[Federal Register Volume 63, Number 155 (Wednesday, August 12, 1998)]
[Notices]
[Pages 43200-43220]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-21724]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 20, 1998, through July 31, 1998. The 
last biweekly notice was published on July 29, 1998 (63 FR 40551).

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 September 11, 1998, the licensee may file a request for a 
hearing with respect to issuance of the amendment to the subject 
facility operating license and any person whose interest may be 
affected by this proceeding and who wishes to participate as a party in 
the proceeding must file a written request for a hearing and a petition 
for leave to intervene. Requests for a hearing and a petition for leave 
to intervene shall be filed in accordance with the Commission's ``Rules 
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2. 
Interested persons should consult a current copy of 10 CFR 2.714 which 
is available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or

[[Page 43201]]

petition; and the Secretary or the designated Atomic Safety and 
Licensing Board will issue a notice of a hearing or an appropriate 
order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs 
Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of amendment request: July 20, 1998.
    Description of amendment request: Baltimore Gas and Electric 
Company (BGE) request a modification involving replacing the service 
water (SRW) heat exchangers with new plate and frame heat exchangers 
having increased thermal performance capability. A similar license 
amendment dated February 8, 1998, was granted to Operating License No. 
DPR-53--Calvert Cliffs Nuclear Power Plant, Unit 1.
    The planned modification for Unit 2 is virtually identical to the 
one just completed for Unit 1 during the spring 1998 refueling outage. 
The only exception is the addition of an extra manual valve in the Unit 
2 system to isolate the bypass line for maintenance. This additional 
manual valve is needed due to the change in location of the tie-in to 
the main header. (The Unit 1 bypass line ties into the main header 
downstream of a control valve; therefore, it did not need a separate 
isolation valve for maintenance.)
    The saltwater and SRW piping configuration will be modified as 
necessary to allow proper fit-up to the new components. A flow control 
scheme to throttle saltwater flow to the heat exchangers and the 
associated bypass lines will be added. Saltwater strainers with an 
automatic flushing arrangement will be added upstream of each heat 
exchanger. The majority of the physical work associated with this 
modification is restricted to the SRW pump room.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    None of the systems associated with the proposed modification 
are accident initiators. The SW and SRW Systems are used to mitigate 
the effects of accidents analyzed in the UFSAR [Updated Final Safety 
Analysis Report]. The SW and SRW Systems provide cooling to safety-
related equipment following an accident. They support accident 
mitigation functions; therefore, the proposed modification does not 
increase the probability of an accident previously evaluated.
    The proposed modification will increase the heat removal 
capacity of the SRW System. The design provided under this activity 
ensures that the safety features provided by the SW and SRW are 
maintained, and in some instances enhanced; i.e., the availability 
of important-to-safety equipment required to mitigate the 
radiological consequences of an accident described in the UFSAR is 
enhanced by the

[[Page 43202]]

flexibility and increased thermal margin provided with this design.
    The redundant cooling capacity of the SW and SRW Systems have 
not been altered. Furthermore, the proposed activity will not 
change, degrade, or prevent actions described or assumed in any 
accident described in the UFSAR. The proposed activity will not 
alter any assumptions previously made in evaluating the radiological 
consequences of any accident described in the UFSAR. Therefore, the 
consequences of an accident previously evaluated in the UFSAR have 
not increased.
    Therefore, the proposed modification does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed activity involves modifying the SW and SRW System 
components necessary to support the installation of new SRW heat 
exchangers. None of the systems associated with this modification 
are identified as accident initiators in the UFSAR. The SW and SRW 
Systems are used to mitigate the effects of accidents analyzed in 
the UFSAR. None of the functions required of the SRW or SW System 
have been changed by this modification. This activity does not 
modify any system, structure, or component such that it could become 
accident initiator, as opposed to its current role as an accident 
mitigator.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The Safety design basis for the SW and SRW System is the 
availability of sufficient cooling capacity to ensure continued 
operation of equipment during normal and accident conditions. The 
redundant cooling capacity of these systems, assuming a single 
failure, is consistent with assumptions used in the accident 
analysis.
    The design, procurement, installation, and testing of the 
equipment associated with the proposed modification are consistent 
with the applicable codes and standards governing the original 
systems, structures, and components. The design of instruments and 
associated cabling ensures that physical and electrical separation 
of the two subsystems is maintained. Common-mode failure is not 
introduced by the activity. The equipment is qualified for the 
service conditions stipulated for that environment. New cable and 
raceways for this design will be installed in accordance with 
seismic design requirements. The additional electrical load has been 
reviewed to ensure the load limits for the vital 1E buses are not 
exceeded. The circuits and components related to the control valves 
control loops are safety-related, are similar to those used for the 
other safety-related flow control functions. The proposed 
modification will not have any adverse effects on the safety-related 
functions of the SW and SRW Systems.
    For the above reasons, the existing licensing bases have not 
been altered by the proposed modification. This activity will not 
reduce the margin of safety as it exists now. In fact, the margin of 
safety has been increased by this activity due to the increase in 
the thermal capacity of the dual train design (i.e., two heat 
exchangers per train versus one heat exchanger per train of the 
original design) and the increased availability of safety-related 
components.
    Therefore, this proposed modification does not significantly 
reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: July 13, 1998.
    Description of amendment request: The proposed amendments would 
revise the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 
BVPS-2) Updated Final Safety Analysis Report (UFSAR) descriptions of 
the Intake Structure main entrance and interconnecting cubicle doors. 
The current UFSAR descriptions state that the cubicle access doors are 
open to permit excess water from a major pipe rupture to flow out of 
the cubicles thereby avoiding internal flooding. The proposed changes 
would address a new failure mode of safety-related equipment that had 
not been previously considered for BVPS-1. The proposed changes would 
state that the cubicle interconnecting flood protection doors are 
normally closed with their inflatable seals depressurized and that the 
associated security/fire doors are normally closed. The proposed door 
closure arrangement is intended to protect the safety-related equipment 
in the interconnecting cubicles from the consequences of potential 
internal flooding.
    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 revises the text of the UFSAR for Unit 1 and 
Unit 2 to describe how protection is provided against potential 
internal floods in the cubicles that house the Unit 1 River Water 
and Unit 2 Service Water Pumps. The previous description concluded 
that the Unit 1 River Water pumps were protected because open 
cubicle access doors will permit excess water to flow out of the 
cubicles. The practice that has changed, and is described in the 
proposed revisions to the Unit 1 and Unit 2 UFSARs, will provide 
protection of the Unit 1 River Water Pumps and the Unit 2 Service 
Water Pumps so that no flooding event can adversely affect more than 
one Unit 1 or Unit 2 pump. Therefore, it can be concluded that the 
proposed changes do not involve any 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 effect of flooding the pump cubicles was considered in BVPS-
1 to have no adverse effect because open cubicle access doors would 
permit excess water to flow out of the cubicles, and pipe cracks in 
moderate energy piping was not part of the design basis. Revising 
the door arrangement described in the BVPS-1 UFSAR such that the 
security/fire doors are normally closed, requires that the effects 
of flooding be considered. Engineering analysis shows that a 
moderate energy pipe crack, (i.e., the BVPS-2 design basis internal 
flood), produces a leak rate of 1162 gpm, which results in a maximum 
water level of 0.82 feet, with the security/fire doors closed. The 
water level in the adjacent cubicle would reach a level at 0.37 
feet. This is below the level which would cause failures of the MCCs 
[Motor Control Centers] in the pump cubicles.
    The maximum leak rate from a failure of a Unit 1 rubber 
expansion joint in a pump cubicle would result in water rising to a 
level which would cause the MCCs to be flooded and fail; therefore, 
maintaining the flood door between the adjacent cubicles closed 
limits the impact to a single train.
    Failure of a single train of River Water is analyzed in the 
USAR; therefore, this change would not introduce a new or different 
type of accident.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change in the Unit 1 and Unit 2 UFSARs describes 
how protection is provided for the Unit 1 River Water, and the Unit 
2 Service Water pumps. Protection of the Unit 1 River Water Pumps 
and the Unit 2 Service Water pumps is provided so that no flooding 
event can adversely affect more than one Unit 1 or Unit 2 pump. 
Therefore, it can be concluded that the proposed changes do not 
involve any 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

[[Page 43203]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: July 9, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.7.1.1 and associated Bases for 
both units. TS 3.7.1.1 currently provides requirements for reducing the 
power range high neutron flux trip setpoint when one or more main steam 
safety valves are inoperable. The current basis for determining the 
amount of trip setpoint reduction has been determined to be non-
conservative. The proposed amendment would specify maximum allowable 
reactor power level based on the number of operable main steam safety 
valves rather than requiring a reduction in reactor trip setpoint. This 
change would be consistent with the NRC staff's guidance provided in 
the NRC's improved Standard Technical Specifications for Westinghouse 
plants (NUREG-1431, Revision 1). The maximum allowable reactor power 
level with inoperable safety valves would be calculated based on the 
recommendations of Westinghouse Nuclear Safety Advisory Letter (NSAL) 
94-01. The proposed change to the Unit 1 TS 3.7.1.1 would also delete 
reference to 2 loop operation since 2 loop operation is not a licensed 
condition for either unit.
    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 will generally incorporate the Improved 
Standard Technical Specification (ISTS) main steam safety valve 
(MSSV) requirements of NUREG-1431 into Specification 3.7.1.1 and 
associated Bases. The Unit 1 specification currently includes 
reference to 2 loop operating requirements in Action ``b'' and Table 
3.7-2. Reference to 2 loop operation is being deleted since it is 
not addressed in the ISTS and is not a licensed condition for these 
plants. The limiting condition for operation has been modified to 
incorporate the ISTS wording and requires MSSV operability in 
accordance with Tables 3.7-1 and 3.7-2. Table 3.7-1 lists the 
maximum allowable power level as a function of the number of 
operable MSSVs per steam generator and continues to require a 
minimum of 2 operable MSSVs per steam generator for continued plant 
operation. Table 3.7-2 specifies the MSSV lift setting and tolerance 
for each MSSV. The valve lift setting remains unchanged along with 
the current tolerance of +1 percent -3 percent. The Applicability 
statement has not been changed since it is consistent with the ISTS 
requirements.
    Proposed Action ``a'' applies with one or more inoperable MSSVs 
and requires that within 4 hours power must be reduced in accordance 
with the value specified in Table 3.7-1; otherwise, shut down. This 
action satisfies the same goal as the current action by restricting 
thermal power so that the energy transfer to the most limiting steam 
generator is not greater than the available relief capacity for that 
steam generator. Proposed Action ``b'' incorporates additional 
conservatism by specifically requiring at least 2 operable MSSVs per 
steam generator. This ensures that a minimum overpressure protection 
is available during all applicable modes of operation. Proposed 
Action ``c'' provides an exception to Specification 3.0.4 which does 
not allow entry into a mode where the Limiting Condition for 
Operation (LCO) is not met and actions require a shutdown. This 
exception is not addressed in the ISTS requirements; however, an 
exception to Specification 3.0.4 allows entry into a mode where the 
LCO applies in conformance with the action statements.
    Proposed Surveillance Requirement 4.7.1.1 requires verification 
of the lift setpoint for each MSSV listed in Table 3.7-2 in 
accordance with the Inservice Test Program. Note (1) is applied to 
Surveillance Requirement 4.7.1.1 to provide clarification of the 
testing requirements, such that this testing is required only in 
Modes 1 and 2 so that the plant can enter Modes 2 and 3 where this 
specification applies without first performing the test. A note (2) 
has been applied to the lift setting in Table 3.7-2 that requires a 
setting corresponding to the ambient conditions of the valve at the 
nominal operating temperature and pressure. The ISTS does not 
include this note but it has been included for consistency with the 
current note and provides a clear reminder to test personnel of the 
required test conditions.
    The safety valve Bases have been revised to generally 
incorporate the ISTS Bases which significantly improve the content 
and understanding of the MSSV requirements. These changes are 
consistent with the UFSAR [Updated Final Safety Analysis Report] 
design description and analysis assumptions where the MSSVs provide 
the required overpressure protection. The proposed changes are 
consistent with the regulations and provide additional assurance 
that the secondary side pressure remains within the bounds of the 
safety analyses; therefore, the proposed changes will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes generally incorporate the ISTS MSSV 
requirements to ensure adequate secondary side overpressure 
protection is available and properly maintained. The revised 
Limiting Condition for Operation (LCO) limits plant power level 
based on the number of operable MSSVs as stated in Table 3.7-1 and 
provides the valve lift settings and tolerances as shown in Table 
3.7-2. The actions require a reduction in power when the number of 
valves is less than the full complement for each steam generator and 
also require at least 2 operable MSSVs per steam generator. When 
these requirements cannot be met a plant shutdown is required. An 
action also provides an exception to Specification 3.0.4 and is 
consistent with the exception currently provided. These actions are 
more conservative than the current requirements and provide 
additional assurance that Specification 3.7.1.1 will continue to 
govern the MSSV limitations in a manner consistent with the accident 
analyses assumptions. The revised surveillance requirement provides 
clearly understandable testing requirements to ensure the MSSVs are 
adequately monitored and will perform in accordance with the 
accident analysis assumptions. The proposed change does not 
introduce any new mode of operation or require any physical 
modification to the plant; therefore, this change will 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 MSSVs ensure the ASME [American Society of Mechanical 
Engineers] Code, Section III requirements are maintained to limit 
the secondary system pressure to within 110 percent of the design 
pressure when passing the design steam flow. This ensures that the 
overpressure protection system can cope with all operational and 
transient events. Operation with less than the full number of MSSVs 
is permitted as long as thermal power is restricted to meet the ASME 
Code requirements. This limitation is provided in the proposed 
technical specifications along with operability and surveillance 
requirements to ensure the level of overpressure protection is 
maintained. MSSV operability is defined as the ability to open 
within the setpoint tolerances, relieve steam generator 
overpressure, and reseat when pressure has been reduced. MSSV 
operability is determined by surveillance testing in accordance with 
the Inservice Test program which provides assurance that the MSSVs 
will perform their designed safety functions to mitigate the 
consequences of accidents that could result in a challenge to the 
reactor coolant pressure boundary. The proposed change continues to 
ensure that the required components are properly maintained and that 
the assumed parameters are verified during the applicable conditions

