[Federal Register Volume 63, Number 203 (Wednesday, October 21, 1998)]
[Notices]
[Pages 56238-56269]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-28069]


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

NUCLEAR REGULATORY COMMISSION


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

Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from September 26, 1998, through October 8, 1998. 
The last biweekly notice was published on October 7, 1998 (63 FR 
53943).

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period.

[[Page 56239]]

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 November 20, 1998, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: September 23, 1998.

[[Page 56240]]

    Description of amendment request: Carolina Power & Light (CP&L) 
proposes to revise the Harris Nuclear Plant Technical Specification 
(TS) 3/4.6.1.3, ``Containment Air Locks,'' to clarify the requirements 
for locking an air lock door shut. CP&L also proposes to revise TS 3/
4.6.1.3 to be consistent with NUREG 1431, Revision 1, ``Standard 
Technical Specifications, Westinghouse Plants,'' dated April 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    Containment Air Locks are not an accident initiating system as 
described in the Final Safety Analysis Report [FSAR]. The proposed 
change implements guidance for Technical Specifications associated with 
air lock doors consistent with NUREG-1431, Revision 1, ``Standard 
Technical Specifications, Westinghouse Plants,'' dated April 1995. 
Additionally, clarification is provided to permit locking an inoperable 
air lock door as required by Technical Specifications [TS]. The 
proposed change does not affect another Structure, System, or 
Component. The operation and design of containment air locks will not 
be affected by this proposed change. The ability of containment to 
mitigate an accident will not be affected by this change.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    Containment Air Locks are designed to form part of the containment 
pressure boundary. The proposed change provides for administrative 
controls and operating restrictions for air lock doors consistent with 
guidance provided by the Commission. Containment Air Locks are not an 
accident initiating system as described in the Final Safety Analysis 
Report. The proposed change does not affect another Structure, System, 
or Component. The operation and design of containment air locks will 
not be affected by this proposed change.
    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 amendment does not involve a significant reduction 
in the margin of safety.
    The proposed change to containment air locks does not affect any of 
the parameters that relate to the margin of safety as described in the 
Bases of the TS or the FSAR. Accordingly, NRC Acceptance Limits are not 
affected by this change.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Pao-Tsin Kuo (Acting).

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1, Monroe County, Michigan

    Date of amendment request: July 17, 1998 (Reference NRC-98-0044).
    Description of amendment request: The proposed amendment will 
revise the License to allow the licensee to possess special nuclear 
material in a quantity totaling no more than 15 grams of uranium-235, 
uranium-233, or plutonium, or any combination thereof and with 
plutonium totaling no more than 2 curies.
    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 using the standards in 10 CFR 50.92(c). The licensee's 
analysis is presented below:
    (1) Does the proposed change significantly increase the probability 
or consequences of an accident previously evaluated?
    The proposed changes do not involve a significant increase in the 
probability or consequences of an accident. Possessing trace amounts of 
special nuclear material cannot affect the probability of the analyzed 
sodium or liquid waste accidents. The ability to possess such material 
does not itself change any methods of handling liquid waste or sodium. 
Possession of special nuclear material could potentially increase the 
consequences of an accident if it was in use or in the vicinity if an 
accident occurs. However, the increase in consequences would not be 
significant due to the limitations on radioactivity content of such 
special nuclear material. The special nuclear material limit is below 
that requiring an emergency plan or maximum dose evaluation per 10 CFR 
70.22(i). Since the quantity is below that requiring an offsite 
emergency plan or evaluation, even if all the special nuclear material 
allowed to be possessed by the proposed amendment were released during 
a postulated accident, the consequences would not be significantly 
increased. If the provision allowing for possession of more than 15 
grams of special nuclear material or 2 curies of plutonium were to be 
used in the future due to identified plant contamination, the 
requirements of 10 CFR 70.22(i) would need to be assessed and a dose 
evaluation performed or an emergency plan submitted if required to 
ensure the analyzed accident is appropriately addressed and mitigated. 
Any such special nuclear material would be contained in the remaining 
plant contamination, since fuel and blanket material were shipped 
offsite during 1973-1975. Therefore, this amendment does not involve a 
significant increase in the probability or consequences of an accident.
    (2) Will the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously analyzed?
    The proposed changes do not create the possibility of a new or 
different type of accident from any previously evaluated. Allowing 
possession of small amounts of special nuclear material does not change 
methods of monitoring the facility or operations or surveillance of any 
systems at Fermi 1. The amount requested is below that requiring 
criticality monitoring per 10 CFR 70.24, and the separation of the 
special nuclear material will not be permitted. Thus, there is no 
identified physical mechanism for creating an accident based on the 
existence of such material in the quantities specified. If the 
provision allowing for possession of more than 15 grams of special 
nuclear material or 2 curies of plutonium if is identified in plant 
contamination in the future were to be invoked, applicable provisions 
to ensure public safety per 10 CFR Part 70, Part 73, and Part 74 will 
apply. For these reasons, allowing Detroit Edison to possess very 
limited amounts of special nuclear material at Fermi 1 will not create 
the possibility of a new or different type of accident.

[[Page 56241]]

    (3) Will the proposed change significantly reduce the margin of 
safety at the facility?
    The proposed changes do not involve a significant reduction in the 
margin of safety at Fermi 1. No changes to any systems, or the status 
of any systems or structures, are created by this amendment. Being able 
to have a very limited amount of special nuclear material at Fermi 1 
will not significantly reduce the margin of safety because a 10 CFR 
Part 20 program is already in place, and the amount of special nuclear 
material is being limited below criteria requiring an emergency plan, 
special nuclear material control program, or criticality monitoring. If 
more than 15 grams of special nuclear material or 2 curies of plutonium 
is identified in plant contamination in the future, the proposed 
license amendment will require the applicable portions of 10 CFR Part 
70, Part 73, and Part 74 to apply for the amount identified. For these 
reasons, this amendment will not significantly reduce the margin of 
safety at Fermi 1.
    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, NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esquire, Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Branch Chief: John W.N. Hickey.

