the UFSAR. Therefore, this change does not 
involve a significant increase in the probability of any previously 
evaluated accidents. The Manual Initiation push button channels 
introduce signals into HPCI and RCIC System isolation logics that 
are redundant to the automatic protective instrumentation and 
provide manual isolation capability only if a system initiation 
signal is present. Technical Specification Section 3.3.6.1 Condition 
G requires isolation of the System flowpath, which renders the 
System inoperable and reduces the availability of the System due to 
the failure of a manually initiated isolation, an isolation which is 
not assumed in any transient or accident analysis in the UFSAR. 
Removal of the Manual Initiation Function

[[Page 17027]]

for HPCI and RCIC from the Primary Containment Isolation 
Instrumentation Section of Technical Specifications does not affect 
the automatic protective instrumentation and the automatic isolation 
capability. Therefore, this change does not significantly increase 
the consequences of a previously analyzed accident.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed change introduces no new mode of plant 
operation and does not involve physical modification to the plant. 
Therefore, it does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety. The proposed change deletes the 
Manual Initiation Function from Technical Specifications, but no 
significant reduction in a margin of safety is involved. Technical 
Specification Section 3.3.6.1 Condition G requires isolation of the 
System flowpath, which renders the System inoperable and reduces the 
availability of the System due to the failure of a manually 
initiated isolation, an isolation that is not assumed in any 
transient or accident analysis in the UFSAR. Removal of the Manual 
Initiation Function for HPCI and RCIC from the Primary Containment 
Isolation Instrumentation Section of Technical Specifications does 
not affect the automatic protective instrumentation and the 
isolation capability. This change is acceptable based on the fact 
that the Manual Initiation Function is not assumed in any accident 
or transient analysis in the UFSAR.
    Based upon the above, we [licensee] have determined that the 
proposed amendment will not involve a significant hazards 
consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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.
    NRC Project Director: T.J. Kim, Acting.

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

    Date of amendment request: March 1, 1999.
    Description of amendment request: The proposed amendment would 
change the Cooper Nuclear Station (CNS) Technical Specifications (TSs) 
to revise the calibration frequency of the reactor recirculation flow 
transmitters from once every 184 days to once every 18 months. This 
calibration is required as part of TS Surveillance Requirement (SR) 
3.3.1.1.10.
    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 amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Changing the calibration frequency of the recirculation 
loop flow transmitters from 184 days to 18 months may increase the 
amount of drift experienced by the transmitters. However, CNS 
calculation (NEDC 98-024 [forwarded by letter dated March 10, 1999]) 
takes into account the 18 month calibration intervals. This 
calculation, performed in accordance with the General Electric (GE) 
setpoint methodology for CNS, demonstrates that the expected drift 
is not significant, and is consistent with past operating 
experience. Changing the calibration frequency of the flow 
transmitters does not change any of the precursors assumed in the 
accident analysis. Therefore, changing the calibration frequency for 
flow transmitters from 184 days to 18 months does not involve a 
significant increase in the probability of an accident previously 
evaluated in the USAR [Updated Safety Analysis Report].
    The proposed change will not create the possibility of a new or 
different kind of accident than evaluated in the USAR. The proposed 
change does not result in any physical change to plant structures, 
systems, or components. The proposed change does not alter the form, 
fit, or function of any equipment or components credited in the 
accident analyses described in the USAR. Therefore, changing the 
test frequency does not create the possibility of a new or different 
kind of accident.
    The proposed change will not involve a significant reduction in 
a margin of safety. This conclusion is based on the fact that the 
proposed change is consistent with the drift assumptions used in CNS 
approved calculation (NEDC 98-024). The calibration frequency of 18 
months is consistent with the operating practices prior to 
conversion to Improved Technical Specifications, and is consistent 
with past operating practice at CNS. Therefore, changing the 
calibration frequency from 184 days to 18 months 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, Nebraska 68305.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, Nebraska 68602-0499.
    NRC Project Director: George Dick, Acting.

Northeast Nuclear Energy Company (NNECO), et al., Docket Nos. 50-336 
and 50-423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New 
London County, Connecticut

    Date of amendment request: March 5, 1999.
    Description of amendment request: The proposed amendment would 
relocate certain Technical Specification (TS) Section 6.0 
administrative controls to the NRC-approved Northeast Utilities Quality 
Assurance Program (NUQAP) Topical Report. Specifically, Sections 6.2.3 
(Unit 3 only), 6.5, 6.6 (partial), 6.7 (partial), and 6.10. The 
proposed amendment would also delete parts of Section 6.6 and 6.7 
because their requirements are duplicated in existing regulations or 
elsewhere in the TS. In addition, the proposed amendment would modify 
the table of contents and other TS sections to incorporate the 
aforementioned changes (e.g., correct references).
    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:

    In accordance with 10 CFR 50.92, NNECO has reviewed the attached 
proposed changes and has concluded that they do not involve a 
Significant Hazards Consideration (SHC). The basis for this 
conclusion is that the three criteria of 10 CFR 50.92 are not 
compromised. The proposed changes are not a SHC because the proposed 
change will not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    No design basis accidents are affected by these proposed 
changes. The proposed changes relocate portions of the Technical 
Specifications to the NUQAP Topical Report or remove duplicate 
sections and are being proposed to eliminate the need for a T.S. 
change each time there is a related change in the administrative 
controls for the site.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no changes in the way the plant is operated due to 
these revisions. The potential for an unanalyzed accident is not 
created. There is no impact on plant response, and no new failure 
modes are introduced. The proposed deletions and

[[Page 17028]]

editorial changes have no impact on safety limits or design basis 
accidents, and have no potential to create a new or unanalyzed 
event.
    3. Involve a significant reduction in a margin of safety.
    These changes do not directly affect any protective boundaries 
nor do they impact the safety limits for the protective boundaries. 
These proposed changes relocate portions of the administrative 
controls to the NUQAP Topical Report or are editorial in nature. 
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: 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 Project Director: Elinor G. Adensam.