[[Page 43204]]

and on a consistent basis; therefore, this change will not reduce 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: July 21, 1998.
    Description of amendment request: The proposed change request would 
permit an alternative to the requirement to perform Control Rod Drive 
(CRD) scram time testing with the reactor pressurized prior to resuming 
power operation. The change would permit: (1) scram time testing with 
the reactor depressurized prior to resuming operation, and (2) a second 
scram time test with the reactor pressure above 800 psig, prior to 
exceeding 40% reactor power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated; (or)
    There will not be an increase in the probability of occurrence 
of an accident previously evaluated in the Safety Analysis Report 
(SAR) because the requested change provides additional assurance 
that the CRD System is able to perform its safety function, and 
therefore does not change the probability of occurrence of an 
accident.
    There will not be an increase in the consequences of an accident 
previously evaluated in the Safety Analysis Report (SAR) because the 
requested change will ensure that the CRD System is able to perform 
its safety function, and therefore does not change the consequences 
of an accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated; (or)
    The requested change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The first issue associated with the requested change is increased 
wear on the CRDs, resulting in increased buffer seal wear or 
failure. This wear or failure of the buffer seal would result in 
difficulty or inability to withdraw the rod subsequent to the 
depressurized scram. The safety function of the rod to insert on a 
scram signal, however, would be unaffected by this seal degradation. 
Therefore, there is no safety concern with the increased wear due to 
performance of the cold scram test.
    The other consideration associated with the new requested change 
is the possible increased risk of stub tube leakage during the cold 
(depressurized) test. Without the download due to reactor pressure, 
the momentary upward loading on the CRD stub tube puts the stub tube 
into tension. Any flaws in the stub tube could grow and eventually 
result in a stub tube leak. The likelihood of flaws in the stub 
tubes, however, is very small, based on the extensive repair work on 
the stub tube surfaces performed prior to plant operation. The 
integrity of the stub tube repairs is verified by the 1000 pound 
leak test performed during every startup of the reactor. This test, 
therefore, poses very minimal risk of stub tube leakage.
    3. Involve a significant reduction in a margin of safety.
    The change will not decrease the margin of safety as defined in 
the basis of any Technical Specification. This is because the 
requested change, like the existing Technical Specification test, 
provides assurance that the CRD System is able to perform its safety 
function, and therefore does not change 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, Pitman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O.Thomas.

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

    Date of amendment request: June 11, 1998
    Description of amendment request: The proposed amendment would 
incorporate an alternative high radiation area control for Three Mile 
Island Nuclear Station, Unit No. 1 (TMI-1) in accordance with 10 CFR 
20.1601(c). The alternative would modify Technical Specification 6.12 
to allow for a conspicuously posted barricade and flashing light in 
individual high radiation areas that are located within large areas 
where no enclosure exists for locking, and no enclosure can be 
reasonably erected. A minor clarification to indicate that the 
requirement of paragraph 6.12.1.a also applies to 6.12.1.b and an 
editorial change were added.
    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 facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
proposed amendment involves changes to the TMI-1 Technical 
Specifications, which are consistent with Regulatory Guide 8.38. 
This change does not involve any change to system or equipment 
configuration. The proposed amendment incorporates an alternative 
high radiation area control, which has been previously found to be 
acceptable by the NRC. The reliability of systems and components 
relied upon to prevent or mitigate the consequences of accidents 
previous evaluated is not degraded by the proposed changes. 
Therefore, this change does not increase the probability or 
consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. This change only 
involves controls for access to high radiation areas. Access to 
plant equipment during normal or accident conditions will not be 
affected by utilizing this alternate method. Therefore, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment is consistent with Regulatory Guide 
8.38. The proposed amendment involves high radiation area access 
control and is not related to the margin of safety associated with 
any plant operation or transients. Therefore, it is concluded that 
operation of the facility in accordance with the proposed amendment 
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.

[[Page 43205]]

    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pitman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

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

    Date of amendment request: May 20, 1998.
    Description of amendment request: The proposed change would revise 
the Refueling Water Storage Tank (RWST) setpoint associated with 
Automatic Switchover to the Containment Sump. This change would require 
a revision to the Engineered Safety Features Actuation System 
Instrumentation Trip Setpoints, Table 3.3-4, Functional Unit 8.b, RWST 
Level--Low-Low, along with associated Bases Section 3/4.3.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors and does not alter the design assumptions 
affecting the ability of the RWST and the ECCS [Emergency Core 
Cooling System] pumps to mitigate the consequences of an accident.
    Revising the RWST Level Low-Low setpoint has a negligible effect 
on the operating margin for the RWST. The revised setpoint assures 
that the minimum RWST volume assumed in the accident analyses is 
injected prior to switchover to the recirculation mode. The effect 
on containment flood level, equipment qualification, and pH of the 
containment sump and the containment spray fluid, remain within the 
limits assumed in the accident analyses.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The setpoint change does not affect the function of the level 
monitoring channels or any function of the accident mitigation 
equipment associated with the RWST. No new components or physical 
changes are involved with this change. There are no changes to the 
source term, containment isolation or radiological release 
assumptions used in evaluating the radiological consequences in the 
Seabrook Station [updated final safety analysis report] UFSAR. The 
new setpoint will continue to initiate the automatic ECCS transfer 
from the injection mode to the recirculation mode and provide the 
alarm to alert the operator(s) to begin the manual actions necessary 
to complete the transfer to the recirculation mode. Manual operator 
action is required to complete the switchover to the recirculation 
mode. With the new setpoint, sufficient time remains available for 
the operator(s) to complete the transfer prior to receipt of the 
RWST EMPTY alarm and reaching the vortexing level in the RWST. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The design bases for the RWST Level Low-Low setpoint is to 
ensure that the minimum volume of water to support the assumptions 
made in the safety analysis is injected prior to switchover and that 
there is adequate time available for the operators to complete the 
manual actions necessary to complete the switchover to the 
recirculation mode prior to actuation of the RWST EMPTY alarm. The 
minimum injection volume assumed in the accident analyses, and time 
required for the operator(s) to initiate and complete manual actions 
to complete switchover to the recirculation mode prior to receipt of 
the RWST EMPTY alarm, remains unaffected by this change. Therefore, 
the proposed change does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Cecil O.Thomas.

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

    Date of amendment request: May 21, 1998.
    Description of amendment request: The proposed change would revise 
selected Technical Specification (TS) surveillance requirements to 
accommodate fuel cycles of up to 24 months for surveillances that are 
currently performed at each 18-month or other specified outage 
interval. Specifically, the following TS surveillance requirements 
would be revised by the proposed change: 4.1.3.3, Digital Rod Position 
Indication; 4.8.1.1.1.b, A.C. Sources--Operating--Transfer of 1E Bus 
Power from Normal to Alternate Source; 4.8.1.1.2.f.1 through 15, A.C. 
Sources--Operating--Emergency Diesel Generator Surveillances; 4.8.3.3, 
Onsite Power Distribution--Trip Circuit For Inverter I-2A; 4.8.2.1.c, d 
& f, D.C. Sources--Operating--125V D.C. Batteries and Chargers; 
4.8.4.2.a.1) & a.2), Containment Penetration Conductor Overcurrent 
Protective Devices and Protective Devices for Class 1E Power Sources 
Connected to Non-Class 1E Circuits; 4.8.4.3, Motor Operated Valves 
Thermal Overload Protection. In addition, the components listed in 
Technical Specification 4.8.2.2, D.C. Sources--Shutdown--125V DC 
Batteries and Chargers, have been evaluated to support an extension in 
frequency to 24 months (+25%).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, 
configuration of the facility or the manner in which the plant is 
operated. The proposed changes do not alter or prevent the ability 
of structures, systems, or components (SSCs) to perform their 
intended function to mitigate the consequences of an initiating 
event within the acceptance limits assumed in the Updated Final 
Safety Analysis Report (UFSAR). The proposed changes are 
administrative in nature and do not change the level of programmatic 
controls or the procedural details associated with aforementioned 
surveillance requirements.
    Changing the frequencies of the aforementioned surveillance 
requirements from at least once per 18 months to at least once per 
refueling interval does not change the basis for the frequencies. 
The frequencies were chosen because of the need to perform these 
verifications under the conditions that are normally found during a 
plant refueling outage, and to avoid the potential of an unplanned 
transient if these surveillances were conducted with the plant at 
power.
    Equipment performance over several operating cycles was 
evaluated to determine the impact of extending the surveillance 
intervals. This evaluation included a review of surveillance 
results, preventative maintenance records, and the frequency and 
type of corrective maintenance activities, a