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

    Date of amendment request: April 30, 1998.
    Description of amendment request: Arkansas Nuclear One--Unit 2 
(ANO-2) Technical Specification (TS) 4.8.1.1.2.c.3 has been revised to 
relocate the specific value for the single largest post-accident load 
to the Bases associated with TS 4.8. The revised TS 4.8.1.1.2.c.3 would 
require the licensee to verify the generator capability to reject a 
load greater than or equal to its associated single largest post-
accident load.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The Diesel Generators (DGs) are not identified as the initiator of 
any accident previously analyzed. The design and function of the DGs 
are unaffected by this proposed change. Applying more restrictive 
acceptance criterion to the single largest load rejection test can not 
result in an increase in the probability of accidents previously 
evaluated and will provide increased assurance that the DGs will 
perform as intended to support the mitigation of accidents previously 
evaluated.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change corrects information contained in the technical 
specification and does not involve any design change, plant 
modification, change in analyzed DG performance, or change in plant 
operation. Since the DGs are not considered to be event initiators, 
their accident mitigation function is unaffected, and normal operation 
is unaffected, the proposed change does not result in new or different 
accidents from those previously analyzed.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The design and function of the DGs are unaffected by the proposed 
change. Applying more restrictive acceptance criterion to the single 
largest load rejection test will provide increased assurance that the 
DGs will perform as intended to support the mitigation of postulated 
accidents. DG performance is proposed to meet a more stringent 
standard.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: May 18, 1998.
    Description of amendment request: The proposed changes delete the 
ANO-2 TS 3.6.2.2 and 4.6.2.2 requirements, and their associated bases, 
for the sodium hydroxide addition system and add new limiting 
conditions for operation, action statements, surveillance requirements, 
and bases information for trisodium phosphate baskets which will be 
installed during the next ANO-2 refueling outage (2R13). The capability 
to add sodium hydroxide to the containment spray system during the 
initial phase of a loss-of-coolant accident will be replaced with 
crystalline trisodium phosphate (TSP) dodecahydrate stored in 
containers located on the floor of the containment building.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change modifies the method of containment spray sump 
pH control. The containment spray function is important for containment 
heat removal/pressure mitigation. However, this change does not affect 
the probability of occurrence of the accident initiators which result 
in the need for containment heat removal and pressure mitigation. Since 
the TSP baskets are seismically mounted passive devices located inside 
the containment, they cannot initiate a transient or affect the 
probability of occurrence of any previously analyzed accident.
    The proposed change only modifies the chemical composition of the 
containment spray and sump fluid. The proposed changes do not affect 
the heat removal/pressure mitigation functions of the system since the 
spray flow rate and droplet size are unchanged. The proposed change 
also will not adversely affect the radiological doses for the design 
basis accident (DBA) loss-of-coolant accident (LOCA) at the exclusion 
area boundary, low

[[Page 56242]]

population zone, control room, or emergency response facility. The 
change does not adversely affect the calculated peak clad temperature 
for the DBA LOCA or the environmental qualification (EQ) of components 
located inside containment.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change allows the use of TSP as a buffering agent for 
the containment sump instead of sodium hydroxide (NaOH) added via the 
containment spray system. The TSP baskets are passive devices that have 
minimal impact on any other system except through water chemistry. The 
change in water chemistry does not adversely affect any safety system 
or required safety functions. The replacement of NaOH additive with TSP 
will not change the probability of a malfunction of safety-related 
equipment.
    Potential malfunctions relating to the proposed modification have 
been evaluated for their effect on plant safety and have been found to 
be non-significant. Additionally, the transient pH behavior of the 
containment spray flow does not adversely affect the EQ of components 
located inside containment.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed change does not adversely affect the ability of the 
containment spray system to perform the functions of containment heat 
removal, pressure mitigation, and fission product (iodine) retention. 
The proposed change does not adversely affect any equipment credited in 
the safety analysis. Also, the proposed change does not increase the 
peak clad temperature or the offsite doses due to the DBA LOCA.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: June 29, 1998.
    Description of amendment request: The proposed amendment would 
revise the as-found lift setting tolerance for the ANO-2 main steam 
safety valves (MSSVs) and pressurizer safety valves (PSVs) will be 
increased. The proposed increase in the lift setting tolerance is 
contingent upon a reduction in a linear power level-high setpoint and 
use of the latest small break loss of coolant accident (SBLOCA) 
methodology for development of the Core Operating Limits Report (COLR).
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This change allows for a larger 3% tolerance versus 
1%, -3% as-found lift setting tolerance. The proposed 
change does not involve any change to the physical characteristics of 
the main steam safety valves (MSSVs) and pressurizer safety valves 
(PSVs), and will have no impact on the as-left settings. During 
testing, the MSSVs and PSVs will continue to adjusted to 1% 
of the Technical Specification (TS) lift setting.
    The impact on the Safety Analysis Report (SAR) analyses when the 
as-found lift setting tolerances are increased has been evaluated and 
the effects upon the impacted events have been found to be within 
acceptable limits, providing the allowable linear power level with 
three inoperable MSSVs is revised from 45% to 36%, and that the latest 
NRC approved C-E small break loss of coolant analysis (LOCA) evaluation 
model, CENPD-137, Supplement 2-P-A, is included as a methodology for 
determination of operating parameters identified within the core 
operating limits report (COLR). With these concurrent changes, plant 
systems required for safe operation and shutdown will continue to be 
available to fulfill their safety function as described in the SAR. 
Steam production in excess of relief capacity is precluded by the 
physical design of the plant and operation of the reactor protection 
system. Revision of the MSSV as-found lift setting tolerance from 
1%, 3% to 3% does not alter safety 
analyses conclusions.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    This change does not create any new plant configuration or 
operational mode. This proposal to increase the MSSV and PSV as-found 
lift setting tolerance does not modify equipment or change the manner 
in which the MSSVs and PSVs will be operated. ASME design requirements 
for maintaining system operating pressure limits below the maximum 
design pressure of 1210 psia for plant secondary systems, and 2750 psia 
for the reactor coolant system (RCS) are not impacted. The reduction in 
allowable linear power level when three MSSVs are inoperable assures 
plant operation within current analysis assumptions. The addition of 
topical report CENPD-137, Supplement 2-P-A, as a reference to develop 
the COLR is bounded by assumptions within the existing safety analysis. 
The cycle specific COLR analyses will continue to be performed 
utilizing NRC approved methodologies. The TS changes do not require any 
new equipment be included in the design basis, and current equipment 
will continue to be operated in a manner consistent with its design.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The upper tolerance limit for design pressure is not affected by 
this change. During the most severe anticipated operational transient, 
the Secondary System pressure and RCS pressure will not exceed 110% of 
design pressure. The MSSV and PSV lift settings will continue to be set 
within -1% of the TS lift setting during surveillance testing.
    The decrease in the peak cladding temperature of the reactor fuel, 
due to a change in the methodology for analysis, does not significantly 
impact previous analytical results. The current and previous analytical 
methodologies are approved by the Staff.
    The impact of the proposed changes on the ANO-2 SAR analyses have 
been evaluated. The evaluation demonstrates that the results of the 
impacted events