PECO Energy Company, Docket No. 50-353, Limerick Generating Station, 
Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: March 11, 1999.
    Description of amendment request: The proposed revision to the 
Technical Specifications (TSs) involves a change to TS Section 2.1 and 
its associated TS Bases to revise the minimum critical power ratio 
(MCPR) Safety Limits for Cycle 6.
    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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The revised MCPR Safety Limits for LGS Unit 2 Technical 
Specifications, and their use to determine cycle-specific thermal 
limits, have been calculated using NRC-approved methods (i. e, 
GESTAR-II, Rev. 13) and are based on LGS, Unit 2, Cycle 6 specific 
inputs. The use of these methods assures that the SLMCPR [safety 
limit minimum critical power ratio] value is within the existing 
design and licensing basis, and cannot increase the probability or 
severity of an accident.
    The basis for the MCPR Safety Limit calculation is to ensure 
that greater than 99.9 percent of all fuel rods in the core avoid 
transition boiling if the limit is not violated. The MCPR Safety 
Limit preserves the existing margin to transition boiling and fuel 
damage in the event of a postulated accident. The probability of 
fuel damage is not increased.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The MCPR Safety Limit is a Technical Specification numerical 
value designed to ensure that fuel damage from transition boiling 
does not occur as a result of the limiting postulated accident. The 
MCPR Safety Limit is not an accident initiator; therefore, it cannot 
create the possibility of any new type of accident. The new MCPR 
Safety Limits are calculated using NRC-approved methods (i.e., 
GESTAR-II, Rev. 13) and are based on LGS, Unit 2, Cycle 6 specific 
inputs.
    Therefore, the proposed TS changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in the margin of safety.
    The margin of safety as defined in the TS Bases will remain the 
same. The new MCPR Safety Limits are calculated using NRC-approved 
methods (i.e., GESTAR-II, Rev. 13), which are in accordance with the 
current fuel design and licensing criteria, and are based on LGS, 
Unit 2, Cycle 6 specific inputs. The MCPR Safety Limit remains high 
enough to ensure that greater than 99.9 percent of all fuel rods in 
the core will avoid transition boiling if the limit is not violated, 
thereby preserving the fuel cladding integrity.

    Therefore, the proposed TS changes do not involve a reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Project Director: Elinor G. Adensam.

PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit 
1, Luzerne County, Pennsylvania

    Date of amendment request: March 12, 1999.
    Description of amendment request: The amendment would modify the 
Susquehanna Steam Electric Station, Unit 1, Technical Specifications 
Table 3.3.5.1-1 ``Emergency Core Cooling System Instrumentation.'' The 
change updates the allowable values for both the Core Spray (CS) and 
Low Pressure Coolant Injection System (LPCI) ``Reactor Steam Dome 
Pressure--Low'' functions for initiation and injection permissive. 
Specifically, the allowable values are being changed from a specified 
minimum pressure to a specified allowable pressure band. This more 
restrictive allowable value range will prevent CS and LPCI system 
overpressurization while still permitting injection to prevent fuel 
clad temperature limits from being exceeded.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposal does not involve an increase in the probability or 
consequences of an accident previously evaluated. The proposed 
amendment changes the ``Reactor Steam Dome Pressure-Low'' Allowable 
Values so to provide further assurance that the Core Spray and RHR 
systems will perform their LOCA [Loss-of-coolant accident] design 
basis function.
    The functional design basis of the Core Spray and LPCI is to 
inject water into the reactor vessel to cool the core during a LOCA 
by opening the Core Spray and LPCI injection valves when reactor 
pressure drops below the reactor vessel low pressure permissive. The 
upper analytical limit for the permissive is the Core Spray and LPCI 
systems' maximum design pressure, and the lower analytical limit is 
the lowest pressure which allows injection to prevent exceeding the 
fuel cladding temperature limit. The new allowable values were 
selected to lie within the upper and lower limits to ensure there 
will be no change in the required logic or functions of the Core 
Spray and LPCI systems. These new values do not affect the LOCA or 
its ``limiting fault'' frequency of occurrence and do not introduce 
any new accidents or malfunctions of equipment important to safety. 
Since they do not affect the LOCA, they do not change the 
probability of occurrence of the LOCA. The new allowable values do 
not change the logic or function of the reactor vessel low pressure 
permissive. These new values simply provide the basis for which the 
associated pressure

[[Page 17029]]

instruments are to be set to ensure proper operation of Core Spray 
and LPCI within the design pressures as described above. Therefore, 
the change in allowable values does not increase the probability of 
occurrence or the consequences of an accident or malfunction of 
equipment important to safety.
    Based upon the analysis presented above, PP&L [PP&L, Inc.] 
concludes that the proposed action does not involve an 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.
    This proposal does not create the probability of a new or 
different type of accident from any accident previously evaluated. 
The new allowable values do not change any plant systems, 
structures, or components, nor do they change any existing or create 
any new Core Spray and LPCI logic or functions. The new allowable 
values were selected to ensure the required operation of the Core 
Spray and LPCI systems within the design pressures described above.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The change does not involve a reduction in the margin of safety. 
Technical Specification Bases Section B3.3.5.1 9 (ECCS 
Instrumentation) identifies that the low reactor steam dome pressure 
signals are used as permissives for operation of the low pressure 
ECCS subsystems. The new allowable values were selected so to not 
impact the logic, redundancy, operability or surveillance 
requirements for these subsystems. The new allowable values maintain 
the margin requirements that the Core Spray and LPCI system 
pressures such that they do not exceed their system maximum design 
pressures and that system pressures are high enough to ensure that 
the ECCS injection prevents the fuel peak cladding temperature from 
exceeding the limits of 10CFR50.46.

    The margin of safety is unaffected by the 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 proposed to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PP&L, Inc., 2 North Ninth St., GENTW3, Allentown, PA 18101-
1179.
    NRC Project Director: Elinor G. Adensam.

PP&L, Inc., Docket No. 50-387, Susquehanna Steam Electric Station, Unit 
1, Luzerne County, Pennsylvania