[[Page 43206]]

failure mode analysis, and consultation with the respective system 
engineer. The evaluations conclude that the subject SSCs are highly 
reliable, that presently do not exhibit time dependent failure modes 
of significance, and that there is no indication that the proposed 
extension could cause deterioration in the condition or performance 
of the subject SSCs. There are no known mechanisms that would 
significantly degrade the performance of the evaluated equipment 
during normal plant operation. Although there have been generic or 
repetitive failures of some components in the past, which may have 
affected the ability of the SSCs to consistently and successfully 
perform their safety function, those items have been resolved 
through design changes and rework such that they have not recurred. 
There have been no repetitive failures or time dependent failures 
that were significant in nature which would have prevented the SSCs 
from performing their intended safety function.
    Deletion of the restriction ``during effect on safe operation of 
the plant is given prior to conduct of a particular surveillance in 
a condition or mode other than shutdown.
    Since the proposed changes only affect the surveillance 
intervals for SSCs that are used to mitigate accidents [sic], the 
changes do not affect the probability or consequence of a previously 
analyzed accident. While the proposed changes will lengthen the 
intervals between surveillances, the increase in intervals has been 
evaluated. Based on the reviews of the surveillance tests, 
inspections, and maintenance activities, it is concluded that there 
is no significant adverse impact on the reliability or availability 
of these SSCs.
    Since there are no changes to previous accident analyses, the 
radiological consequences associated with these analyses remain 
unchanged, therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed changes do not alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. There are no changes to the source term, 
containment isolation or radiological release assumptions used in 
evaluating the radiological consequences in the Seabrook Station 
UFSAR. Existing system and component redundancy is not being changed 
by the proposed changes. The proposed changes have no adverse impact 
on component or system interactions. The proposed changes are 
administrative in nature and do not change the level of programmatic 
controls and procedural details associated with the aforementioned 
surveillance requirements. Therefore, since there are no changes to 
the design assumptions, conditions, configuration of the facility, 
or the manner in which the plant is operated and surveilled, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    There is no adverse impact on equipment design or operation and 
there are no changes being made to the Technical Specification 
required safety limits or safety system settings that would 
adversely affect plant safety. The proposed changes are 
administrative in nature and do not change the level of programmatic 
controls and procedural details associated with the aforementioned 
surveillance requirements.
    From the evaluations performed on the subject SSCs there are no 
indications that potential problems would be cycle-length dependent 
or that potential degradation would be significant for the time 
frame of interest and, therefore, increasing the surveillance 
interval to the bounding limit of 30 months (24 months plus 25%) 
will have little, if any, adverse affect on safety.
    The proposed changes to the surveillance intervals are still 
consistent with the basis for the intervals and the intent and 
method of performing the surveillance is unchanged. Deletion of the 
restriction ``during shutdown'' where this restriction is stated 
will permit performance of certain maintenance and testing 
activities during conditions or modes other than shutdown. North 
Atlantic will ensure, through the implementation of appropriate 
administrative controls, that proper regard to their effect on safe 
operation of the plant is given prior to conduct of a particular 
surveillance in a condition or mode other than shutdown. In 
addition, use of the subject SSCs during normal plant operation, 
combined with their previous history of availability and 
reliability, provide assurance that the proposed changes will not 
affect the reliability of the subject SSCs. Thus, it is concluded 
that the subject SSCs would be available upon demand to mitigate the 
consequences of an accident and, therefore, there is no impact on 
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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Cecil O. Thomas.

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

    Date of amendment request: July 2, 1998.
    Description of amendment request: The proposed amendment would 
revise the updated Final Safety Analysis Report (FSAR) by changing FSAR 
Sections 9.7.2, ``Service Water,'' and 9.4, ``Reactor Building Closed 
Cooling Water,'' to discuss the use of various types of internal 
protective coatings and liners used in the piping and components of the 
systems. The proposed change also indicates that periodic maintenance, 
surveillances, and inspections would be conducted to ensure that 
coating or liner degradation would be promptly detected and corrected 
to provide reasonable assurance that the systems can perform their 
safety-related functions.
    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 significant hazards 
consideration because the changes would not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The SWS [Service Water System] provides cooling water directly 
or indirectly to a multitude of mitigating and support systems such 
as safety injection, containment spray, and RBCCW [Reactor Building 
Closed-Cooling Water]. Therefore either directly or indirectly, the 
SWS is credited in the mitigation of virtually all analyzed 
operating events and accidents. However, there are no failures of 
the SWS which would directly initiate any of the licensing basis 
accidents. Therefore, the probability of occurrence of accidents 
previously evaluated is not increased by this activity.
    The SWS is comprised of two separate and independent trains, 
each capable of providing the cooling capacity required for normal 
and accident operation. Therefore, the failure of a single heat 
exchanger or train will not influence the consequences of an 
accident. Only a common mode loss of SWS function could affect 
accident consequences. It can be postulated that lining material 
could be released as a result of the SWS response to an accident or 
as a result of a seismic event, resulting in heat exchanger blockage 
in both trains (common mode). However, the discussion below provides 
the basis for concluding that lining degradation will not increase 
the consequences of an accident.
    In response to a Safety Injection Actuation Signal or a Loss of 
Normal Power event, the quantity of flow in safety related SWS heat 
exchangers may increase significantly, imparting higher loads on the 
pipe linings than are typically present during normal operation. In 
spite of this flow increase, it is considered to be much more likely 
that any lining degradation will occur and be detected under normal 
operating conditions, and will be corrected prior to the occurrence 
of an event of the type discussed above. SWS pump flow 
surveillances, performed periodically during normal operation, 
subject significant portions of the SWS to flow levels which equal 
or exceed those expected to occur during accidents. Any degraded 
lining material prone to be released during an

[[Page 43207]]

accident is expected to be released during these pump surveillances. 
The inspections, operating procedures, and surveillances ensure that 
significant lining releases will be promptly detected and 
investigated. In addition, SWS design features provide the system 
with a significant level of protection against degraded lining 
debris (e.g., standby spare RBCCW heat exchanger and EDG [Emergency 
Diesel Generator] engine cooler strainers) both during normal 
operation and while responding to an accident.
    An evaluation was performed to assess the significance of 
loading on the linings due to a postulated seismic event. The 
importance of seismic loads depends upon their magnitude relative to 
normal operating loads, and on their relative frequency of 
occurrence. Normal operating loads include steady state flow loads 
as well as transients due to pump swaps and realignments for 
surveillances. The evaluation determined that normal operating loads 
are significantly greater than anticipated seismic loads concurrent 
with steady state flow loads. Therefore, if normal operating loads 
do not cause lining to become detached, it is very unlikely that a 
random seismic event would cause detachment. In addition, while flow 
loads are continuously present in most of the system and normal 
transients occur many times during an operating cycle, seismic 
events at the Millstone site are very infrequent (the repetition 
rate of an OBE [Operating Basis Earthquake] is hundred of years). 
Should normal operating loads cause lining detachment, it is much 
more probable that this released material will be detected, and the 
degraded condition corrected, prior to the occurrence of a seismic 
event.
    Based upon these discussions, and given the random nature of 
lining degradation and the scrutiny with which the SWS is operated 
and maintained, it is not considered to be credible that the 
operability of both SWS trains will be simultaneously impaired by 
lining degradation and release.
    Therefore, there is no significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    As discussed above, the failure of a single heat exchanger or a 
single SWS train will not cause an accident. Only a common mode loss 
of SWS function could create the possibility of a previously 
unanalyzed accident, and this loss would not directly initiate an 
accident. However, for the reasons discussed above, lining 
degradatiion will not cause common mode failures to occur.
    Therefore, the change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margins of safety of the protective boundaries (fuel matrix/
cladding, reactor coolant system pressure boundary, and containment) 
would not be impacted by the postulated release of lining material 
into the SWS. The accident analyses in the FSAR [Final Safety 
Analysis Report] demonstrate the performance of the protective 
boundaries. As discussed previously, it is not considered to be 
credible that lining degradation will cause a common mode loss of 
SWS function. Therefore, since the accident analyses credit only one 
SWS train, released lining would not affect accident analyses 
assumptions. On this basis, it is concluded that margins of safety 
as demonstrated by the accident analyses would not be affected by 
postulated lining material release.
    Therefore, the change will not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: July 17, 1998.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) surveillance requirements for 
the onsite emergency diesel generators (EDGs) to achieve an overall 
improvement in the EDGs reliability and availability. The proposed 
changes would modify the requirement for operability tests of an EDG 
when the other EDG is inoperable, delete the requirement for 
operability tests when one or both offsite A.C. sources are inoperable, 
eliminate fast loading of the EDGs except for the 18-month testing, and 
eliminate fast starts (15 seconds) except for once per 6 months and 
during the 18-month testing. These proposed changes are generally 
consistent with the guidance provided in Generic Letter (GL) 84-15, 
``Proposed Staff Actions to Improve and Maintain Diesel Generator 
Reliability,'' dated July 2, 1984, and GL 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation,'' dated September 27, 1993. 
Justification for deviations from the guidance provided in the GLs is 
provided in the licensee's submittal.
    In addition, the licensee proposes to revise the wording in the TS 
requirements for offsite circuits to be consistent with NUREG-0212, 
``Standard Technical Specifications for Combustion Engineering 
Pressurized Water Reactors,'' Revision 2, fall 1980, and the guidance 
provided in GL 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate 24-Month Fuel Cycle,'' dated April 2, 1991. 
The associated TS Bases will be updated to reflect the proposed 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The LCOs [Limiting Conditions for Operation] for Technical 
Specifications [TSs] 3.8.1.1 and 3.8.1.2 will be changed to require 
a transmission network between offsite power and the onsite Class 1E 
distribution system, instead of just between offsite and the 
switchyard. This change, which will expand the requirement, is 
consistent with the current Millstone Unit No. 2 interpretation of 
the required distribution system. Therefore, the proposed changes 
will not result in a significant increase in the probability or 
consequences of an accident previously analyzed.
    The diesel generators (DGs) supply power to the emergency busses 
at Millstone Unit No. 2 in the event of a loss of normal power 
(LNP). The emergency busses supply the vital equipment used to 
mitigate the consequences of design basis accidents. Therefore, the 
diesel generators are vital equipment used to mitigate the 
consequences of design basis accidents. Failure of the DGs will not 
cause a design basis accident to occur. However, failure of the DGs 
will affect the consequences of design basis accidents if a 
concurrent LNP occurs.
    The proposed changes will revise the action requirements 
regarding operability testing of the DGs. The requirement to test 
the DGs if offsite circuits are inoperable will be deleted. An 
inoperable offsite circuit, by itself, will not affect the 
operability of the DGs. The requirement to test the remaining 
operable DG if one DG is inoperable will be modified. Testing will 
not be required provided a common cause failure is not the reason 
for declaring the DG inoperable. The requirement contained in the 
first footnote (*) to Technical Specification 3.8.1.1 to complete 
the test of the remaining DG will be deleted. The need to test the 
remaining DG will be based on the determination of a common cause 
failure. These changes will improve DG reliability by reducing the 
number of unnecessary starts and by requiring more appropriate 
testing of the DGs when there is a potential for common mode