[[Page 56243]]

remained within the acceptable limits providing the maximum linear 
power level percentage for three inoperable MSSVs is reduced. This 
reduction in maximum allowable linear power level assures that adequate 
steam relief capacity will be available to prevent overpressurizing the 
secondary steam system during the most severe anticipated operational 
transient.
    Addition of topical report CENPD-137, Supplement 2-P-A, will not 
reduce the existing TS operability and surveillance requirements. The 
cycle specific COLR limits for future reloads will continue to be 
developed based on NRC-approved methodologies. The ANO-2 TSs will 
continue to require that the core be operated within these limits.
    The cumulative impact of all of the proposed changes and the 
results of the impacted events have been found to be within acceptable 
limits. The system capabilities to mitigate and/or prevent accidents 
will be the same as they were prior to these changes.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: June 29, 1998.
    Description of amendment request: These proposed changes are in 
Technical Specification 3.4.2, ``Reactor Coolant System--Safety 
Valves--Shutdown,'' and Technical Specification 3.4.12, ``Reactor 
Coolant System--Overpressure Protection'' regarding the low temperature 
overpressure protection system. The specific changes include modifying 
the requirements for the pressurizer code safety valve requirements 
specified by Technical Specification 3.4.2 and a modification of the 
safety injection tank isolation requirements specified in Technical 
Specification 3.4.12.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The reactor coolant system (RCS) is designed with overpressure 
protection devices to be used in all modes of operation. The changes to 
Technical specification (TS) 3.4.2 will ensure that, if no pressurizer 
code safety valves are operable, the RCS will be cooled down to the 
mode of applicability of the low temperature overpressure protection 
(LTOP) system (TS 3.4.12) within 12 hours. The LTOP relief valves 
provide sufficient relief capacity to protect the RCS from 
overpressurization when the RCS inlet temperature (Tc) less 
than or equal to 220 deg. F. Therefore, this change will ensure the 
proper actions will be taken that will ensure adequate overpressure 
protection of the RCS. These actions are not accident initiators, and 
therefore do not involve a significant increase in the probability of 
any accident previously evaluated.
    The proposed change to TS 3.4.12 provides additional operational 
flexibility for the use of the safety injection tanks (SITs) as an 
additional inventory source during Modes 4, 5, and 6 when the RCS is in 
LTOP conditions. The ability to use the SITs, with a pressure less than 
300 psig is within the existing LTOP analysis. The LTOP analysis 
ensures that under the analyzed worst case overpressurization event, 
the RCS is protected. The 300 psig SIT pressure limit, corrected for 
instrument uncertainty, will prevent a challenge to the LTOP relief 
valves and therefore the RCS will be assured of overpressure 
protection. The SIT pressure limit will also be low enough to prevent 
an inadvertent isolation of the shutdown cooling system and thus 
prevent a loss of shutdown cooling due to placing an SIT in service. 
The remaining changes included in this amendment request are considered 
administrative in nature and are therefore considered acceptable.
    Based on the above discussions, these changes do not involve a 
significant increase in the probability or consequences of any accident 
previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes included in this amendment request provide 
additional operational flexibility for the use of the SITs and specify 
the proper actions to be taken that will ensure adequate overpressure 
protection of the RCS. The LTOP relief valves have already been 
evaluated for operation below 220 deg. F. The changes do not introduce 
any new plant configurations. No new accident possibilities are being 
introduced by these changes. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does Not involve a significant reduction in the margin of 
safety.
    The proposed change to the TS 3.4.2 action statement requires the 
Tc be less than or equal to 220 deg. F when no pressurizer 
code safety valves are available. When Tc is less than or equal to 
220 deg. F, the LTOP system operability is required by TS 3.4.12. This 
action will provide assurance that the RCS will be protected from an 
overpressurization event and therefore increases the margin of safety.
    The requirements to maintain one pressurizer code safety valve in 
Mode 4 when Tc is less than or equal to 220 deg. F and in 
Mode 5 has been removed by the proposed revision to TS 3.4.2. The LTOPs 
provide adequate RCS over pressure protection during these modes 
without reliance on the pressurizer code safeties. Maintaining the 
requirement to require one pressurizer code safety to be operable at 
the same time as the LTOP system is required to be operable, provides 
no additional plant safety. An operable LTOP system prevents RCS 
pressure from increasing high enough to challenge the pressurizer code 
safety lift setpoints.
    The current TS 3.4.12 LTOP limits are based on an analysis that 
uses the methodology outlined in the ASME Code Case N-514. This code 
case defines the margin of safety for the current LTOP limits. This 
code case was utilized in the development of TS 3.4.12. The safety 
factor utilized by the code case provides a reasonable vessel 
overpressure allowance for conditions expected during a low temperature 
transient. The margin of safety is not reduced with SITs in service and 
pressurized to less than 300 psig because this condition is bounded by 
the existing LTOP analysis. Therefore, this change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 56244]]