    Date of amendment request: March 12, 1999.
    Description of amendment request: This proposed amendment would 
revise the minimum critical power ratio safety limit in Technical 
Specification (TS) Section 2.1.1.2. Also, the proposed amendment would 
modify the references in TS Section 5.6.5 in order to include only 
those references that directly support the generation of the Core 
Operating Limit and to remove the reference for the Lead Use 
Assemblies, which will be discharged during the next Unit 1 refueling 
outage.
    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 applicable sections of the FSAR are Chapters 4.4 and 15. 
FSAR Chapter 4.4 describes the MCPR Safety Limit, and Chapter 15 
describes the transient and accident analyses. The reference to be 
added to Section 5.6.5 of the Unit 1 Technical Specifications 
describes a NRC approved critical power correlation for 
ATRIUMTM-10 fuel. This correlation is appropriate for use 
in conservative methodologies for generating MCPR Safety Limits and 
MCPR Operating Limits to assure safe operation of Unit 1 with 
ATRIUMTM-10 fuel. A discussion of the impact of the 
proposed Technical Specification change is provided below.
    The proposed change in critical power correlation does not 
physically affect the plant or its systems. Thus, it does not 
increase the probability of an accident previously evaluated.
    A Unit 1 Cycle 12 MCPR Safety Limit analysis was performed for 
PP&L by SPC. This analysis used NRC approved methods described in 
ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2. These methods 
will be used each cycle to calculate the Unit 1 Safety Limits. For 
Unit 1 Cycle 12, the critical power performance of the 9 x 9-2 and 
ATRIUMTM-10 fuel was determined using the NRC approved 
ANFB and ANFB-10 correlations, respectively. The SAFETY LIMIT MCPR 
calculations statistically combine uncertainties on feedwater flow, 
feedwater temperature, core flow, core pressure, core power 
distribution, and uncertainties in the Critical Power Correlation. 
The SPC analysis used cycle specific power distributions and 
calculated MCPR values such that at least 99.9% of the fuel rods are 
expected to avoid boiling transition during normal operation or 
anticipated operational occurrences. The resulting two-loop and 
single-loop MCPR Safety Limits are included in the proposed 
Technical Specification change. Thus, the cladding integrity and its 
ability to contain fission products are not adversely affected.
    Analyses of the Single Loop Pump Seizure accident with the NRC 
approved ANFB-10 correlation for ATRIUMTM-10 fuel 
(Reference 1) will be performed to demonstrate that the NRC 
acceptance criterion (i.e., small fraction of 10CFR100 dose limits) 
is met. Analyses will also be performed to validate the conclusion 
that two-loop transients are more severe than those events analyzed 
in single-loop operation.
    Changes to Section 2.1.1.2 reflect the change from a flow 
dependent MCPR Safety Limit to a single value MCPR Safety Limit for 
two-loop operation and single-loop operation.
    Changes to Reference 5.6.5 delete the methodology used for 
critical power analyses for ATRIUMTM-10 fuel and add the 
NRC approved ANFB-10 methodology to the list of approved 
methodologies. Other changes in Reference 5.6.5 are administrative 
in nature because they delete references not directly related to the 
generation of Core Operating Limits. No new analysis approaches are 
used due to these changes.
    Changes to BASES Sections 2.1.1 and 3.2.2 reflect the inclusion 
of the ANFB-10 critical power correlation. The range of the 
applicability of the ANFB-10 is valid for pressures > 571 psia and 
bundle mass fluxes > 0.115  x  10\6\ lb/hr-ft \2\. These values 
assure that a valid CPR calculation will result at or above 25% of 
rated core thermal power, that is, reactor steam dome pressure 
 785 psig and core flow  10 Mlbm/hr.
    Changes to BASES Sections 3.2.1, 3.2.2, 3.2.3, and 3.2.4 reflect 
the removal of Reference 7 for the ABB LUAs, since the four LUAs 
will be discharged from Unit 1 during the Unit 1 11th Refueling and 
Inspection Outage.
    The consequences of transients and accidents will remain within 
the criteria approved by the NRC. The methodology used to perform 
the analyses has been previously approved by the NRC. Thus, analysis 
results using the new methodology will continue to provide assurance 
that the reactor will perform its design safety function during 
normal operation and design basis events. Therefore, the proposed 
action does not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the Unit 1 Technical Specifications 
(MCPR Safety Limits, removal of methodology references not directly 
supporting the generation of Core Operating Limits, removal of the 
two references describing previously approved methodology for 
applying ANFB to ATRIUMTM-10 fuel, removal of the ABB LUA 
reference, and inclusion of the ANFB-10 correlation reference) do 
not require any physical plant modifications, physically affect any 
plant components, or entail changes in plant operation. Removal of 
the Unit 1 Cycle 11 footnote allows Unit 1 Cycle 12 and future cycle 
operation with NRC

[[Page 17030]]

approved methodology. Thus, the proposed change does not create the 
possibility of a previously unevaluated operator error or a new 
single failure. The consequences of transients and accidents will 
remain within the criteria approved by the NRC. Therefore, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The applicable Technical Specification Sections include 2.1.1.2 
and 5.6.5.
    The changes to the Unit 1 Technical Specifications discussed in 
Item 1 above do not require any physical plant modifications, 
physically affect any plant components, or entail changes in plant 
operation. Therefore, the proposed change will not jeopardize or 
degrade the function or operation of any plant system or component 
governed by Technical Specifications. The consequences of transients 
and accidents will remain within the criteria approved by the NRC. 
The proposed MCPR Safety Limits and use of the ANFB-10 critical 
power correlation described in the reference added to Section 5.6.5 
do not involve a significant reduction in the margin of safety as 
currently defined in the Bases of the applicable Technical 
Specification sections.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Bryan A. Snapp, Esquire, PP&L, Inc., 2 North 
Ninth St., Allentown, PA 18101.
    NRC Project Director: Elinor G. Adensam.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: March 2, 1999.
    Description of amendment request: The proposed amendment would 
clarify the use of a ``check valve with flow through the valve 
secured'' as a means to isolate an affected containment penetration 
(i.e., a penetration with an inoperable penetration barrier) in 
Technical Specification 3.6.3 Action 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?
    The proposed change does not involve an increase in the 
probability or consequences of an accident previously evaluated. The 
proposed change does not involve any hardware changes. The proposed 
change will clarify Technical Specification 3.6.3 Action b to allow 
the use of a check valve with the flow through the valve secured as 
a means to isolate an inoperable containment penetration. This 
change is consistent with the changes identified in NUREG-1431, 
``Improved Standard Technical Specifications for Westinghouse 
Plants'', Specification 3.6.3 (Containment Isolation Valves), which 
identifies check valves with flow through the valve secured as a 
type of deactivated automatic valve, and with 10 CFR 50 Appendix A 
General Design Criteria 55 and 56, which include the use of check 
valves as ``automatic isolation valves''. The proposed change will 
not affect the containment isolation valve OPERABILITY requirements 
or associated isolation time limits established in the 
Specifications. Therefore the proposed change will not affect any 
safety margin or safety limit applicable to the facility. Therefore 
no increase in the probability or consequences of any accident 
previously evaluated will occur.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change will clarify Technical Specification 3.6.3 
Action b to allow the use of a check valve with the flow through the 
valve secured as a means to isolate an inoperable containment 
penetration. The proposed change will not involve any physical 
change to plant systems, structures, or components (SSC). This 
change is consistent with the changes identified in NUREG-1431, 
``Improved Standard Technical Specifications for Westinghouse 
Plants'', Specification 3.6.3 (Containment Isolation Valves), which 
identifies check valves with flow through the valve secured as a 
type of deactivated automatic valve, and with 10 CFR 50 Appendix A 
General Design Criteria 55 and 56, which include the use of check 
valves as ``automatic isolation valves''. The proposed change only 
provides clarification to the existing Specification 3.6.3, and will 
not affect the established containment isolation valve OPERABILITY 
requirements or associated isolation time limits. Since the proposed 
change does not impact operation of the facility as presently 
approved, no possibility exists for a new or different kind of 
accident from those previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will clarify Technical Specification 3.6.3 
Action b to allow the use of a check valve with the flow through the 
valve secured as a means to isolate an inoperable containment 
penetration. This change is consistent with the changes identified 
in NUREG-1431, ``Improved Standard Technical Specifications for 
Westinghouse Plants'', Specification 3.6.3 (Containment Isolation 
Valves), which identifies check valves with flow through the valve 
secured as a type of deactivated automatic valve, and with 10 CFR 50 
Appendix A General Design Criteria 55 and 56, which include the use 
of check valves as ``automatic isolation valves''. The proposed 
change only provides clarification to the existing Specification 
3.6.3, and will not affect the established containment isolation 
valve OPERABILITY requirements or associated isolation time limits. 
The proposed change does not involve a significant reduction in a 
margin of safety because the ability to isolate containment in the 
event of a release of radioactive material to the containment 
atmosphere or pressurization of the containment will be maintained. 
The margin of safety is defined by the established containment 
isolation valve OPERABILITY requirements and associated isolation 
time limits. The proposed change does not alter these operating 
restrictions and the margin of safety which assures the ability to 
isolate containment is not affected.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: George Dick, Acting.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: March 9, 1999.
    Description of amendment request: The amendment request proposes 
that reference to the Independent Safety Engineering Group be removed 
from Technical Specification requirements, with supporting changes to 
the Operations Quality Assurance Plan.
    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 does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment is a programmatic and administrative 
change that