[[Page 43208]]

failure. The proposed changes to the action requirements will not 
change the response of the DGs to an LNP. Therefore, the proposed 
changes will not result in a significant increase in the probability 
or consequences of an accident previously analyzed.
    The requirement contained in the second footnote (**) to 
Technical Specification 3.8.1.1 to allow a one time extension of the 
allowed outage time to 7 days will be deleted. This provision is no 
longer necessary since the Millstone Unit No. 1 work has been 
completed. The statements that a successful test of the DG performed 
for the current Action Statements c, d, or e will satisfy the 
required testing of Action States a or b are no longer necessary 
with the proposed changes. These statements will be deleted. The 
removal of these items will not change the response of the DGs to an 
LNP. Therefore, these proposed changes will not result in a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    The proposed changes to the DG surveillance requirements will 
allow an engine prelube period before all DG tests starts, allow 
slow starting of the DGs, and allow the DGs to be loaded in 
accordance with manufacturer recommendations. This will decrease the 
wear on the DGs. The proposed changes will also allow adequate time 
for the completion of all manufacturer recommended DG engine prelube 
procedures. Modifying starting and loading requirements, consistent 
with the manufacturer recommendations, is intended to enhance diesel 
reliability by minimizing severe test conditions which can lead to 
premature failures. In addition, specifying that the 184 day DG SRs 
[surveillance requirements] will satisfy the 31 day DG starting and 
loading SRs will eliminate redundant testing. These proposed changes 
will minimize unnecessary DG testing while maintaining DG 
reliability. The proposed changes will not change the response of 
the DGs to an LNP. Therefore, these changes will not result in a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    The ASTM [American Society for Testing and Materials] standards 
referenced for diesel fuel oil sampling will be modified in SR 
4.8.1.1.2.b. The proposed changes will replace an outdated standard, 
and will remove the year of issuance or revision from the ASTM 
standards referenced. This will allow use of the current approved 
ASTM standard. These proposed changes do not affect the sampling 
frequency or acceptance criteria of this SR. Therefore, the proposed 
changes will not result in a significant increase in the probability 
or consequences of an accident previously analyzed.
    The proposed wording changes to eliminate any possible confusion 
when SRs 4.8.1.1.1 and 4.8.1.1.2 are referenced by SR 4.8.1.2, to 
state that the DGs start from standby conditions instead of ambient 
conditions, and to remove the requirement to perform a DG 
surveillance only during shutdown will not affect any technical 
aspect of the SRs. Therefore, the proposed changes will not result 
in a significant increase in the probability or consequences of an 
accident previously analyzed.
    SRs will be added to test the DGs every 184 days at conditions 
similar to the current 31 day SRs. These conditions are more 
restrictive than the new proposed 31 day SRs. The 184 day SRs will 
require the diesel generators to start and obtain speed and voltage 
within 15 seconds and will also require the diesel generators to be 
synchronized, loaded, and to maintain the load for at least 60 
minutes. However, it will allow gradual loading, based on 
manufacturer recommendations, to be used. A 184 day surveillance 
interval is sufficient to verify DG fast-start capability, and is 
consistent with GL [Generic Letter] 84-15, GL 93-05, and NUREG-1432. 
Therefore, the posed changes will not result in a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    The list of SRs, contained in SR 4.8.1.2, that do not have to be 
performed for the operable diesel generator in Modes 5 and 6 will be 
expanded to take into account the 184 day DG SR that will be added. 
This proposed change will exclude the one operable DG from being 
loaded when the 184 day SR is performed. This is consistent with the 
current SR which excludes performance of SR 4.8.1.1.2.a.3. Loading 
the one required operable diesel generator could subject this diesel 
generator to grid faults which could adversely affect its ability to 
perform its safety function. Therefore, the proposed change will not 
result in a significant increase in the probability or consequences 
of an accident previously analyzed.
    The Bases of these Technical Specifications will be modified and 
expanded to discuss the proposed changes, and to provide guidance to 
ensure the requirements are correctly applied. Therefore, the 
proposed changes will not result in a significant increase in the 
probability or consequences of an accident previously analyzed.
    These proposed changes do not alter the way any structure, 
system, or component functions. The intent of the proposed changes 
is to improve the reliability of the DGs by eliminating unnecessary 
surveillance testing and allowing most of the surveillance testing 
to be performed in accordance with the recommendations of the 
manufacturer. There will be no adverse effect on equipment important 
to safety. The response of the DGs to an LNP, as described in the 
Millstone Unit No. 2 FSAR [Final Safety Analysis Report], will 
remain the same. There will be no effect on any of the design basis 
accidents previously evaluated. Therefore, this License Amendment 
Request will not result in a significance increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of an 
accident from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    This License Amendment Request proposes to modify the LCOs for 
electrical power sources, DG surveillance requirements and the 
required actions for inoperable electrical power sources contained 
in the Millstone Unit No. 2 Technical Specifications. The proposed 
changes will revise LCO wording to be consistent with the required 
offsite power distribution requirements and improve DG reliability 
by minimizing excessive wear of the DGs, and changing the starting 
and loading requirements of the DGs, in accordance with manufacturer 
recommendations, during most DG surveillance and operability tests. 
Improving the reliability of the DGs will help ensure the DGs will 
respond to an LNP as described in the Millstone Unit No. 2 FSAR. 
Therefore, this License Amendment Request will not result in a 
significant reduction in the margin of safety as defined in the 
Bases for the Technical Specifications addressed 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 proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: July 21, 1998.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) by changing various Reactor 
Protection System (RPS) and Engineered Safety Features Actuation System 
(ESFAS) setpoints and allowable values; correct the specified maximum 
reactor power level limited by the high power level RPS trip; add new 
TS and requirements associated with the automatic isolation of steam 
generator blowdown; and make several editorial and changes to correct 
various errors

[[Page 43209]]

and to provide needed clarification. The applicable TS Bases sections 
would also be changed to reflect the proposed changes, correct previous 
errors identified during the licensee's review of the TS, eliminate 
redundant information, and expand the TS Bases to discuss the new 
requirements for the automatic isolation of the steam generator 
blowdown.
    Specifically, the proposed changes would modify TS 2.1.1, ``Safety 
Limits--Reactor Core,'' TS 2.2.1, ``Limiting Safety System Settings--
Reactor Trip Setpoints,'' TS 3.3.1.1, ``Instrumentation--Reactor 
Protective Instrumentation'' TS 3.3.2.1, ``Instrumentation--Engineered 
Safety Features Actuation System Instrumentation,'' and would add a new 
TS 3.7.1.8, ``Plant Systems--Steam Generator Blowdown Isolation 
Valves.'' As previously noted, the applicable TS Bases sections will be 
updated to reflect the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to correct the maximum reactor power level 
from 112% to 111.6% is consistent with the maximum high power trip 
setpoint of 106.6%, plus 5% uncertainty, currently used in the 
safety analyses. This does not change the Technical Specification 
required high power reactor trip setpoint. There will be no adverse 
effect on any design basis accident previously evaluated or on any 
equipment important to safety. Therefore, the proposed change will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to the trip setpoints and allowable values 
for the Reactor Protection System (RPS) trips on high pressurizer 
pressure, high containment pressure, low steam generator pressure, 
and low steam generator level are the result of revisions to the 
instrument loop uncertainty and setpoint calculations. These 
calculations were revised to incorporate calculation methodology 
changes, analytical limit changes, correct errors identified, and to 
include the effects of a harsh environment (pressure, temperature, 
and radiation), where appropriate. The proposed setpoints and 
allowable values will ensure a reactor trip signal is generated at, 
or before the analytical limits used in the respective accident 
analyses are reached. There will be no adverse effect on any design 
basis accident previously evaluated or on any equipment important to 
safety. Therefore, the proposed changes will not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes to the trip setpoints and allowable values 
for the Engineered Safety Features Actuation System (ESFAS) 
actuations on low pressurizer pressure, high containment pressure, 
low steam generator pressure, low refueling water storage tank 
level, and low steam generator level are the result of revisions to 
the instrument loop uncertainty and setpoint calculations. These 
changes were revised to incorporate calculation methodology changes, 
analytical limit changes, correct errors identified, and to include 
the effects of a harsh environment (pressure, temperature, and 
radiation), where appropriate. The proposed setpoints and allowable 
values will ensure an ESF [engineered safety feature] actuation 
signal is generated at, or before the analytical limits used in the 
respective accident analyses are reached. There will be no adverse 
effect on any design basis accident previously evaluated or on any 
equipment important to safety. Therefore, the proposed change will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to add Technical Specification requirements 
for the steam generator blowdown isolation valves will provide 
additional assurance that the automatic isolation of steam generator 
blowdown will occur as assumed in the loss of main feedwater 
accident analysis. There will be no adverse effect on any design 
basis accident previously evaluated or on any equipment important to 
safety. Therefore, the proposed changes will not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change to the value of steam generator pressure 
when the steam generator low pressure reactor trip can be bypassed 
(from 780 psia to 800 psia) will reduce the range of plant operation 
when this trip is required to be available. However, this will not 
affect the range of plant operation when this RPS trip is required 
to be operable. This RPS trip is required in Modes 1 and 2. The 
expected steam generator pressure during a reactor startup (entry 
into Mode 2) is approximately 900 psia, which corresponds to a 
Reactor Coolant System (RCS) temperature of approximately 532 deg.F. 
The proposed change will require the bypass to be automatically 
removed prior to exceeding a steam generator pressure of 800 psia. 
There will be no adverse effect on any design basis accident 
previously evaluated or on any equipment important to safety. 
Therefore, the proposed change will not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to the value of pressurizer pressure (from 
1750 psia to 1850 psia) when the pressurizer low pressure ESF 
actuations (SIAS, CIAS, and EBFAS) [safety injection actuation 
system, containment isolation actuation system, and enclosure 
building filtration actuation system] can be blocked will reduce the 
range of plant operation when these functions are required to be 
available. However, since the plant would normally be in Mode 3 when 
pressurizer pressure is in this range, automatic actuation of these 
ESF functions on high containment pressure, as well as manual 
actuation, is required to be operable. In addition, the plant would 
not normally maintain pressurizer pressure between 1750 psia and 
1850 psia. Therefore, since automatic actuation of these ESF 
functions on high containment pressure, as well as manual actuation, 
should be operable, and the time the plant will operate between 1750 
psia and 1850 psia is small, the ESFAS will continue to function as 
before. There will be no adverse effect on any design basis accident 
previously evaluated or on any equipment important to safety. 
Therefore, the proposed change will not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to the value of steam generator pressure 
(from 600 psia to 700 psia) when the steam generator low pressure 
ESF actuation (main steam line isolation) can be blocked will reduce 
the range of plant operation when this function is required to be 
available. However, since the plant would be in Mode 3 when steam 
generator pressure is in this range (RCS temperature of 
approximately 486 deg.F to 503 deg.F), automatic actuation of this 
ESF function on high containment pressure, as well as manual 
actuation, is required to be operable. In addition, the plant would 
not normally maintain steam generator pressure between 600 psia and 
700 psia. Therefore, since automatic actuation of this ESF function 
on high containment pressure, as well as manual actuation, should be 
operable, and the time the plant will operate between 600 psia and 
700 psia is small, the ESFAS will continue to function as before. 
There will be no adverse effect on any design basis accident 
previously evaluated or on any equipment important to safety. 
Therefore, the proposed change will not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The minor editorial and non-technical changes to correct 
spelling errors, correct a capitalization error, add page amendment 
numbers, add the specific plant parameter (steam generator pressure) 
to use if an RPS or ESF function can be bypassed, change the value 
of the parameter (pressurizer pressure) used in action statements, 
and a ``[less than or equal to]'' symbol, change ``value'' to 
``setpoint,'' and update the index will have no effect on plant 
operation. These changes will not result in any technical changes to 
the Millstone Unit No. 2 Technical Specifications. There will be no 
adverse effect on any design basis accident previously evaluated or 
on any equipment important to safety. Therefore, the proposed change 
will not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to the Technical Specification Bases will 
incorporate the RPS and ESFAS setpoint changes, correct errors, 
eliminate redundant information, and expand the Bases to discuss the 
new requirements for steam generator blowdown isolation. These 
changes will have no effect on equipment operation. There will be no 
adverse effect on any design basis accident