    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: June 29, 1998.
    Description of amendment request: The proposed change to the 
Arkansas Nuclear One Unit 2 Technical Specifications would provide a 
range of acceptable values for the 4160 Volt bus loss of voltage 
values. The present Technical Specification Table 3.3-4, item 7.a 
provides a single value for both the trip and the allowable values for 
the 4160 Volt bus loss of voltage requirements. These table entries do 
not include an acceptable range or an explicit indication of the 
allowed tolerance that the actual setting is allowed to vary from the 
indicated value. The proposed change replaces the specific trip value 
with an explicit range of acceptable allowable values.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The two 4160 Volt (V) vital bus loss of voltage protection relays 
that are provided on each of the 4160 V safety buses are provided to 
detect loss of voltage, isolate the safety buses, initiate load 
shedding, and start the associated emergency diesel generator. This 
safety function is unchanged by the proposed setpoint revisions. The 
revised settings for the loss of voltage protection relays will 
continue to provide the safety function with no appreciable additional 
time delay. The proposed time delays are within those assumed in the 
ANO-2 safety analyses. Additionally, the lower voltage settings will 
prevent unnecessary isolations from the off-site power sources which 
will contribute to reducing the probability of a loss of off-site power 
due to off-site power system transients.
    The ANO-2 technical specifications will continue to require the 
4160 V loss of voltage functions to be surveillance tested at their 
present frequency without changing the modes in which the surveillance 
is required or the modes of applicability for these components. The 
technical specifications will continue to require the same actions as 
currently exist for the inoperability of one or more of the 4160 V loss 
of voltage channels. Therefore, this change does not involve a 
significant increase in the probability or consequences of any accident 
previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed change introduces no new modes of plant operation or 
new plant configuration. The 4160 V vital bus loss of voltage 
protection relays are required to operate following a complete loss of 
off-site power to initiate the bus power source transfer to on-site 
power, i.e., the emergency diesel generators, to prevent a loss of all 
AC power. This safety function is unchanged by the proposed setpoint 
revisions, and the proposed setpoints continue to provide the required 
actions consistent with the ANO-2 safety analysis. Therefore, this 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The two undervoltage relays located on each 4160 V safety bus are 
provided to detect loss of voltage, isolate the safety buses, initiate 
load shedding, and start the emergency diesel generators. This safety 
function is unchanged by the proposed setpoint revisions.
    The lower loss of voltage values do not affect the safety function 
since there is no appreciable time difference in reaching the lower 
setpoints during a loss of voltage event. The maximum proposed time 
delay setting with the minimum loss of voltage relay setting is within 
those used in the ANO-2 safety analysis. The revised settings for the 
relays will continue to provide the safety function with no appreciable 
additional time delay.
    Removal of the trip value from the technical specifications is 
consistent with that which is presented in NUREG-1432, ``Standard 
Technical Specifications for Combustion Engineering Plants.'' The 
current ANO-2 technical specifications and NUREG-1432 both indicate 
that if the setpoint is outside the allowable value column, the 
associated channel is declared inoperable. This approach is consistent 
with this proposed technical specification change.
    The trip and allowable values listed in the technical 
specifications for the loss of voltage protection for the 4160 V buses 
are presently the same. With these values being the same, if the trip 
value is exceeded, the allowable value will also be exceeded. This 
change provides a range of acceptable allowable values for these 
relays. By relocating the trip values in the surveillance test 
procedures, the procedural limits for the voltage and time delay 
settings can be adjusted to ensure margin to the allowable values. 
Additionally, the lower voltage settings will help to prevent 
unnecessary isolation from the off-site power sources due to off-site 
perturbations in the electrical grid, and thus contribute to increasing 
the margin of safety. Therefore, this change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: June 29, 1998.
    Description of amendment request: The proposed Technical 
Specification change revises the surveillance testing requirements for 
the Arkansas Nuclear One--Unit 2 (ANO-2) direct current (DC) electrical 
distribution system. ANO-2 is planning on modifying the 120 volt vital 
alternating current (AC) electrical distribution system by installing 
new inverters during the next scheduled refueling outage (2R13). This 
modification will increase the normal 125 volt vital DC system loads by 
adding the inverters as a normal load. The power for each 125 volt 
vital DC system is normally supplied by its associated battery charger. 
ANO-2 is in the process of replacing the vital DC battery chargers by 
plant modification to ensure all the battery chargers are of sufficient 
capacity to provide the necessary current requirements for the normal 
125 volt vital DC loads. The proposed change to specification 
4.8.2.3.c.4 is required to ensure the new chargers are adequately 
tested to support the associated inverter replacement.

[[Page 56245]]