[[Page 17031]]

does not physically alter safety-related systems, nor does it affect 
the way in which safety-related systems perform their functions. The 
functions assigned to the Independent Safety Engineering Group are 
addressed by other organizations. Because the design of the facility 
and system operating parameters are not being changed, the proposed 
amendment does not involve an increase in the probability or 
consequences of any accident previously evaluated.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment is a programmatic and administrative 
change that does not physically alter safety-related systems, nor 
does it affect the way in which safety-related systems perform their 
functions. The functions assigned to the Independent Safety 
Engineering Group are addressed by other organizations. Because the 
design of the facility and system operating parameters are not being 
changed, the proposed amendment does not create the possibility of a 
new or different kind of accident previously evaluated.
    The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed amendment is a programmatic and administrative 
change that provides assurance that plant operations continue to be 
conducted in a safe manner. The functions assigned to the 
Independent Safety Engineering Group are addressed by other 
organizations. As stated above the proposed amendment does not 
physically alter safety-related systems, nor does it affect the way 
in which safety-related systems perform their functions. Because the 
design of the facility and system operating parameters are not being 
changed, the proposed amendment 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: George F. Dick, Acting.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: March 15, 1999 (Supplement to October 
29, 1998).
    Description of amendment request: The proposed amendments were 
submitted by application dated October 29, 1998, to relocate Technical 
Specification (TS) 3/4.7.9 requirements for snubbers to the Technical 
Requirements Manual. The Commission issued a Notice of Consideration of 
Issuance of Amendments regarding its proposed no significant hazards 
consideration determination that was published in the Federal Register 
on December 16, 1998 (63 FR 69346).
    Subsequently, by letter dated March 15, 1999, supplemental 
information was submitted to include TS 6.10.3.l to be relocated to the 
Technical Requirements Manual. This information is being noticed to 
provide for public comment on the issue of no significant hazards 
consideration.
    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 supplement to the amendment request relocates the record 
keeping requirements of Technical Specification 6.10.3.l to the 
Technical Requirements Manual. The change does not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed) or make changes in the methods 
governing normal plant operation. The change will not impose 
different requirements, and adequate control of information will be 
maintained. This change will not alter assumptions made in the 
safety analysis and licensing basis.
    The Technical Requirements Manual is incorporated in the South 
Texas Project Updated Final Safety Analysis Report and will be 
maintained pursuant to 10 CFR 50.59. In addition, snubber 
operability is addressed in existing surveillance procedures that 
are also controlled by 10 CFR 50.59 and subject to the change 
control provisions imposed by plant administrative procedures, which 
endorse applicable regulations and standards.
    Therefore, the supplement to the proposed 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 supplement to the amendment request relocates the record 
keeping requirements of Technical Specification 6.10.3.l to the 
Technical Requirements Manual. The change does not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed) or make changes in the methods 
governing normal plant operation. The change will not impose 
different requirements, and adequate control of information will be 
maintained. This change will not alter assumptions made in the 
safety analysis and licensing basis.
    The Technical Requirements Manual is incorporated in the South 
Texas Project Updated Final Safety Analysis Report and will be 
maintained pursuant to 10 CFR 50.59. In addition, snubber 
operability is addressed in existing surveillance procedures that 
are also controlled by 10CFR50.59 and subject to the change control 
provisions imposed by plant administrative procedures, which endorse 
applicable regulations and standards.
    Therefore, the change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The supplement to the amendment request relocates the record 
keeping requirements of Technical Specification 6.10.3.l to the 
Technical Requirements Manual. The relocated requirements remain the 
same as the existing Technical Specifications. The change will not 
reduce a margin of safety because it has no impact on any safety 
analysis assumptions. Future changes to the relocated requirements 
will be evaluated per the requirements of 10CFR50.59.
    Therefore, the supplement will not result in 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: George F. Dick, Acting.

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

    Date of amendment request: February 15, 1999.
    Description of amendment request: The proposed amendment would 
revise requirements of Technical Specifications Section 6, 
``Administrative Controls,'' related to (1) plant manager's 
responsibilities, (2) plant staff titles and organization, (3) offsite 
and onsite review committee (4) reportable events, and (5) actions 
required in event of a safety limit violation.
    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

[[Page 17032]]

consideration, which is presented below:

    The proposed amendment will not change the intent of the TS or 
decrease WPSC's management support or involvement in activities at 
the Kewaunee Plant. Furthermore, it will not result in a decrease in 
the engineering or technical support supplied by the plant staff or 
the corporate support staff. The proposed changes are administrative 
in nature. They primarily involve the relocation of existing 
requirements to owner controlled documents; therefore, there are no 
significant hazards associated with this change. As an 
administrative change this will not result in a significant increase 
in the probability of occurrence or consequences of an accident. As 
an administrative change this will not create the possibility of a 
new or different kind of accident from any previously analyzed. This 
administrative change relocates existing requirements, and 
therefore, will not involve a significant decrease in the margin of 
safety.

    In addition, the staff analyzed the proposed changes in accordance 
with the provisions of 10 CFR 50.92. The proposed change will not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    The analyses for the previously evaluated accidents are presented 
in Chapter 14 of the Updated Safety Analysis Report. There are 19 
postulated accidents addressed therein. The proposed amendment would 
not affect the safety analysis assumptions or analytical models used 
for any of these analyses. Also, the calculated dose consequences for 
analyzed accidents would be unaffected. Therefore the proposed changes 
do not involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed accident does not involve any physical change to the 
design of the physicality, or operation of the facility outside the 
bounds of the existing analyses. Thus, there is no possibility of 
creating a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed changes do not involve any physical changes to any of 
the fission product barriers or to the design or operation of any 
safety systems. Also, no safety limits, limiting safety systems 
settings, limiting conditions for operation or testing requirements 
would be affected. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.
    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: Cynthia A. Carpenter.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power 
Station, Franklin County, Massachusetts

    Date of amendment request: March 17, 1999.
    Description of amendment request: Licensee submitted a License 
Amendment request to delete administrative Technical Specification (TS) 
requirements related to overtime restrictions. The licensee stated it 
will provide appropriate constraints on excessive overtime in its 
Administrative Procedures.
    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 changes are administrative in nature and simply 
eliminate outdated requirements from the YNPS Technical 
Specifications. As such the changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The administrative 
nature of the changes will not affect safety-related systems or 
components or their mode of operation and therefore, will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different accident from 
any previously evaluated. The proposed changes do not modify any 
plant systems or components and, therefore, do not create the 
possibility of a new or different accident from any previously 
evaluated.
    3. Involve a significant reduction in the margin of safety. The 
changes are administrative in nature involving the deletion of 
outdated requirements in the technical specifications; therefore, 
there will be no reduction in the margin of safety.