[[Page 43210]]

previously evaluated or on any equipment important to safety. 
Therefore, the proposed changes will not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes have no adverse effect on any of the design 
basis accidents previously evaluated and have no adverse effect on 
how the RPS and ESFAS function to mitigate the consequences of 
design basis accidents. Therefore, the license amendment request 
does not impact the probability of an accident previously evaluated 
nor does it involve a significant increase in the consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will correct the maximum reactor power 
level specified; change RPS trip setpoints, allowable values, and 
bypass setpoints; change ESFAS trip setpoints, allowable values, and 
block setpoint changes; add a new Technical Specification and 
additional requirements associated with the automatic isolation of 
steam generator blowdown; and make various minor editorial and non-
technical changes. There will be no adverse effect on equipment 
important to safety. The RPS and ESFAS will continue to function as 
designed to mitigate the consequences of design basis accidents. 
Therefore, there will be no significant reduction of the margin of 
safety as defined in the Bases for the Technical Specifications 
affected 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 proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna 
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

    Date of amendment request: June 19, 1998.
    Description of amendment request: The amendment to Unit 1 Technical 
Specifications (TS) involves the addition of a new section entitled 
``Oscillation Power Range Monitoring (OPRM) Instrumentation'' and 
revisions to Section 3.4.1 ``Recirculation Loops Operating'' to remove 
the specifications related to thermal power stability which will not be 
required after the installation of the OPRM instrumentation. Unit 1 is 
currently operating under Interim Corrective Actions (ICAs) defined in 
TS 3.4.1 that specify restrictions on plant operation and actions by 
operators in response to instability events. The OPRM system provides 
an automatic long-term solution to the instability issue and eases the 
burden on the operator.
    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 OPRM most directly affects the APRM and LPRM portions of the 
Power Range Neutron Monitoring system. Its installation does not 
affect the operation of these sub-systems. None of the accidents or 
equipment malfunctions affected by these sub-systems are affected by 
the presence or operation of the OPRM.
    The APRM channels provide the primary indication of neutron flux 
within the core and respond almost instantaneously to neutron flux 
changes. The APRM Fixed Neutron Flux-High function is capable of 
generating a trip signal to prevent fuel damage or excessive reactor 
pressure. For the ASME overpressurization protection analysis in 
FSAR Chapter 5, the APRM Fixed Neutron Flux-High function is assumed 
to terminate the main steam isolation valve closure event. The high 
flux trip, along with the safety/relief valves, limit the peak 
reactor pressure vessel pressure to less than the ASME Code limits. 
The control rod drop accident (CRDA) analysis in Chapter 15 takes 
credit for the APRM Fixed Neutron Flux-High function to terminate 
the CRDA. The Recirculation Flow Controller Failure event (pump 
runup) is also terminated by the high neutron flux trip. The APRM 
Fixed Neutron Flux-High function is required to be OPERABLE in MODE 
1 where the potential consequences of the analyzed transients could 
result in the Safety Limits (e.g., MCPR and Reactor pressure) being 
exceeded.
    The installation of the OPRM equipment does not increase the 
consequences of a malfunction of equipment important to safety. The 
APRM and RPS systems are designed to fail in a tripped (fail safe) 
condition; the OPRM will have no affect on the consequence of the 
failure of either system. An inoperative trip signal is received by 
the RPS any time an APRM mode switch is moved to any position other 
than Operate, an APRM module is unplugged, the electronic operating 
voltage is low, or the APRM has too few LPRM inputs. These functions 
are not specifically credited in the accident analysis, but are 
retained for the RPS as required by the NRC approved licensing 
basis.
    The OPRM allows operation under current operating conditions 
presently restricted by the current Technical Specifications by 
providing automatic suppression functions in the area of concern in 
the event an instability occurs. The consequences of any accident or 
equipment malfunction are not increased by operating under those 
conditions. Although protected by the OPRM from thermal-hydraulic 
core instabilities above 30% core power, operation under natural 
core recirculation conditions is not allowed. No accidents or 
transients of a type not analyzed in the FSAR are created by 
operating under these conditions with the protection of the OPRM 
system.
    This change does not increase the probability of an accident as 
previously evaluated. The OPRM is designed and installed to not 
degrade the existing APRM, LPRM, and RPS systems. These systems will 
still perform all of their intended functions. The new equipment is 
tested and installed to the same or more restrictive environmental 
and seismic envelopes as the existing systems. The new equipment has 
been designed and tested to the electromagnetic interference (EMI) 
requirements of Reference 2, which assures correct operation of the 
existing equipment. The new system has been designed to single 
failure criteria and is electrically isolated from equipment of 
different electrical divisions and from non-1E equipment. The 
electrical loading is within the capability of the existing power 
sources and the heat loads are within the capability of existing 
cooling systems. The OPRM allows operation under operating 
conditions presently forbidden or restricted by the current 
Technical Specifications. No other transient or accident analysis 
assumes these operating restrictions.
    Based upon the analysis presented above, PP&L 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 OPRM system is a monitoring and accident mitigation system that 
cannot create the possibility for an accident.

[[Page 43211]]

    The OPRM will allow operation in conditions currently restricted 
by the current Technical Specifications. Although protected by the 
OPRM from thermal-hydraulic core instabilities above 30% core power, 
operation under natural circulation conditions is not allowed. No 
accidents or transients of a type not analyzed in the FSAR are 
created by operating under these conditions with the protection of 
the OPRM system. No new failure modes of either the new OPRM 
equipment or of the existing APRM equipment have been introduced. 
Quality software design, testing, implementation and module self-
health testing provides assurance that no new equipment malfunctions 
due to software errors are created. The possibility of an accident 
of a new or different type than any evaluated previously is not 
created.
    The new OPRM equipment is designed and installed to the same 
system requirements as the existing APRM equipment and is designed 
and tested to have no impact on the existing functions of the APRM 
system. Appropriate isolation is provided where new interconnections 
between redundant separation groups are formed. The OPRM modules 
have been designed and tested to assure that no new failure modes 
have been introduced.
    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.
    There has been no reduction in the margin of safety as defined 
in the basis for the Technical Specifications. The OPRM system does 
not negatively impact the existing APRM system. As a result, the 
margins in the Technical Specifications for the APRM system are not 
impacted by this addition.
    Current operation under the ICAs provides an acceptable margin 
of safety in the event of an instability event as the result of 
preventive actions and Technical Specification controlled response 
by the control room operators. The OPRM system provides an increase 
in the reliability of the protection of the margin of safety by 
providing automatic protection of the MCPR safety limit, while the 
protection burden is significantly reduced for the control room 
operators. This protection is demonstrated as described above, and 
in the NRC reviewed and approved Topical Reports NEDO-32465-A and 
CENPD-400-P-A.
    Replacement of the ICA operating restrictions from Technical 
Specifications with the OPRM system does not affect the margin of 
safety associated with any other system or fuel design parameter.

    Therefore, the change does not involve a reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: July 6, 1998
    Description of amendment request: The proposed Technical 
Specification (TS) changes represent revisions to the Radiological 
Effluent Technical Specification (RETS) Section 3.5.b.1, ``Main 
Condenser Steam Jet Air Ejector (SJAE)'' and Table 3.10-1 ``Radiation 
Monitoring Systems that Initiate and/or Isolate Systems'' including 
associated TS Bases. The existing RETS for radiation monitoring 
instrumentation systems that initiate and/or isolate systems will be 
changed by adding Allowable Outage Times (AOTs) and incorporating 
editorial and administrative changes to clarify requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The inherent redundancy and reliability of the protective 
instrumentation trip systems ensure that the consequences of an 
accident are not significantly increased. In addition, the 
restrictive Allowable Outage Time (AOT) interval limits the 
probability of the protective instrument channel being unavailable 
and an accident requiring its function from occurring 
simultaneously. The requirement that the associated trip function 
maintains trip capability for selected instrumentation ensures that 
the protective instrumentation response will occur such that the 
consequences of an accident are not different from those previously 
evaluated. The proposed changes provide AOTs for test and repair of 
plant instrumentation. The changes do not introduce any new modes of 
plant operation, make any physical changes, or alter any operational 
setpoints. Therefore, the changes do not degrade the performance of 
any safety system assumed to function in the accident analysis. 
Consequently, there is no effect on the probability of occurrence of 
an accident.
    Regarding the consequences of an accident, the GE Licensing 
Topical Reports (References 1 and 2) [GE Topical Report NEDC-31677P-
A, ``Technical Specification Improvement Analysis for BWR Isolation 
Actuation Instrumentation,'' July 1990 and GE Topical Report GENE-
770-06-1-A, ``Bases for Changes to Surveillance Test Intervals and 
Allowed Out-Of-Service Times for Selected Instrumentation Technical 
Specifications,'' December 1992] conclude that the proposed AOT for 
the safety system instrumentation results in an insignificant change 
in the core damage frequency. The AOTs result in a slight increase 
in the unavailability of the safety functions. The overall effect on 
the probability of an accident is negligible. The NRC concurred in 
their SERs [safety evaluation reports] (References 3 and 4) [NRC 
Safety Evaluation Report, letter from Charles E. Rossi, NRC to S.D. 
Floyd, BWR Owners Group, ``General Electric Company Topical Report 
NEDC-31677P, Technical Specification Improvement Analysis for BWR 
Isolation Actuation Instrumentation'', June 18, 1990 and NRC Safety 
Evaluation Report, letter from Charles E. Rossi, NRC to R.D. Binz, 
BWR Owners Group, ``General Electric Company Topical Report GENE-
770-06-1, Bases for Changes to Surveillance Test Intervals and 
Allowed Out-Of-Service Times for Selected Instrumentation Technical 
Specifications,'' July 21, 1992] with this conclusion. Consequently, 
there is not a significant increase in the consequences of an 
accident.
    Since the editorial and administrative items do not alter the 
meaning or intent of any requirements, they do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the protective instrumentation trip 
system specifications do not create the possibility of a new or 
different kind of accident because they do not introduce any new 
operational modes or physical modifications to the plant.
    For systems with only one channel (Main Control Room 
Ventilation) or two-out-of-two logic system (SJAE Radiation 
Monitors) a six-hour surveillance AOT is being proposed and a repair 
time AOT is not allowed. This is consistent with GE Topical Reports 
referenced in current TS Bases 4.2 and STS [Standard Technical 
Specifications] and therefore, will not introduce a new or different 
kind of accident than previously evaluated.
    Since the editorial and administrative items do not alter plant 
configurations or operating modes, they do not create the 
possibility of a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The protective instrumentation surveillance requirements provide 
verification of the operability of the trip system instrumentation 
channels. In addition, the redundant channel that monitors the 
identical Trip Function maintains trip capability for the relatively 
short duration of the test or repair time period. This ensures that 
protective