    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Technical Specification (TS) surveillance requirement (SR) 
4.8.2.3.b.2 requires the battery banks for each of the vital 125 volt 
direct current (DC) systems to be inspected to ensure that no visible 
corrosion exists at the terminals or the connectors. This SR has been 
modified to allow the present corrosion inspection, or the measurement 
of the resistance of the associated battery connections. The resistance 
measurement provides an indication of physical damage or abnormal 
deterioration that could potentially degrade battery performance and 
has been an accepted alternative to the visual inspection requirement.
    The Bases change associated with TS 3.8.2.3 Action ``b'' is 
considered administrative in nature and simply clarifies the intent of 
the action without changing the requirements of the action or its 
required completion time. The station batteries are not classified as 
accident initiators in the ANO-2 accident analysis. The 125 volt class 
1E batteries are credited for accident mitigation in the accident 
analysis. The above described changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    Each battery charger is required to have sufficient capacity to 
restore the battery from the design minimum charge to its fully charged 
state while supplying normal steady state loads. The minimum specified 
TS surveillance required charger amperage limit will ensure this 
capacity. The additional charger output is presently accounted for in 
the emergency diesel generator loading tables in the Safety Analysis 
Report (SAR). Loss of one train of the vital 125 volt DC system is an 
accident that has been evaluated in the SAR. The capacity of the 
battery chargers is not a factor in the probability of this accident 
occurring. Therefore, the changes associated with this technical 
specification amendment request do not increase the probability of any 
accident previously evaluated.
    The proposed technical specification changes do not modify the 
limiting condition for operation or the associated action statements 
regarding operability of the battery chargers other than clarifying 
these requirements. The frequency at which the battery charger 
operability is demonstrated by surveillance testing is not being 
modified by this technical specification change request. The proposed 
battery charger surveillance testing acceptance criterion will more 
appropriately demonstrate the capability of this equipment. This change 
does not affect the consequences of any of the previously evaluated 
accidents.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    Technical specification SR 4.8.2.3.b.2 requires the battery banks 
for each of the 125 volt systems to be inspected to ensure that no 
visible corrosion exists at the terminals or the connectors. This SR 
has been modified to allow the present corrosion inspection, or to 
perform resistance readings on the associated battery connections. The 
visual inspection is required to detect corrosion of the battery 
connections. The resistance measurement of the associated battery 
connections provides an acceptable alternative to the visual inspection 
requirement and provides an indication of physical damage or abnormal 
deterioration that could potentially degrade battery performance.
    The availability of an extra battery charger for each train 
following the plant modification provides a more reliable configuration 
without introduction of any new modes of plant operation. No new 
accident possibilities are being introduced by the proposed change to 
the surveillance testing specification for battery charger amperage. 
Increasing the surveillance testing amperage limit for the battery 
chargers does not create the potential for any different accident since 
the new value remains within the design capacity of the components.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    TS SR 4.8.2.3.b.2 has been modified to allow resistance readings on 
the associated battery connections or the performance of the present 
visual inspection requirements. The resistance measurement of the 
associated battery connections provides an acceptable alternative to 
the visual inspection requirement and provides an indication of 
physical damage or abnormal deterioration that could potentially 
degrade battery performance without a significant reduction in the 
margin of safety.
    The proposed technical specification surveillance requirements for 
the battery chargers continues to require testing of battery chargers 
at the present duration and frequency. These requirements will also 
apply to the second charger being installed for each Class 1E battery 
train. Each of the new battery chargers has sufficient capacity to 
restore the battery from the design minimum charge to its fully charged 
state while supplying normal steady state loads. The proposed 
surveillance specification change does not involve a significant 
reduction in the margin to safety since the demonstrated capacity will 
be of a higher amperage requirement than is demonstrated during the 
surveillance test with the existing configuration. Increasing the 
required amperage value assures the surveillance test will continue to 
demonstrate the chargers can provide significantly more current than is 
necessary to meet the design requirements. Therefore, this change does 
not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: August 6, 1998.
    Description of amendment request: The proposed technical 
specification change revises the Action requirements for the Arkansas 
Nuclear One--Unit 2 (ANO-2) Control Element Assembly (CEA) position 
indicator channels. The Action requirements listed in Specification 
3.1.3.2 are being modified consistent with the requirements of NUREG-
1432, ``Standard Technical Specifications for Combustion Engineering 
Plants.'' The proposed changes also include the relocation of Technical 
Specification Table 3.8-1, ``Containment Penetration Conductor 
Overcurrent Protective Devices'' per

[[Page 56246]]

NRC Generic Letter 91-08, ``Removal of Component Lists From Technical 
Specifications.''
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This technical specification (TS) change request contains the 
relocation of Table 3.8-1, Containment Penetration Conductor 
Overcurrent Protective Devices, and changes to the control element 
assembly (CEA) position indication.
    Generic Letter (GL) 91-08, ``Removal of Component Lists From 
Technical Specifications,'' was issued as a TS line item improvement by 
the NRC. Table 3.8-1 is one of the specific lists of components 
contained in the GL. TS Table 3.8-1 and all its references have been 
removed from Specification 3/4.8.2.5 in accordance with the GL. This 
change is considered administrative in nature because the requirements 
for operability, the limiting conditions for operation, the 
surveillance requirements and their frequencies for the containment 
penetration conductor overcurrent protective devices remains the same. 
This amendment request fundamentally modifies the physical location of 
the devices listed in Table 3.8-1 from the TS to the plant procedures. 
These changes have no affect on the probability or consequences of any 
accident previously evaluated.
    The remaining changes included in this amendment request are those 
relating to the CEA position indication. The Action requirements for TS 
3.1.3.2 were modified to be consistent with the requirements of NUREG-
1432, ``Standard Technical Specifications for Combustion Engineering 
Plants.'' The most recent revision of NUREG-1432 was used to produce 
this change because it represents the latest guidance for the TS CEA 
position indication requirements that are applicable to ANO-2 and 
acceptable to the NRC.
    The requirement was removed from TS 3.1.3.2 that restricted each 
CEA group to a maximum of one CEA with less than two of the required 
position indicator channels. NUREG-1432 places no requirements on the 
number of CEAs in a group with less than two of the required position 
indicator channels. NUREG-1432 would allow all the CEAs in a group to 
have only one of the required CEA position indications operable. In 
this situation, the associated CEAs with less than two of the required 
position indicator channels would have to be placed at their ``Full 
In'' or ``Full Out'' limits.
    TS 3.1.3.2 was modified to allow the use of the ``Full In'' or 
``Full Out'' limits which ensures this specification is consistent with 
its bases and NUREG-1432. The TS will still maintain the requirements 
for two independent means of determining CEA position with this 
amendment request. With two independent means of determining CEA 
position, reliable determination of actual CEA position will be 
maintained.
    Additionally, NUREG-1432 does not require the placement of any 
other CEAs in the associated group at the ``Full Out'' limit when one 
of the CEAs in the group has only one of the required position 
indication systems operable. All of the remaining CEAs in the 
associated group still have at least two independent means of CEA 
position indication or they would already be required to be positioned 
to the ``Full Out'' limit to restore the second position indication. 
The TS retains the requirements for the individual and group CEA 
alignment in accordance with Specifications 3.1.3.1 and 3.1.3.6. These 
requirements also eliminate the need for pulling the remaining CEAs in 
the group to the ``Full Out'' limit as long as the alignment 
requirements are maintained.
    These changes will allow the operator more time to focus on the 
individual CEA position indication problem rather than moving the 
remainder of the CEAs in the group unnecessarily. Anytime that a CEA is 
moved, a small probability exists for it to slip or drop into the core. 
If this were to occur while attempting to align the group to the ``Full 
Out'' limit, a reactor transient would be initiated. Additionally, 
anytime the CEAs are operated, a small probability of an error exists. 
Removing the unnecessary requirement for the group withdrawal could 
decrease the probability of CEA misoperation. CEA position indication 
is not considered as an accident initiator. Retaining the requirements 
to maintain at least two independent means of determining CEA position 
will ensure the consequences of all the accidents previously evaluated 
remain unchanged.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The portions of this change that are made in accordance with GL 91-
08 are considered administrative in nature and do not result in the 
creation of a new or different kind of accident from any previously 
evaluated.
    The bases for TS 3.1.3.2 state that the action statements 
applicable to inoperable CEA position indicators permit continued 
operation when the positions of CEAs with inoperable position 
indicators can be verified by the ``Full In'' or ``Full Out'' limits. 
Although TS 3.1.3.2 may have originally been intended to allow 
continued operation using the ``Full In'' limits, it has never been 
clearly addressed in the specification. NUREG-1432 allows the use of 
both the ``Full In'' or ``Full Out'' limits. This amendment request 
will not change the methods for CEA operation, although it will reduce 
unnecessary CEA manipulations due to CEA position indication problems.
    The requirements of Specification 3.1.3.1 will ensure that an 
individual CEA is maintained in proper alignment with the remaining 
CEAs in the group. Specification 3.1.3.6 will ensure the CEA groups are 
maintained within the proper withdrawal sequence and insertion limits. 
Specification 3.1.3.5 will ensure the shutdown CEA groups are 
maintained in the ``Full Out'' position. The CEA position indication 
changes allowed by this amendment request, including the allowance to 
use the ``Full In'' limits, can produce a CEA configuration that is 
different from that allowed by the current TSs. However, the allowed 
configurations will be bounded by the TS 3.1.3.2 Action ``c'' 
requirements for compliance with Specifications 3.1.3.1, 3.1.3.5, and 
3.1.3.6. Therefore, the action requirements of TS 3.1.3.2 will ensure 
the CEAs are operated consistent with the safety analysis assumptions.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The portions of this change that are made in accordance with GL 91-
08 are considered administrative in nature and have no effect on the 
margin of safety. The remaining changes can result in a lower 
probability of CEA misoperation and reduce the potential of plant 
transients due to CEAs that slip or drop into the core while performing 
unnecessary group realignments. These changes can also reduce 
unnecessary plant shutdowns, due to unneeded restrictions on CEA 
position indication. An unnecessary plant shutdown produces an 
opportunity for plant