Based on the considerations noted above, it is concluded that the 
proposed changes will not endanger the public health and 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: Greenfield Community College, 
1 College Drive, Greenfield, Massachusetts 01301.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One 
International Place, Boston, Massachusetts 02110-2624.
    NRC Project Director: Seymour H. Weiss.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power 
Station, Franklin County, Massachusetts

    Date of amendment request: March 17, 1999.
    Description of amendment request: Licensee submitted a License 
Amendment request to transfer Technical Specification Sections 6.7--
Procedures and Programs and 6.9--Record Retention to the Yankee 
Decommissioning Quality Assurance Program (YDQAP).
    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 changes are administrative in nature. 
Administrative requirements in Sections 6.7 and 6.9 of the YNPS 
Technical Specifications are to be transferred to the YDQAP which is 
the current location of related administrative requirements. As such 
the changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The administrative 
nature of the changes will not affect safety-related systems or 
components or their mode of operation and therefore, will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different accident from 
any previously evaluated. The proposed changes do not modify any 
plant systems or components and, therefore, will not create the 
possibility of a new or different accident from any previously 
evaluated.
    3. Involve a significant reduction in the margin of safety. The 
changes are administrative in nature involving the relocation of 
administrative requirements from one licensing document to another 
licensing document currently containing related requirements; 
therefore, there will be no significant reduction in the margin of 
safety.

Based on the considerations noted above, it is concluded that the 
proposed changes will not endanger the public health and safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 17033]]

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: Greenfield Community College, 
1 College Drive, Greenfield, Massachusetts 01301.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One 
International Place, Boston, Massachusetts 02110-2624.
    NRC Project Director: Seymour H. Weiss.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power 
Station, Franklin County, Massachusetts

    Date of amendment request: March 17, 1999.
    Description of amendment request: Licensee submitted a License 
Amendment request to consolidate management positions and to transfer 
Technical Specification review and audit functions to the Yankee 
Decommissioning Quality Assurance Program (YDQAP).
    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 changes are administrative in nature and reflect a 
streamlining of the YAEC/YNPS management structure and procedures 
consistent with the on-going requirement to complete the remaining 
scope of YNPS decommissioning safely and efficiently. As such the 
changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. The administrative 
nature of the changes will not affect safety-related systems or 
components or their mode of operation and therefore, will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different accident from 
any previously evaluated. The proposed changes do not modify any 
plant systems or components and, therefore, will not create the 
possibility of a new or different accident from any previously 
evaluated.
    3. Involve a significant reduction in the margin of safety. 
Elimination of the Manager of Operations position and the Plant 
Superintendent position will not eliminate any of the 
responsibilities or functions currently assigned to these positions. 
These responsibilities or functions will be reassigned to an 
appropriately qualified YAEC/YNPS manager, i.e., the Decommissioning 
Manager. This change and replacement of the PORC and the NSARC 
review and audit functions with an independent safety review and an 
IRAC are consistent with the significant reduction in the scope and 
the complexity of activities at YNPS as the facility moves into the 
later stages of the decommissioning effort; therefore, there will be 
no significant reduction in the margin of safety.

Based on the considerations noted above, it is concluded that the 
proposed changes will not endanger the public health and 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: Greenfield Community College, 
1 College Drive, Greenfield, Massachusetts 01301.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One 
International Place, Boston, Massachusetts 02110-2624.
    NRC Project Director: Seymour H. Weiss.

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.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas

    Date of application for amendment: February 24, 1999.
    Brief description of amendment: The amendment would revise 
Technical Specification Table 3.3-1, ``Reactor Protective 
Instrumentation,'' Action 2, for Arkansas Nuclear One, Unit No. 2. The 
proposed change would add a footnote to Action 2 that would allow 
startup and operation with the functional units associated with the 
Channel ``D'' ex-core nuclear instrumentation to be maintained in the 
bypassed or tripped condition following the restart from Refueling 
Outage 2R13. This footnote is intended to support normal plant 
operations until such time that the Channel ``D'' ex-core detector 
assembly can be restored to an operable status.
    Date of publication of individual notice in Federal Register: March 
8, 1999 (64 FR 11067).
    Expiration date of individual notice: April 7, 1999.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

[[Page 17034]]

CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz 
Mill, Pennsylvania

    Date of application for amendment: September 28, 1998 supplemented 
on November 17, 1998.
    Brief description of amendment: This amendment changes the license 
to reflect the new legal name of the licensee for the Westinghouse Test 
Reactor to CBS Corporation.
    Date of issuance: March 25, 1999.
    Effective Date: March 25, 1999.
    Amendment No: 9.
    Facility License No. TR-2: This amendment changes the license.
    Date of initial notice in Federal Register: December 16, 1998, (63 
FR 69334).
    The Commission has issued a Safety Evaluation for this amendment 
dated March 25, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document: N/A.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

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

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

    Date of application for amendments: December 17, 1998.
    Brief description of amendments: The amendments revised the 
respective facility Technical Specifications (TS) by adding a new 
Limiting Condition for Operations that provided an administrative 
enhancement by allowing testing required to return equipment to service 
to be conducted under administrative controls.
    Date of issuance: March 16, 1999.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 172, 167; 184, 181; 132, 117.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29, DPR-30, 
NPF-11 and NPF-18.
    The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4153) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021; for LaSalle, the Jacobs Memorial Library, 815 North 
Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.

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: August 14, 1998, as 
supplemented on October 13 and December 23, 1998.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to reflect the use of Siemens Power 
Corporation (SPC) ATRIUM-9B fuel. Specifically, the amendments 
incorporate the following into the TSs: (1) new methodologies that will 
enhance operational flexibility and reduce the likelihood of future 
plant derates; (2) administrative changes that adopt Improved Standard 
Technical Specification (iSTS) language where appropriate; and (3) 
changes to the Minimum Critical Power Ratio.
    Date of issuance: March 16, 1999.
    Effective date: Immediately, to be implemented prior to startup of 
Cycle 9 for Unit 1 and prior to startup of Cycle 8 for Unit 2.
    Amendment Nos.: 131, 116.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59588). The December 23, 1998, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
March 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.