[[Page 43212]]

instrumentation reliability is maintained. The proposed change 
provides for a specific time period to perform required 
surveillances on instrument channels without trips present in 
associated trip systems. This time allotment tends to enhance the 
margin of safety by decreasing the probability of unnecessary 
challenges to safety systems and inadvertent plant transients. The 
evaluations presented in the referenced GE Licensing Topical Reports 
concluded that the overall effect of the proposed changes provides a 
net increase in plant safety.
    The only action resulting from the proposed changes to RETS is 
to add AOTs for selected instrumentation. Spurious signals during 
testing could initiate plant transients. These transients are 
bounded by the current transient analysis. These tests do not 
subject the instruments to any conditions beyond their design 
specifications and are performed in accordance with approved testing 
standards. This testing ensures equipment operability by identifying 
degraded conditions, initiating corrective action and properly 
retesting them. Therefore, the proposed RETS do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director

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

    Date of amendment request: June 25, 1998.
    Description of amendment request: The proposed changes affect 
Technical Specification (TS) Surveillance Requirement 4.5.1.d.2.b by 
deleting the requirement to perform in-situ functional testing of the 
Automatic Depressurization System (ADS) safety relief valves (SRVs) 
during startup testing activities. The proposed changes also affect TS 
Surveillance Requirement 4.4.2.1.b such that the 18-month channel 
calibration for the SRV acoustic monitors will no longer require an 
exception to the provisions of TS 4.0.4, nor adjustments to SRV full 
open noise levels.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS change does not involve any physical changes to 
plant structures, systems or components (SSC). The ADS will continue 
to function as designed. The ADS is an Emergency Core Cooling System 
(ECCS) designed to mitigate the consequences of an accident, and 
therefore, can not contribute to the initiation of any accident. The 
ADS utilizes five of the 14 main steam line SRVs as the primary 
method for depressurizing the reactor pressure vessel to permit low 
pressure core cooling capability in the event of a small break Loss-
of-Coolant-Accident (LOCA) if the high pressure cooling systems 
(i.e., High Pressure Cooling Injection (HPCI) and Reactor Core 
Isolation Cooling (RCIC) systems) fail to maintain adequate reactor 
vessel water level.
    Deleting the TS surveillance requirements to perform the in-situ 
testing of the ADS/SRVs during startup, as proposed, should reduce 
the probability of an inadvertent opening of an SRV as discussed in 
Section 15.1.4 of the Hope Creek [Updated Final Safety Analysis 
Report] UFSAR since deleting this testing requirement will eliminate 
a known initiator of SRV pilot leakage and subsequent erosion. This 
proposed TS change will have a tendency to increase, rather than 
decrease, the reliability of the ADS/SRVs by eliminating the in-situ 
ADS functional startup testing. The probability of the ADS/SRVs to 
open on demand has been demonstrated to be extremely high and is not 
measurably improved through the in-situ ADS functional startup 
testing.
    Using the provisions of 10CFR50.59, PSE&G will establish a 
method for performing SRV acoustic monitor channel calibration that 
does not require reactor steam pressure or SRV opening. This testing 
method will comply with the current TS definition of CHANNEL 
CALIBRATION. Since the notes associated with TS Surveillance 
Requirement 4.4.2.1 (providing a compliance exception to the 
provisions of TS 4.0.4 to allow for proper reactor steam pressure to 
perform the test and an allowance for noise level adjustments) are 
no longer needed, their removal will not affect plant operation or 
testing and will not involve an increase in the probability or 
consequences of an accident previously evaluated.
    This proposed TS change will not increase the probability of 
occurrence of a malfunction of any plant equipment important to 
safety. Alternate testing methods at Hope Creek and at the offsite 
test facility adequately demonstrate proper ADS valve operation and 
assure that the valves will continue to function as designed. 
Existing surveillance testing and inspections of the ADS/SRVs at 
Hope Creek verify that the ADS initiation logic, solenoid valve 
operation, pneumatic gas supply integrity and air operator assembly 
(including pilot rod) will operate as designed. Offsite testing 
verifies pilot disc operation, setpoint calibration, stroke time and 
main valve disc operation.
    Deleting the in-situ testing requirement, as proposed, will 
reduce the probability of increasing SRV leakage, which should 
reduce the probability of an inadvertent opening of an SRV. 
Therefore, any SRV pilot leakage that can be eliminated would reduce 
the probability of occurrence of a malfunction of that SRV. Deleting 
the ADS/SRV in-situ functional test will in no way increase any 
consequences of a malfunction of plant equipment important to 
safety. The consequences of a malfunction of an ADS/SRV as discussed 
in the Hope Creek UFSAR remain unchanged.
    In addition, eliminating a known initiator of SRV leakage, as 
proposed in this TS change, would help reduce operator workarounds 
in the form of suppression pool cooling and letdown operation 
activities. As a result, this will reduce the unnecessary operation 
of the Residual Heat Removal (RHR) and its supporting systems.
    Therefore, the proposed TS change 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 TS changes do not involve any physical changes to 
plant SSC. The design and operation of the ADS/SRVs are not changed 
from that currently described in the UFSAR. The ADS will continue to 
function as designed to mitigate the consequences of an accident. No 
changes of any kind are being made to the valves, auxiliary 
components or ADS logic. Deleting the requirement to perform the ADS 
in-situ functional test during plant startup as proposed in this TS 
change request reduces the likelihood of an SRV developing a leak 
and degrading throughout the subsequent operating cycle. Therefore, 
there is no possibility that implementing this proposed TS change 
would create a different type of malfunction to the ADS/SRVs than 
any previously evaluated.
    Eliminating the requirement to perform the in-situ testing of 
the ADS/SRVs during startup activities does not create a new or 
different type of accident than any previously evaluated. There is 
no accident scenario associated with testing the ADS/SRVs other than 
the inadvertent opening of a relief valve, which is currently 
discussed in Section 15.1.4 of the UFSAR. The proposed TS changes do 
not alter the conclusions described in the UFSAR regarding an 
inadvertent opening of an SRV. No new or different type of accident 
will be created as a result of these proposed changes.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Using the provisions of 10CFR50.59, PSE&G will establish a 
method for performing SRV acoustic monitor channel calibration that 
does not require reactor

[[Page 43213]]

steam pressure or SRV opening. This testing method will comply with 
the current TS definition of CHANNEL CALIBRATION. Since the notes 
associated with TS Surveillance Requirement 4.4.2.1 (providing a 
compliance exception to the provisions of TS 4.0.4 to allow for 
proper reactor steam pressure to perform the test and an allowance 
to perform noise level adjustments) are no longer needed, their 
removal will not affect plant operation or testing and will not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.

    The proposed TS change involves deleting the requirement to 
perform in-situ functional testing of the ADS/SRVs during startup 
activities. This testing imposes an unnecessary challenge on the 
ADS/SRVs and has been linked to SRV degradation (e.g., pilot valve 
and/or main valve leakage). This proposed TS change should reduce 
SRV leakage and improve ADS/SRV reliability by reducing the 
potential for spurious SRV actuation. Since ADS operability can be 
readily demonstrated with extremely high confidence by the existing 
surveillance tests and inspections performed for the ADS, there will 
be no reduction in any margin of safety resulting from this proposed 
TS change. Therefore, the proposed TS change does not involve a 
significant reduction in a margin of safety.
    Using the provisions of 10CFR50.59, PSE&G will establish a 
method for performing SRV acoustic monitor channel calibration that 
does not require reactor steam pressure or SRV opening. This testing 
method will comply with the current TS definition of CHANNEL 
CALIBRATION. Since the notes associated with TS Surveillance 
Requirement 4.4.2.1 (providing a compliance exception to the 
provisions of TS 4.0.4 to allow for proper reactor steam pressure to 
perform the test and an allowance to perform noise level 
adjustments) are no longer needed, their removal will not affect 
plant operation or testing and will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: Robert A. Capra.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: February 18, 1998.
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear Plant (WBN) Technical Specifications (TS) 
and associated Bases to address a new condition (Condition B) and 
associated actions in which one train (consisting of two valves) of 
Steam Generator Atmospheric Dump Valves (ADVs), although functional, 
would be considered technically INOPERABLE in the event of one train of 
the auxiliary control air system (ACAS) was out of service. The action 
required for the new condition is to restore the ADV lines to OPERABLE 
status within 72 hours. In addition, the proposed amendment would make 
a correction to the required action for Condition B (new Condition C) 
to clarify that the required action for two or more inoperable ADV 
lines (with the exception of new Condition B) is to restore all but one 
ADV line to operable status. The current Required Action for Condition 
B incorrectly states that only one ADV line must be restored to 
operable status.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The addition of the 72 hour completion time and clarification to 
existing TS do not increase the probability of an accident 
previously evaluated since these changes do not result in hardware 
or procedural changes which will affect probability of occurrence of 
an accident. The probability of an accident occurring during the 72 
hour period as compared to the 24 hour completion time currently in 
the TS remains small. Further, addition of the 72 hour completion 
time and clarification to existing TS does not increase the 
consequences of an accident previously evaluated since sufficient 
equipment and procedures remain available to mitigate accidents 
previously evaluated. With two ADVs inoperable under this LCO, two 
ADVs remain in service. As indicated in the Applicable Safety 
Analysis of the TS Basis, two valves are adequate to cool the unit 
to the RHR [residual heat removal] entry conditions subsequent to 
accidents accompanied by a loss of offsite power. In addition, as 
indicated in the background discussion of the Bases of 3.7.4, the 
ADVs can be operated by use of a bottled nitrogen system designed to 
open the valves in the event of loss of normal and emergency air 
supplies. The valves may also be operated manually by using the 
valve hand wheels. Consequently, the two inoperable ADVs under this 
LCO are still expected to remain functional and could be placed in 
service and used to cool the steam generators, if necessary, in the 
event of an accident. Based on the above, the addition of the 72 
hour completion time and clarifications to existing TS in accordance 
with this proposed amendment do not significantly increase the 
probability or consequences of an accident previously evaluated.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The addition of the 72 hour completion time and clarifications 
to existing TS does not cause the initiation of any accident nor 
create any new credible limiting failure for safety-related systems 
and components. The change does not result in an event previously 
deemed incredible being made credible. As such, it does not create 
the possibility of an accident different than any evaluated in the 
FSAR [Final Safety Analysis Report]. The change has an insignificant 
effect on the ability of the safety-related systems to perform their 
intended safety functions. Although the period during which a 
safety-related function (ACAS air supply) is assumed inoperable is 
extended from 24 to 72 hours, sufficient remaining equipment (two 
ADVs supplied by the opposite train ACAS) is available to mitigate 
the limiting [steam generator tube rupture] SGTR accident, assuming 
no single failure occurs. Also, additional redundant and diverse 
equipment (normal control air, emergency bottled nitrogen, and the 
valve hand wheels) is available and expected to remain functional to 
ensure the ADVs accomplish their function following an accident. The 
change does not create failure modes that could adversely impact 
safety-related equipment. Therefore, the change will not create the 
possibility of a malfunction of equipment important to safety 
different than previously evaluated in the FSAR. Thus, the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The TS currently allow two or more ADVs to be out of service for 
24 hours, based on low probability of an event occurring during the 
period which would require use of the ADVs, and based on 
availability of the steam dump valves and the MSSVs [main steam 
safety valves]. Providing a 72 hour completion time specifically for 
loss of two ADV valves due to loss on one train of ACAS to the ADVs 
does not significantly reduce the margin of safety since the 
probability of an event occurring during the 72 hour period is still 
small, and the capability exists to use the inoperable ADVs by 
manually operating the valves using the valve hand wheels, or by 
connecting the valve nitrogen bottle system, which was designed to 
operate the valves upon loss of air. In addition, the MSSVs, and the 
condenser steam dump valves would normally also be available. Thus, 
the proposed change does not significantly reduce the margin of 
safety.
    Further, the NRC staff notes that the proposed change to the TS 
action statement for two or more ADV lines inoperable to