[[Page 56247]]

upsets that can be avoided by this change. The proposed TS provide an 
equivalent level of safety as those specifications that currently 
exist. Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: September 17, 1998.
    Description of amendment request: The proposed amendment addresses 
a problem associated with the existing technical specifications being 
inconsistent with the design of the plant protection system (PPS). The 
PPS uses a design in which a single bistable is used to automatically 
enable the selected core protection calculator (CPC) trip functions 
whenever a permissive exists to bypass the high logarithmic power level 
trip function. The technical specifications allow the bypass of the 
high logarithmic power trip when power is above 10-4 percent 
power and allow bypasses of the affected CPC trips when power is below 
10-4 percent power. The proposed technical specification 
change establishes a range for the bistable setpoint to be within such 
that it is possible to meet both of its design functions while also 
meeting the technical specification 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. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This technical specification (TS) change request modifies the power 
level at which two of the three operating bypasses can be set to 
operate. This change is necessary because the present plant bistable 
design requires a range for this bistable to operate within rather than 
a specific setpoint as required by the present TS. The single bistable 
associated with these operating bypasses is designed with an inherent 
hysteresis loop and therefore requires an operating range. The band of 
10-4% to 10-2% of rated thermal power provides 
the bistable an adequate operating range to account for the inherent 
bistable hysteresis, allow for bistable drift, and provides margin for 
the applicable uncertainties. Regardless of the actual bistable 
setpoint within this band, the bistable design ensures that either the 
high logarithmic power level or the core protection calculator (CPC) 
generated trips are available to provide reactor trip protection. The 
CPC and logarithmic power operating bypasses and their setpoints are 
not considered credible accident initiators and therefore modifying 
their setpoints does not involve a significant increase in the 
probability of an accident previously evaluated.
    The automatic removal function of these operating bypasses is 
designed to mitigate the consequences of accidents. As described within 
the background section of the TS change request, the safety analyses 
associated with operating bypasses have been reviewed for the 
acceptability of these changes. This review concluded that these 
changes are considered bounded by the existing safety analyses. Since 
these TS changes are bounded within the present safety analyses, they 
do not involve a significant increase in the consequences of an 
accident previously evaluated.
    The remaining changes included in this TS change request are being 
made to clarify the existing requirements for the operating bypasses 
and to establish consistency with the above described changes. The 
remaining changes have been found acceptable because they are 
considered administrative in nature and have no effect on the 
probability or consequences of an accident previously evaluated.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    There are no physical plant modifications being made to the plant 
as a result of this change. The only function that is required by the 
TS and modified by this change is associated with the allowed setpoint 
for the automatic bypass removal feature of the CPCs. This feature will 
still be required by the TS, but will be allowed a slightly higher 
setpoint. The system connections and the reactor trip setpoints are not 
affected by this change. The CPC and logarithmic power operating 
bypasses and their setpoints are not considered as credible accident 
initiators. Therefore, this change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The safety analyses associated with these operating bypasses have 
been reviewed for the acceptability of these changes. This review 
concluded that the changes associated with this TS change request are 
considered bounded within the existing safety analyses. The associated 
safety analyses have been considered to be acceptable because they have 
produced acceptable results and thus provide an acceptable margin to 
safety. Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: June 29, 1998.
    Description of amendment request: The proposed changes modify 
Technical Specification (TS) 3.7.6.1 (Control Room Emergency Air 
Filtration System--Modes 1-4), TS 3.7.6.2 (Control Room Emergency Air 
Filtration System--Modes 5 and 6), TS 3.7.6.3 (Control Room Air 
Temperature--Modes 1-4), TS 3.7.6.4 (Control Room Air Temperature--
Modes 5 & 6), and TS 3.7.6.5 (Control Room Isolation and 
Pressurization), and the associated Bases.
    The proposed changes to the control room ventilation TS affects the 
Applicability and the Actions. These changes will make the TS 
consistent with NUREG-1432 (Standard Technical Specifications 
Combustion Engineering