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 application for amendments: August 14, 1998, as 
supplemented by letters dated October 13, 1998, and December 23, 1998.
    Brief description of amendments: The amendments changed the Quad 
Cities Technical Specifications (TS) to reflect the use of Siemens 
Power Corporation (SPC) ATRIUM-9B fuel. Specifically, the amendments 
incorporate the following into the TS: (a) new methodologies that will 
enhance operational flexibility and reduce the likelihood of future 
plant derates; (b) administrative changes that eliminate the cycle-
specific implementation of ATRIUM-9B fuel and adopt Improved Standard 
Technical Specification language where appropriate; and (c) changes to 
the Minimum Critical Power Ratio (MCPR).
    The amendment for Unit 1 also reflects the removal of Unit 1 
specific pages incorporated into Unit 1 TS by Amendment No. 182 and are 
no longer applicable. The August 14, 1998, application superseded an 
August 29, 1997, application in its entirety (63 FR 2274).
    Date of issuance: March 17, 1999.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 185 & 182.
    Facility Operating License Nos. DPR-29, DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48258) and November 4, 1998 (63 FR 59588). The October 13, 1998, 
submittal changed a reference to a recently NRC-approved additive 
constant uncertainty (ACU) generic methodology for ATRIUM-9B fuel (ANF-
1125 (P)(A), supplement 1, Appendix E) from Appendix D which provided 
an interim value for ACU. This change was noticed on November 4, 1998 
(63 FR 48258). The December 23, 1998, submittal provided additional 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated March 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

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 application for amendments: November 30, 1998.
    Brief description of amendments: The amendments changed the 
technical specifications (TSs) by decreasing the Allowed Outage Time 
(AOT) from 67 days to 14 days for the Safe Shutdown Makeup Pump (SSMP).
    Date of issuance: March 26, 1999.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 186 & 183.

[[Page 17035]]

    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 13, 1999 (64 FR 
2246).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    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: September 30, 1998.
    Brief description of amendments: The amendments increase the 
maximum fuel rod internal pressure in the spent fuel pool from 1200 
pounds per square inch gauge (psig) to 1300 psig by changing the 
Updated Final Analysis Report (UFSAR) reference to the computer code 
used to determine the fuel rod internal pressure (TACO3 computer code 
would be added) in UFSAR Chapter 15. In addition, the amendments 
justify not increasing the overall effective decontamination factor for 
iodine as a consequence of a fuel handling accident and change the 
terminology used in the UFSAR from ``fuel assembly gap gas pressure'' 
to ``fuel rod internal pressure.''
    Date of Issuance: March 26, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-301; Unit 2-301; Unit 3-301.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments authorized change(s) to the FSAR.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59590).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit 
No. 2, Pope County, Arkansas.

    Date of application for amendment: February 25, 1999.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Table 3.3-1, ``Reactor Protective Instrumentation,'' 
Action 2, for Arkansas Nuclear One, Unit No. 2 (ANO-2). This change 
adds a footnote to Action 2 that allows startup and operation with the 
functional units associated with the Channel ``D'' ex-core nuclear 
instrumentation to be maintained in the bypassed or tripped condition 
following the restart from Refueling Outage 2R13. This footnote is 
intended to support normal plant operations until such time that the 
Channel ``D'' ex-core detector assembly can be restored to an operable 
status. This footnote will be in effect for a time period not to extend 
beyond Mid-Cycle Outage 2P99, which is the next planned entry into cold 
shutdown conditions for ANO-2. A Notice of Enforcement Discretion 
(NOED) related to TS Table 3.3-1, Action 2, was issued verbally on 
February 23, 1999. The NOED is documented in a letter dated February 
25, 1999.
    Date of issuance: March 23, 1999.
    Effective date: As of the date of issuance.
    Amendment No.: 202.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (64 FR 11067 dated March 8, 1999). The notice 
provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided for an opportunity to request a hearing by April 7, 1999, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final NSHC determination are contained in a 
Safety Evaluation dated March 23, 1999.
    Attorney for Licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington DC 20005-3502.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio.

    Date of application for amendment: November 2, 1995, and as 
supplemented by submittal dated January 7, 1999.
    Brief description of amendment: This amendment revises technical 
specification requirements for handling irradiated fuel in the Primary 
Containment and the Fuel Handling Building, and selected specifications 
associated with performing core alterations.
    Date of issuance: March 11, 1999.
    Effective date: March 11, 1999.
    Amendment No.: 102.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62497).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not change the scope of the original application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 11, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio.

    Date of application for amendment: August 27, 1996, as supplemented 
by submittals dated April 9, 1997, July 22, 1998, December 3, 1998, and 
January 18, 1999.
    Brief description of amendment: This amendment revised Technical 
Specification 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs),'' and 3.6.1.9, ``Main Steam Isolation Valve (MSIV) Leakage 
Control System (LCS).'' The amendment reflects implementation of the 
revised accident source term in NUREG-1465, ``Accident Source Terms for 
Light-Water Nuclear Power Plants'' and permits the licensee to 
eliminate the MSIV LCS and increase the allowable leak rates of the 
MSIVs.
    Date of issuance: March 26, 1999:
    Effective date: March 26, 1999.
    Amendment No.: 103.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53958).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.

[[Page 17036]]

    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio.

    Date of application for amendment: October 27, 1998.
    Brief description of amendment: This amendment revised the minimum 
critical power ratio (MCPR) safety limit contained in TS 2.1.1.2. In 
addition, the amendment removes a note to TS 2.1.1.2 and a footnote to 
TS 5.6.5.b that references MCPR safety limit values as cycle specific.
    Date of issuance: March 26, 1999:
    Effective date: March 26, 1999.
    Amendment No.: 104.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66603).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio.

    Date of application for amendment: September 9, 1998, as 
supplemented by submittals dated January 6, March 4, and March 18, 
1999.
    Brief description of amendment: This amendment revises the design 
and licensing basis of containment isolation valves in the feedwater 
system. The amendment revises (1) Surveillance Requirement 3.6.1.3.11 
of Technical Specification (TS) 3.6.1.3, ``Primary Containment 
Isolation Valves (PCIVs)'' to exclude the feedwater check valves from 
the hydrostatic test program, (2) TS 5.5.2, ``Primary Coolant Sources 
Outside Containment,'' to stipulate that water leakage past the 
feedwater motor-operated containment isolation valves and the reactor 
water cleanup system return to feedwater line is added to the program, 
and (3) TS 5.5.12, ``Primary Containment Leakage Rate Testing 
Program,'' to state that the feedwater check valves will be tested in 
accordance with the Inservice Testing Program (TS 5.5.6).
    Date of issuance: March 26, 1999.
    Effective date: March 26, 1999.
    Amendment No.: 105.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56262).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.