[[Page 43214]]

require restoration of all but one of the four ADV lines, instead of 
the previous requirement to restore only one ADV line to operable 
status, is more restrictive and more conservative than the action 
statement as currently written. The change also makes the action 
statement consistent with the existing TS Bases in Section B 3.7.4, 
Action B.1. Accordingly, the staff proposes to find that this 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated, 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated, and 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 and the staff's additional assessment as provided above, 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: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: May 6, 1998.
    Description of amendment request: The proposed amendment would 
modify the Watts Bar Nuclear Plant (WBN) Technical Specifications (TSs) 
by revising the allowed enrichment of fuel stored in the new fuel 
storage racks from 4.3 to 5.0 weight percent uranium-235 (U-235). The 
revision also places limitations on fuel storage locations that may be 
utilized in the storage racks and provides additional limits on 
k(effective) when flooded with unborated water.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change to the allowed enrichment of new fuel stored 
in the new fuel storage racks does not change the criticality 
potential with the proposed fuel arrangement requirements for the 
storage racks. The potential keff values are maintained 
the same as the current TS requirements. In addition, the storage 
racks are not modified and the processes for loading and unloading 
fuel in these racks and the controls for these racks remain the same 
except for the storage limitations dictated by the criticality 
analysis. Additional controls are required with appropriate 
verification to assure the fuel is stored within the analysis 
assumptions. Handling procedures contain additional steps to 
specifically verify prohibited cells remain empty after fuel 
movement. This verification assures that the probability of a 
criticality event is not increased by the enrichment change. Since 
the keff limits and operating processes are unchanged by 
the proposed revision, there is no increase in the probability of an 
accident previously evaluated. Likewise, there is no impact to the 
consequences of an accident or increase in offsite dose limits as a 
result of the proposed TS change because the criticality 
requirements are unchanged and plant equipment will be utilized and 
operated without change considering the fuel storage location limits 
imposed by this request.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    As stated above, the plant equipment and operating processes 
will not be altered by the proposed TS change with the exception of 
allowed fuel storage locations in the new fuel storage racks. The 
limitations on acceptable fuel storage locations in the racks ensure 
that the k(effective) limits are maintained at the same limits as 
currently required. TVA has not postulated a criticality event at 
WBN for the spent or new fuel storage locations because the design 
of the associated storage racks, potential moderation, and TS 
allowable fuel enrichments do not support the potential for this 
condition. Therefore, this change does not create the potential for 
a new accident from any previously analyzed.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed TS change maintains the existing requirements for 
criticality by utilizing limited storage locations in the new fuel 
pit storage racks. There is no change to operating practices 
associated with the use and control of these racks except for the 
storage limitations. For these reasons, there will be no reduction 
in the margin [of] the safety as a result of implementing the 
proposed TS change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: July 13, 1998.
    Description of amendment request: The proposed license amendment 
would revise Perry Nuclear Power Plant Technical Specification 3.4.4, 
``Safety/Relief Valves (S/RVs),'' by increasing the present [plus or 
minus] 1% tolerance on the safety mode lift setpoint for the safety/
relief valves to [plus or minus] 3%. This change would be performed in 
accordance with General Electric Topical Report NEDC-31753P, ``BWROG 
In-Service Pressure Relief Technical Specification Revision Licensing 
Topical Report.''
    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 
identified.
    The proposed change allows an increase in the as-found safety 
relief valve (SRV) safety mode setpoint tolerance, determined by 
test after the valves have been removed from service, from [plus or 
minus] 1% to [plus or minus] 3%. The proposed change does not alter 
the Technical Specification requirements on the nominal SRV safety 
mode lift setpoints, the SRV relief mode setpoints, the required 
frequency for the SRV lift setpoint tests, or the number of SRVs 
required to be operable. This change does not involve physical 
changes to the SRVs, nor does it change the operating 
characteristics or safety function of the SRVs.
    Consistent with current requirements, this change continues to 
require that the SRVs be adjusted to within [plus or minus] 1% of 
their nominal lift setpoints following testing. This change does not 
change the behavior and operation of any SRV and therefore has no 
significant impact to reactor operation. It also has no significant 
impact on response to any perturbation of reactor operation 
including transients and accidents previously analyzed in the 
Updated Safety Analysis Report. In addition, this change does not 
change SRV actuation. Therefore, this change will not increase the 
probability of an accident previously evaluated.
    Generic considerations related to the change in setpoint 
tolerance were addressed

[[Page 43215]]

in NEDC-31753P, ``BWROG In-Service Pressure Relief Technical 
Specification Revision Licensing Topical Report,'' and were reviewed 
and approved by the NRC. The plant specific evaluations, required by 
the NRC's Safety Evaluation for NEDC-31753P and performed to support 
this proposed change, are contained in NEDC-32307P, ``Safety Review 
for PNPP Safety/Relief Valve Setpoint Tolerance Relaxation/Out-of-
Service Analyses,'' dated May 1994. These analyses and evaluations 
show that there is adequate margin to the design core thermal limits 
and to the reactor vessel pressure limits using a [plus or minus] 3% 
SRV setpoint tolerance. They also show that operation of the high 
pressure injection systems will not be adversely affected; and the 
containment response from a loss of coolant accident will be 
acceptable.
    (2) The proposed change would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change to allow an increase in the SRV safety mode 
setpoint tolerance from [plus or minus] 1% to [plus or minus] 3% 
does not alter the nominal SRV lift setpoints or the number of SRVs 
required to be operable. This change does not involve physical 
changes to the SRVs, nor does it change the operating 
characteristics or the safety function of the SRVs. The proposed 
change does not involve a physical alteration of the plant. No new 
or different equipment is being installed. The proposed change does 
not impact core reactivity nor the manipulation of fuel bundles. 
There is no alteration to the parameters within which the plant is 
normally operated. As a result no new failure modes are being 
introduced. There are no changes in the methods governing normal 
plant operation, nor are the methods utilized to respond to plant 
transients altered.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) The proposed change will not involve a significant reduction 
in the margin of safety.
    The margin of safety is established through the design of the 
plant structures, systems, and components, the parameters within 
which the plant is operated, and the establishment of the setpoints 
for the actuation of equipment relied upon to respond to an event. 
The proposed change does not significantly impact the condition or 
performance of structures, systems, and components relied upon for 
accident mitigation. The proposed change does not significantly 
impact any safety analysis assumptions or results.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ronald R. Bellamy (Acting).

Previously Published Notices of Consideration of Issuance of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of amendment request: July 8, 1998.
    Description of amendment request: The proposed amendments would 
allow temporary noncompliance with the Penetration Room Ventilation 
System air flow surveillance requirements of Technical Specification 
4.5.4.1.b.1 until modifications can be completed to support testing in 
accordance with ANSI Standard N510-1975, as required by the Technical 
Specifications.
    Date of publication of individual notice in Federal Register: July 
16, 1998 (63 FR 38433).
    Expiration date of individual notice: August 17, 1998.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

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

    Date of application for amendment: June 18, 1998.
    Brief description of amendment: Amend the Crystal River Unit 3 
(CR3) Improved Technical Specifications to allow operation with a 
number of indications previously identified as tube end anomalies and 
multiple tube end anomalies in the CR3 Once Through Steam Generator 
tubes.
    Date of publication of individual notice in the Federal Register: 
June 30, 1998 (63 FR 35615).
    Expiration date of individual notice: July 15, 1998.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.

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

    Date of amendment request: June 19, 1998 (supersedes April 11, 
1997, application), as supplemented July 1, 1998, and information 
provided in a letter of May 5, 1997.
    Brief description of amendment request: The proposed amendment 
would revise Section 3.6.C, Coolant Chemistry, and 3/4.17.B, Control 
Room Emergency Filtration System, of the Technical Specifications (TS), 
Appendix A of the Operating License for the Monticello Nuclear 
Generating Plant. The changes were proposed to establish TS 
requirements consistent with modified analysis inputs used for the 
evaluation of the radiological consequences of the main steam line 
break accident. This amendment request was originally noticed in the 
Federal Register on May 6, 1998 (63 FR 25115). On June 19, 1998, 
supplemented July 1, 1998, the licensee submitted an application that 
superseded in its entirety the licensee's previous submittal dated 
April 11, 1997.
    Date of publication of individual notice in Federal Register: July 
28, 1998 (63 FR 40321).
    Expiration date of individual notice: August 27, 1998.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Date of application for amendment: February 24 1998, as 
supplemented by letter dated May 27, 1998.
    Brief description of amendment: The amendment would support a 
modification to the Callaway Plant, Unit 1 to increase the storage 
capacity of the spent fuel pool.
    Date of individual notice in Federal Register: July 13, 1998 (63 FR 
37598).

[[Page 43216]]

    Expiration date of individual notice: August 12, 1998.
    Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

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 
letter dated May 28, 1998.
    Brief description of amendment: The amendment would support a 
modification to the Wolf Creek Nuclear Generating Station, Unit 1 to 
increase the storage capacity of the spent fuel pool.
    Date of individual notice in Federal Register: July 13, 1998 (63 FR 
37601).
    Expiration date of individual notice: August 12, 1998.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

Notice of Issuance of 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.

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

    Date of application for amendment: February 20, 1998.
    Brief description of amendment: This amendment changed the Pilgrim 
Nuclear Power Station Technical Specification (TS) 3/4.5.B and its 
Bases to incorporate the ultimate heat sink (UHS) temperature of 75 
deg.F, as required by Amendment No. 173. The introduction of a UHS 
temperature restriction requires new specifications, actions, and 
surveillances for the salt service water system. The amendment also 
replaced existing specification 3.5.B ``Containment Cooling System'' 
with new Specification 3/4.5.B.1 ``Residual Heat Removal (RHR) 
Suppression Pool Cooling'', 3/4.5.B.2 ``Residual Heat Removal (RHR) 
Containment Spray'', 3/4.5.B.3 ``Reactor Building Closed Cooling Water 
(RBCCW) System'', and 3/4.5.B.4 ``Salt Service Water (SSW) System and 
Ultimate Heat Sink (UHS)''.
    Date of issuance: July 28, 1998.
    Effective date: July 28, 1998.
    Amendment No.: 176.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17221).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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

    Date of application for amendment: September 19, 1997, as 
supplemented June 15, 1998.
    Brief description of amendment: The amendment relocates the 
Radioactive Effluent Technical Specifications and the Radiological 
Environmental Monitoring Program to the Offsite Dose Calculation 
Manual, in accordance with the recommendations of Generic Letter 89-01. 
Changes are also being made to other sections of the Technical 
Specifications to align them with NUREG-1433, to minimize changes when 
converting to the Improved Standard Technical Specifications.
    Date of issuance: July 31, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 177.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications and the license.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9591).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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

    Date of application for amendment: June 26, 1998, as supplemented 
July 22, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.7.8, ``Ultimate Heat Sink (UHS),'' to permit an 8-
hour delay in the UHS temperature restoration period prior to entering 
the plant shutdown required actions. This TS amendment is given as a 
one-time amendment change effective until September 30, 1998, after 
which the TS will revert back to the original TS provisions.
    Date of issuance: July 29, 1998.
    Effective date: July 29, 1998.
    Amendment No.: 179.
    Facility Operating License No. DPR-23. Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (63 FR 36967 dated July 8, 1998). 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 August 7, 1998, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.