[[Page 56248]]

Plants), as applicable, and the accident analysis. The proposed changes 
to the TS Bases make the Bases consistent with the TS and also clarify 
that suspending movement of irradiated fuel assemblies shall not 
preclude movement to a safe conservative position.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise the control room ventilation Technical 
Specifications (TS) Actions to delete the Action statement to suspend 
all operations involving positive reactivity changes, and adds an 
Applicability and Action related to the movement of irradiated fuel 
assemblies. The changes also add an Applicability footnote and revise 
the Bases to allow irradiated fuel assemblies to be placed in a safe 
conservative position when movement is required to be suspended. Other 
changes to the Bases are being made to be consistent with the TS. These 
changes do not affect the probability of an accident. The control room 
ventilation systems (ventilation, temperature, or envelope) do not 
affect the initiators of an accident; therefore, the changes do not 
alter the initiators of any analyzed events.
    The administrative and more restrictive changes do not affect the 
consequences of an accident. The administrative changes add an 
Applicability footnote and revise the TS Bases to make them consistent 
with the TS. This will ensure the applicable control room ventilation 
system TS are entered during movement of irradiated fuel assemblies and 
that there is no confusion associated with the Bases being 
inconsistent. The more restrictive change of adding the Applicability 
during movement of irradiated fuel assemblies and the Action to suspend 
movement of irradiated fuel assemblies eliminates the precursor to the 
fuel handling accident which prevents the fuel handling accident from 
occurring when the control room ventilation systems are inoperable. The 
addition of this Action ensures the event that may release 
radioactivity is precluded when the control room ventilation systems 
are inoperable.
    The less restrictive changes (deleting the requirement to suspend 
positive reactivity changes and a Bases change which allows irradiated 
fuel assemblies to be placed in a safe conservative position when 
movement has been suspended) do not affect the consequences of an 
accident because no accident mitigator is affected. The safety analysis 
credits instrumentation to detect a boron dilution accident and alert 
the control room staff. After the control room staff is alerted, the 
accident is terminated without a radioactive consequence. These 
instruments are required to be Operable and if one is inoperable, 
positive reactivity changes are required to be suspended. If both 
instruments become inoperable, along with suspension of positive 
reactivity additions, boron concentration is required to be determined 
at frequencies specified in the Core Operating Limits Report (only when 
source range neutron flux monitors are inoperable). Also, the shutdown 
margin (SDM) is required to be met. If the SDM requirements are not 
met, action must be taken to borate (addition of negative reactivity) 
until the SDM is restored. Therefore, if the control room ventilation 
systems are inoperable, suspension of positive reactivity changes are 
not required. The added statement in the Bases allows irradiated fuel 
assemblies to be placed in a safe conservative position to preclude a 
fuel handling accident from occurring. These Actions ensure that 
appropriate measures are taken to preclude events that would require 
the control room to be isolated when any of the control room 
ventilation systems are inoperable.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this proposed 
change create the possibility of a new or different type of accident 
from any accident previously evaluated?
    Response: No.
    The proposed changes revise the control room ventilation TS Actions 
to delete the Action statement to suspend all operations involving 
positive reactivity changes, and adds an Applicability and Action 
related to the movement of irradiated fuel assemblies. The changes also 
add an Applicability footnote and revise the Bases to allow irradiated 
fuel assemblies to be placed in a safe conservative position when 
movement is required to be suspended. Other changes to the Bases are 
being made to be consistent with the TS. These changes do not alter the 
design or configuration of the plant. There has been no physical change 
to plant systems, structures, or components. The proposed changes will 
not reduce the ability of any of the safety-related equipment required 
to mitigate Anticipated Operational Occurrences (AOOs) or accidents. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this proposed 
change involve a significant reduction in a margin of safety?
    Response: No.
    The proposed changes revise the control room TS Actions to delete 
the Action statement to suspend all operations involving positive 
reactivity changes, and adds an Applicability and Action related to the 
movement of irradiated fuel assemblies. The changes also add an 
Applicability footnote and revise the Bases to allow irradiated fuel 
assemblies to be placed in a safe conservative position when movement 
is required to be suspended. Other changes to the Bases are being made 
to be consistent with the TS. The margin of safety is not affected 
because the proposed changes to delete one Action and add an 
Applicability and Action ensures the assumptions of the accident 
analysis are being met. The administrative changes ensure the 
applicable TS are entered and eliminate confusion associated with the 
discrepancies between the TS and Bases. The more restrictive changes of 
adding an Applicability and Action eliminates the precursor to an event 
(fuel handling accident) that may release radioactivity when the 
control room ventilation systems are inoperable. The less restrictive 
changes revises the TS to rely on the instrumentation credited in the 
accident analysis and to allow irradiated fuel assemblies to be placed 
in a safe position to preclude a fuel handling accident. The 
instruments are required to be operable per TS. Compliance with these 
TS and also the SDM TS ensures that boron dilution event is precluded 
or can be mitigated. Therefore, suspension of positive reactivity 
changes is not required when the control room ventilation systems are 
inoperable. These Actions ensure that appropriate measures are taken to 
preclude events that would require the control room to be isolated when 
any of the control room ventilation systems are inoperable. Therefore, 
the proposed 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