    Date of application for amendments: August 24, 1998.
    Brief description of amendments: These amendments change the St. 
Lucie Technical Specifications (TSs) by both removing obsolete license 
conditions and revising the TSs. The amendments change the TSs to 
modify the St. Lucie Unit 1 TSs to add components, not previously 
described in the TSs, to the list of components that comprise an 
operable control room emergency ventilation system, to modify the Unit 
1 and Unit 2 TSs surveillance requirements to clarify component 
operations, not previously described, that must be verified in response 
to a containment sump recirculation actuation signal, to delete from 
the facility operating license No. NPF-16 for Unit 2, license condition 
2.C.19 to reflect the completion of the Unit 1 spent fuel pool re-rack 
and delete license condition 2.I to reflect the resolution of 
litigation and to modify license condition 2.B.5 to restore the 
original syntax of the license condition and license condition 2.F to 
update the references to current license conditions.
    Date of Issuance: March 17, 1999.
    Effective Date: These amendments shall be implemented within 30 
days of receipt.
    Amendment Nos.: 160 and 99.
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the TSs.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50937).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey.

    Date of application for amendment: September 3, 1998.
    Brief description of amendment: The amendment revises Technical 
Specifications 3.4.A.10.e and 3.5.a.2.e to incorporate a Condensate 
Storage Tank water level of greater than 35 feet.
    Date of Issuance: March 17, 1999.
    Effective date: March 17, 1999, to be implemented within 30 days
    Amendment No.: 204.
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6698).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 17, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 
1, DeWitt County, Illinois.

    Date of application for amendment: January 20, 1999, as 
supplemented February 4, 8, and 25, and March 5, 1999.
    Brief description of amendment: The amendment changes the 
undervoltage relay setpoints.
    Date of issuance: March 26, 1999.
    Effective date: March 26, 1999.
    Amendment No.: 122.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1999 (64 FR 
4474).
    The four supplemental submittals provided additional information 
and did not change the requested amendment or affect the proposed no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.

[[Page 17037]]

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of application for amendment: April 13, 1998, as supplemented 
November 5, 1998.
    Brief description of amendment: The proposed amendment would revise 
the Appendix A Technical Specifications to base the Limiting Condition 
for Operation for the fuel storage pool water level on a revised 
analysis of the fuel handling accident and a new analysis for 
radiological shielding during movement of irradiated fuel.
    Date of issuance: March 16, 1999.
    Effective date: March 16, 1999 (and shall be implemented no later 
than 30 days).
    Amendment No.: 162.
    Facility Operating License No. DPR-36: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 20, 1998 (63 FR 
27763). The November 5, 1998, submittal provided additional clarifying 
information and did not change the initial proposed no significant 
hazards determination and did not expand the scope of the original 
application.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 16, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of application for amendment: December 30, 1998.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Tables 3.6.14-2 and 4.6.14-2 regarding the noble gas 
activity monitor channel operability requirement and daily sensor check 
surveillance requirement to be consistent with the conditions specified 
in TS 3.1.3.a for operability of the emergency cooling system. Also, 
this amendment corrects a clerical error in TS 4.6.15.d.
    Date of issuance: March 16, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 165.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6699).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 16, 1999.
    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.

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

    Date of application for amendment: November 19, 1998.
    Brief description of amendment: This amendment changes surveillance 
frequencies in Technical Specifications 4.8.4.4a and 4.8.4.5a to 
require testing of the Electrical Protection Assemblies once every 6 
months with the plant on-line rather than shut down.
    Date of issuance: March 18, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 86.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71970).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 18, 1999.
    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.

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

    Date of amendment request: May 20, 1998, as supplemented by letter 
dated January 28, 1999.
    Description of amendment request: Revise Technical Specifications 
Table 3.3-4 and associated bases to depict a change to the refueling 
water storage tank low-low level setpoint
    Date of issuance: March 12, 1999.
    Effective date: As of its date of issuance, to be implemented 
within 60 days.
    Amendment No.: 60.
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43205).
    The supplemental letter provided clarifying information and did not 
change the staff's proposed no significant hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

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

    Date of application for amendment: September 9, 1998, as 
supplemented February 19 and 26, 1999.
    Brief description of amendment: The amendment resolves several 
previously identified technical specifications (TSs) compliance issues. 
Specifically, the amendment: (1) changed TS definitions 1.24, ``Core 
Operating Limits Report,'' 1.27, ``Engineering Safety Feature Response 
Time,'' and 1.31, ``Radiological Effluent Monitoring and Offsite Dose 
Calculation Manual (REMODCM)''; (2) changed TS 3.0.2, ``Limiting 
Condition for Operation,'' by adding a new TS 3.0.6 to the Limiting 
Condition for Operation TS section; (3) changed TS 4.0.5, 
``Surveillance Requirements''; (4) changed the mode applicability of TS 
3.2.3, ``Total Unrodded Integrated Radial Peaking--FrT''; 
(5) changed TS 3.3.2.1, ``Engineered Safety Features Actuation System 
Instrumentation,'' by modifying TS Table 4.3-2 Table Notation (1) which 
it references; and (6) changed TS 3.4.1.1, ``Reactor Coolant System--
Reactor Coolant System Vents.'' The associated TS Bases sections were 
also changed.
    Date of issuance: March 11, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 230.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56251).
    The supplemental letters provided clarifying information that did 
not change the original proposed no significant hazards consideration 
determination or expand the scope of the original Federal Register 
notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 11, 1999.

[[Page 17038]]

    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: July 17, 1998, as supplemented 
November 10, 1998, and February 11, 1999.
    Brief description of amendment: The amendment revises certain 
diesel generator (DG) action statements and surveillance requirements 
to improve overall DG reliability and availability.
    Date of issuance: March 12, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 231.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43207).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: November 25, 1997, as 
supplemented September 25 and November 11, 1998, and January 28, 1999.
    Brief description of amendment: The amendment revises the Technical 
Specifications for the condensate storage tank (CST) low level suction 
transfer setpoint for the high pressure coolant injection (HPCI) and 
reactor core isolation cooling (RCIC) systems to allow removing one CST 
from service for maintenance.
    Date of issuance: March 19, 1999.
    Effective date: March 19, 1999, with full implementation within 30 
days.
    Amendment No.: 105.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1998 (63 
FR 69344) The November 25, 1997, letter and September 25 and November 
11, 1998, supplements were referenced in the original Federal Register 
notice. The January 28, 1999, supplement provided an updated Technical 
Specification page following the incorporation of Amendment 103, issued 
December 23, 1998. This information was within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards considerations determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 19, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: November 25, 1998.
    Brief description of amendments: The amendments revise Technical 
Specifications 3.2 and Table 3.5-2B to allow limited inoperability of 
boric acid storage tank level channels and transfer logic channels to 
provide for required testing and maintenance of the associated 
components.
    Date of issuance: March 17, 1999.
    Effective date: March 17, 1999, with full implementation within 30 
days
    Amendment Nos.: 143 and 134.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69345).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 19, 1999.
    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.