[[Page 43217]]

    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of NSHC are contained in 
a Safety Evaluation dated July 29, 1998.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: P. T. Kuo, Acting.

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

    Date of application for amendments: March 30, 1998.
    Brief description of amendments: The amendments will (1) restore 
Custom Technical Specifications (CTS) and the associated license 
conditions that had been replaced by Improved Technical Specifications 
(ITS), (2) change certain management titles and responsibilities to 
reflect the permanently shutdown condition of the plant, (3) allow use 
of Certified Fuel Handlers in lieu of licensed operators, (4) modify 
shift crew composition, and (5) eliminate verbiage that imples the 
units are operational.
    Date of Issuance: July 24, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 179 & 166.
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25105). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

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

    Date of application for amendment: June 6, 1997, as supplemented 
September 25, 1997.
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) Table 4.1-2, Frequency for Sampling Tests, to 
delete the requirement to sample the spray additive tank and delete the 
requirement for a sodium hydroxide (NaOH) spray additive in TS Section 
5.2.C.1.
    Date of issuance: July 29, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 197.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4310).
    The September 25, 1997, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: March 3, 1998, as supplemented 
by letters dated April 24, May 7, and July 22, 1998.
    Brief description of amendments: The amendments revise Figure 5.1-1 
of the Technical Specifications (TS) to show the new location of the 
meteorological tower. The meteorological tower will be relocated to a 
new location to facilitate use of the current location as a 
construction site. The proposed TS change does not change the related 
TS Section 5.1.1.
    Date of issuance: July 30, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--179; Unit 2--161.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 29, 1998 (63 FR 
35293).
    The July 22, 1998, submittal provided clarifying information that 
did not change the scope of the March 3, 1998, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.

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

    Date of application for amendments: June 19, 1998, as supplemented 
June 23, 1998.
    Brief description of amendments: These amendments revise the BVPS-1 
and BVPS-2 Technical Specifications (TSs) definitions of a channel 
calibration to add two sentences stating that (1) the calibration of 
instrument channels with resistance temperature detector or 
thermocouple sensors may consist of an inplace qualitative assessment 
of sensor behavior and normal calibration of the remaining adjustable 
devices in the channel and (2) whenever a sensing element is replaced, 
the next required channel calibration shall include an inplace cross 
calibration that compares the other sensing elements with the recently 
installed sensing element. This change makes the BVPS-1 and BVPS-2 TS 
definition of channel calibration consistent with the definition of a 
channel calibration contained in the NRC's improved Standard Technical 
Specifications for Westinghouse Plants (NUREG-1431, Revision 1).
    Date of Issuance: July 28, 1998.
    Effective date: Both units, effective immediately, to be 
implemented within 30 days.
    Amendment Nos.: 216 and 93.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 26, 1998 (63 FR 
34939).
    The June 23, 1998, letter provided minor editorial changes to the 
TS pages that did not change the initial proposed no significant 
hazards consideration determination or expand the amendment request 
beyond the scope of the June 26, 1998 Federal Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of application for amendment: March 20, 1998, and supplemented 
May 22, 1998.

[[Page 43218]]

    Brief description of amendment: The amendment proposed to revise 
Improved Technical Specification Safety Limits and Administrative 
Controls to replace the titles of the Senior Vice President, Nuclear 
Operations and the Vice President, Nuclear Production with the position 
of Chief Nuclear Officer.
    Date of issuance: July 20, 1998.
    Effective date: July 20, 1998.
    Amendment No.: 168.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25109).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 20, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: December 29, 1997, as 
supplemented by June 15, 1998.
    Brief description of amendment: The amendment will modify the 
Technical Specifications for selected cycle-specific reactor physics 
parameters to refer to the St. Lucie Unit 2 Core Operating Limits 
Report for limiting values.
    Date of Issuance: July 24, 1998.
    Effective Date: July 24, 1998.
    Amendment No.: 92.
    Facility Operating License No. NPF-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6985).
    The June 15, 1998, supplement provided clarifying information that 
did not change the scope of the December 29, 1997 application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.

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

    Date of amendment request: February 10, 1997, as supplemented 
December 26, 1997, and July 16, and July 28, 1998.
    Brief description of amendment: The amendment revised the Technical 
Specifications to reflect the adoption of the BWR Owner's Group Long-
Term Solution Stability System Option 1-D in addressing reactor 
operation in or near a region of potential thermal hydraulic 
instability.
    Date of issuance: July 29, 1998.
    Effective date: July 29, 1998, to be implemented within 30 days.
    Amendment No.: 177.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14462).
    The December 26, 1997, July 16, and July 28, 1998, submittals 
provided clarifying information and an administrative change that did 
not alter the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.

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: January 15, 1998, as 
supplemented May 29, 1998.
    Brief description of amendments: The amendment allows a reduction 
in the required number of incore instrumentation detectors for the 
remainder of Unit 1, Cycle 19 operation.
    Date of issuance: July 28, 1998.
    Effective date: July 28, 1998, with full implementation within 30 
days.
    Amendment Nos.: 136.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 30, 1998 (63 FR 
4676) The May 29, 1998, supplement provided clarifying information 
within the scope of the Federal Register notice and did not change the 
staff's initial proposed no significant hazards considerations 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: December 12, 1997.
    Brief description of amendment: The amendment revises the working 
hours for operating personnel to allow 8- to 12-hour work days, nominal 
40-hour weeks. In addition, associated changes are being made to 
surveillance intervals to maintain the same frequency.
    Date of issuance: July 24, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 244.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4321).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: March 31, 1997, as supplemented 
June 18, 1997, October 10, 1997, October 20, 1997, November 11, 1997, 
December 22, 1997, January 15, 1998, January 27, 1998, March 30, 1998, 
April 23, 1998, April 27, 1998, May 8, 1998, and May 22, 1998.
    Brief description of amendment: This amendment changes the 
Technical Specifications to accommodate the modification of the spent 
fuel pool by replacing the three Region 1 rack modules with seven new 
borated stainless steel rack modules scheduled for implementation in 
1998. Six new peripheral modules would be added at some future date. 
Two of the seven new modules planned to be installed in 1998 are to be 
designated as part of Region 2, effectively increasing the Region 2 
area. The other five new modules compose Region 1, resulting in a total 
of 294 storage positions in Region 1. Region 2, with 1075 storage 
positions, consists of three rack types, Type 1, Type 2, and Type 4. 
Type 1 cells are the Boraflex cells that form Region 2 for the existing 
license. Two racks of Type 2 cells, containing borated stainless steel 
(BSS) absorber plates are be added to increase

[[Page 43219]]

the storage capacity of Region 2. In addition, the capacity of Region 2 
could be increased in the future by the addition of Type 4 racks, which 
also contain BSS absorber plates. The amendment increases the boron 
concentration from 300 ppm to 2300 ppm.
    Date of issuance: July 30, 1998.
    Effective date: July 30, 1998.
    Amendment No.: 72.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1998 (63 FR 
35617).
    The May 8 and 22, 1998, letters provided clarifying information 
that did not change the proposed no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

Southern Nuclear Power Company, Inc., et al. Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: May 8, 1998.
    Brief description of amendments: The amendments revise VEGP 
Technical Specification 5.5.7, ``Reactor Coolant Pump Flywheel 
Inspection Program,'' to provide an exception to the examination 
requirements of Regulatory Position C.4.b of Regulatory Guide 1.14, 
Revision 1, dated August 1975.
    Date of issuance: July 21, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--103; Unit 2--81.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 17, 1998 (63 FR 
33108).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 21, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

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: February 25, 1998 (TS 97-06).
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) by revising the surveillance requirements 
for the emergency diesel generators.
    Date of issuance: July 22, 1998.
    Effective date: To be implemented no later than 45 days after 
issuance.
    Amendment Nos.: Unit 1--234; Unit 2--224.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17235).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 22, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendments: May 2, 1995, as supplemented 
October 12, 1995, March 26, 1996, December 15, 1997, and May 27, 1998 
(TSCR 172).
    Brief description of amendments: These amendments revise the 
Technical Specifications (TS) Table 15.4.1-1, ``Minimum Frequencies For 
Checks, Calibrations, and Tests Of Instrument Channels,'' to change the 
test frequency of the containment high range radiation monitor, revise 
note 7, and revise item 36 to clarify which monitors in the radiation 
monitoring system support current TS or meet the requirements of 10 CFR 
50.36. In addition several administrative changes to referenced TS 
sections and plant system titles were made to correct omissions from 
previous amendments.
    Date of issuance: July 17, 1998.
    Effective date: July 17, 1998. The TS are to be implemented within 
45 days from the date of issuance. Implementation shall also include 
relocation of certain TS requirements to licensee-controlled documents, 
as described in the licensee's application dated May 2, 1995, as 
supplemented October 12, 1995, March 26, 1996, December 15, 1997, and 
May 27, 1998, and evaluated in the staff's safety evaluation attached 
to these amendments.
    Amendment Nos.: 185 and 189.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25122). The May 27, 1998, submittal provided additional clarifying 
information and updated TS pages. This information was within the scope 
of the original Federal Register notice and did not change the staff's 
initial no significant hazards considerations determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

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

    Date of application for amendments: May 15, 1998 (TSCR 205, NPL-98-
0303).
    Brief description of amendment: This amendment revises the schedule 
for implementing the boron concentration changes from refueling outage 
24 to refueling outage 23 for the planned conversion of Unit 2 to 18-
month fuel cycles.
    Date of issuance: July 21, 1998.
    Effective date: July 21, 1998, with full implementation within 45 
days.
    Amendment No.: 190.
    Facility Operating License No. DPR-27: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 17, 1998 (63 FR 
33111).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 21, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

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

    Date of amendment request: July 17, 1998.
    Brief description of amendment: The amendment revised Technical 
Specification 3/4.7.5, Ultimate Heat Sink, by adding a new Action 
Statement to be used in the event that plant inlet water temperature 
exceeds 90 degrees F.
    Date of issuance: July 18, 1998.
    Effective date: July 18, 1998.
    Amendment No.: 118.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.

[[Page 43220]]

    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated July 
18, 1998.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW, Washington, D.C. 20037.
    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: March 24, 1995, as supplemented by 
letters dated July 26, 1995, and September 5, 1996.
    Brief description of amendment: The amendment adds a new action 
statement to Technical Specification (TS) 3.5.1 which provides a 72-
hour allowed outage time (AOT) for one accumulator to be inoperable 
because its boron concentration did not meet the 2300-2500 parts per 
million band. In addition, TS surveillance requirements are changed to 
incorporate the guidance of Generic Letter 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements for 
Testing During Operation'' that is applicable to the accumulators, and 
the TS Bases section for TS 3/4.5.1 is revised to reflect the changes 
described above. Instrumentation surveillance requirements associated 
with the accumulator are being relocated from the technical 
specifications to Chapter 16 of the Updated Safety Analysis Report.
    Date of issuance: July 21, 1998.
    Effective date: July 21, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 119.
    Facility Operating License No. NPF-42. The amendment revised the 
Operating License and Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18632).
    The July 26, 1995, and September 5, 1996, supplemental letters 
provided additional clarifying information and did not change the 
initial no significant hazards consideration. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 21, 1998.
    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 5th day of August 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-21724 Filed 8-11-98; 8:45 am]
BILLING CODE 7590-01-P