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 application for amendments: July 30, 1997, as supplemented 
by letter dated December 23, 1998.
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant (DCPP) 
Unit Nos. 1 and 2 by adding a Limiting Condition for Operation, trip 
setpoints, and surveillance requirements for a residual heat removal 
pump trip on refueling water storage tank level-low.
    Date of issuance: March 26, 1999.
    Effective date: March 26, 1999, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 1--130; Unit 2--128.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68312).
    The December 31, 1997 supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noted, and did not change the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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 application for amendments: August 26, 1997, as 
supplemented by letters dated October 14 and November 13, 1997, and 
January 29, 1998.
    Brief description of amendments: The amendments approve a 
modification to the Diablo Canyon Power Plant (DCPP), Unit Nos. 1 and 2 
auxiliary saltwater (ASW) system to bypass approximately 800 feet of 
Unit 1 and 200 feet of Unit 2 Class 1 ASW pipe, a portion of which is 
buried below sea level in the tidal zone outside the intake structure.
    Date of issuance: March 26, 1999.
    Effective date: March 26, 1999, and shall be implemented in the 
next

[[Page 17039]]

periodic update to the FSAR Update in accordance with 10 CFR 50.71(e).
    Amendment Nos.: Unit 1--131; Unit 2--129.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Final Safety Analysis Report Update.
    Date of initial notice in Federal Register: September 16, 1997 (62 
FR 48677).
    The October 14 and November 13, 1997, and January 29, 1998, 
supplemental letters provided additional clarifying information, did 
not expand the scope of the application as originally noticed, and did 
not change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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 application for amendments: September 25, 1996, as 
supplemented on October 29, 1997, March 16, 1998, and February 8, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications by revising the voltage and frequency 
acceptance criteria and the start-timing methodology for the emergency 
diesel generator surveillance testing.
    Date of issuance: March 23, 1999.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment Nos: 218 and 200.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 23, 1996 (61 FR 
5039).
    The October 29, 1997, March 16, 1998, and February 9, 1999, letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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 application for amendments:  September 17, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.8.2, ``Electrical Power Sources--Shutdown,'' for the 
AC distribution system and the 125-volt and 28-volt DC distribution 
systems. Specifically, the amendments change the Applicability and 
Action Statements, if less than the complement of equipment and buses 
are operable, to eliminate the need to establish containment integrity 
and to add the action to suspend core alterations, positive reactivity 
additions, and movement of irradiated fuel assemblies.
    Date of issuance: March 24, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 219 and 201.
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56257).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 24, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: April 6, 1995, as supplemented 
on August 21, 1995. (TS 95-19).
    Brief description of amendments: The amendments change the licenses 
for Sequoyah Nuclear Plant, Units 1 and 2 by removing the license 
conditions that reference the post-accident sampling system (PASS). The 
PASS information has been placed in the Sequoyah Final Safety Analysis 
Report (FSAR). This Change is consistent with NUREG-1431, ``Standard 
Technical Specifications--Westinghouse Plants.''
    Date of issuance: March 16, 1999.
    Effective date: March 16, 1999.
    Amendment Nos.: 243 and 233.
    Facility Operating License Nos. DPR-77 and DPR-79: The amendments 
revise the licenses.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20527). The August 21, 1995, letter provided clarifying information 
that did not change the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 16, 1999.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendment: April 23, 1998, as supplemented 
on January 25, 1999.
    Brief description of amendment: The amendment changes the existing 
requirements for the Residual Heat Removal Service Water (RHRSW), 
Station Service Water (SSW) and Alternate Cooling Tower Systems (ACS) 
as identified in Technical Specifications (TSs) 4.5.C and 3/4.5.D.
    Specifically, the changes are as follows:
    (1) Specifications 3.5.D.3 and 4.5.D.3: This requirement is revised 
to delete the existing allowance for 7 days of operation after both SSW 
subsystems are made or found to be inoperable.
    (2) Specification 4.5.C.1 and Specification 4.5.D.1: These 
requirements have been revised to relocate testing information related 
to pump flow and pressure testing characteristics for the RHRSW and SSW 
Systems, respectively, to the Technical Requirements Manual.
    (3) Specifications 3.5.D.1, 3.5.D.2, 3.5.D.3, 4.5.D.2, 4.5.D.3, and 
associated Bases: All references to SSW ``subsystem'' have been 
replaced by ``essential equipment cooling loop'' to more accurately 
reflect the Vermont Yankee design and operation. In addition, certain 
operability clarifications have been made to the Bases relative to 
affected Specifications.
    (4) Bases for Specification 3.5.D: The Bases have been revised to 
omit statements that imply that the ACS could provide adequate heat 
removal following a postulated accident. Other Bases additions have 
been made that include certain operability clarifications relative to 
affected Specifications.
    Date of Issuance: March 11, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 169.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.

[[Page 17040]]

    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6713).
    The January 25, 1999, supplement did not affect the original 
proposed no significant hazards consideration.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 11, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301.

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

    Date of amendment request: March 20, 1998, as supplemented by 
letters dated May 28, June 30, August 28, September 4, November 20, and 
December 8, 1998.
    Brief description of amendment: The amendment revised the technical 
specifications (TS) to support a modification to the plant to increase 
the storage capacity of the spent fuel pool and increase the nominal 
fuel enrichment to 5% weight percent of U-235. The amendment also 
revised the TS to allow the storage of an additional 279 assemblies in 
the cask loading pit.
    Date of issuance: March 22, 1999.
    Effective date: March 22, 1999, to be fully implemented no later 
than December 31, 1999, except that the racks in the cask loading pit 
may be installed at a future time after the completion of the next 
refueling outage.
    Amendment No.: 120.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 13, 1998 (63 FR 
37601). The June 30, August 28, September 4, November 20, and December 
8, 1998, supplemental letters provided additional clarifying 
information, did not expand the scope of the application as originally 
noticed, and did not change the staff's proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

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

    Date of amendment request: February 4, 1998, as supplemented by 
letter dated October 20, 1998.
    Brief description of amendment: The amendment revises the 
requirements in Technical Specification Tables 3.3-3, 3.3-4 and 4.3-2 
regarding the engineered safety features actuation system (ESFAS) 
Functional Unit 6.f, and adds a note to Table 4.3-2 to clarify the 
verification of time delays associated with ESFAS Functional Units 8.a 
and 8.b.
    Date of issuance: March 23, 1999.
    Effective date: March 23, 1999, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 121.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1998 (63 FR 
14491). The October 20, 1998, supplemental letter provided additional 
clarifying information, did not expand the scope of the application as 
originally noticed and did not change the staff's original proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated March 23, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

    Dated at Rockville, Maryland, this 31st day of March 1999.

    For the Nuclear Regulatory Commission.
Suzanne C. Black,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-8503 Filed 4-6-99; 8:45 am]
BILLING CODE 7590-01-P