[[Page 56249]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: August 12, 1998
    Description of amendment request: The proposed amendment will 
change Technical Specifications (TS) 3.1.2.8, 3.5.1, 3.5.4, Figure 3.1-
1, and Bases 3/4.5.2 for Waterford 3. It increases the maximum boron 
concentration in the Safety Injection Tanks (SITs) and the Refueling 
Water Storage Pool (RWSP) from 2300 ppm to 2900 ppm.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this proposed 
change involve a significant increase in the probability or consequence 
of any accident previously evaluated?
    Response: No.
    The proposed change increases the maximum boron concentration in 
the SITs and the RWSP from 2300 ppm to 2900 ppm. This change does not 
affect the probability of any accident. This increase in boron 
concentration affects the pH of water in the safety injection sump 
during a LOCA [Loss of Coolant Accident] and the potential for boron 
precipitation. The amount of TSP in containment is adequate to maintain 
the pH above 7.0. The revised long term cooling analysis shows that 
boron precipitation will not occur at the higher boron concentrations. 
Therefore, this change will not adversely impact post-LOCA core 
cooling. Thus, the consequences of a LOCA are not affected.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequence of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    Response: No.
    The proposed change will not create any new system connection or 
interactions. Thus, no new modes of failure are introduced. There is no 
significant impact on the corrosion rate in the safety injection system 
due to the slightly higher acidic solution with the higher boron 
concentration.
    Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this proposed 
change involve a significant reduction in margin of safety?
    Response: No.
    Sufficient TSP [Trisodium Phosphate Dodecahydrate] is provided in 
the containment to ensure that the pH of the safety injection sump 
water during a LOCA remains above 7.0 as stated in the Technical 
Specification bases. Adequate time and HPSI [High Pressure Safety 
Injection] flow exist to avoid boron precipitation during a LOCA. The 
higher boron concentration limit will also allow higher refueling boron 
concentrations which will increase the available shutdown margin.
    Therefore, the proposed change does not involve a significant 
reduction in margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: John N. Hannon

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

    Date of amendment request: August 31, 1998.
    Description of amendment request: The proposed amendment would 
revise Improved Technical Specification (ITS) 5.6.2.10, ``Steam 
Generator (OTSG [once-through steam generator]) Tube Surveillance 
Program,'' to include a new repair process, called a ``repair roll'' or 
``re-roll.'' The process would be used to repair steam generator tubes 
with defects within the upper tubesheet. Changes to inservice 
inspection and reporting requirements are proposed for tubes which are 
repaired using this process. The proposed revision would also require 
inspection of both OTSGs during each inservice inspection. In addition, 
several format and editorial changes are proposed to ITS 5.6.2.10 and 
to ITS 5.7.2, ``Special Reports,'' for clarification purposes.
    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 LAR [license amendment request] addresses several 
editorial and format changes which do not impact accident analyses. LAR 
#235 also proposes to implement the repair roll (re-roll) process.
    The qualification of the re-roll joint is based on establishing a 
mechanical roll length which will carry all structural loads imposed on 
the tubes with required margins. A series of tests and analyses were 
performed to establish this length. Tests that were performed included 
leak, tensile, fatigue, ultimate load and eddy current measurement 
uncertainty. The analyses evaluated plant operating and faulted loads 
in addition to tubesheet bow effects. Any tube leakage will be bounded 
by the main steam line break (MSLB) evaluation presented in the Final 
Safety Analysis Report (FSAR). The proposed change also requires 
inspections of the joints created by the repair roll process. The 
addition of this inspection does not change any accident initiators. 
The proposed inspections after re-roll installation, and during future 
inservice inspections, assure continuous monitoring of these tubes such 
that inservice degradation of tubes repaired by the re-roll process 
will be detected. Based on the Framatome Technologies qualification, as 
well as the history for similar industry repair rolls, there are no new 
safety issues, as defined in BAW-2303P, Revision 3, associated with the 
repair roll. Therefore, this change does not involve a significant 
increase in the probability or consequences of any accident previously 
evaluated.
    (2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No new failure modes or accident scenarios are created by the re-
roll process. The new pressure boundary joint created by the repair 
roll process

[[Page 56250]]

has been shown by testing and analysis to provide structural and 
leakage integrity equivalent to the original design and construction 
for all normal operating and accident conditions. Furthermore, the 
testing and analysis demonstrate the repair roll process creates no new 
adverse effects for the repaired tube and does not change the design or 
operating characteristics of the OTSGs. In the unlikely event that a 
tube with a repair roll should fail and sever completely at the 
transition of the re-roll region, the tube would remain engaged in the 
tubesheet bore, preventing interaction with other surrounding tubes. In 
this case, leakage is bounded by the steam generator tube rupture 
(SGTR) accident analysis. Therefore, this change does not create a 
possibility of a new or different kind of accident from any previously 
evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The repair roll process effectively removes the defective/degraded 
area of the tube from service. The new roll expanded interface created 
with the tubesheet satisfies all the necessary structural, leakage and 
heat transfer requirements. The joint is constrained within the 
tubesheet bore; thus, there is no additional risk associated with tube 
rupture. The accident leakage is shown to be well within the initial 
assumption of the MSLB analysis of one gallon per minute primary-to-
secondary leakage. Therefore, the FSAR analyzed accident scenarios 
remain bounding, and the use of the repair roll process does 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 50.92(c) are 
satisfied.
    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC--A5A, P.O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: August 31, 1998.
    Description of amendment request: The proposed amendment will 
change the Improved Technical Specifications (ITS) to add three 
additional Regulatory Guide (RG) 1.97 Type A Category 1 post-accident 
monitoring (PAM) instrumentation variables and one Type B Category 1 
PAM instrumentation variable to ITS Table 3.3.17-1, Post-Accident 
Monitoring Instrumentation. The Type A Category 1 variables added are 
low pressure injection (LPI) pump run status, LPI suction from reactor 
building (RB) sump isolation valves DHV-42 and DHV-43 open position, 
and high pressure injection (HPI) pump run status. The Type B Category 
1 variable added is reactor coolant system (RCS) low range pressure.
    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 addition of post-accident monitoring instrumentation to the CR-
3 ITS and ITS Bases is to ensure instrumentation is available for use 
by the operators for performing manual actions, or to verify automatic 
actions have occurred, which are required to mitigate the effects of a 
design basis accident. The instrumentation is used for monitoring by 
the operators only after an accident occurs, performs no automatic 
functions, and there are no credible failures of this instrumentation 
which could initiate any accident previously evaluated. Therefore, the 
probability of occurrence of any accident previously evaluated is 
unaffected.
    The availability and use of this instrumentation ensures that the 
prescribed manual operator actions for mitigating the consequences of 
an accident will be implemented when necessary, and that the operator 
has sufficient information to verify required automatic actions have 
occurred when necessary. Therefore, the availability and use of the 
instrumentation provides assurance that the consequences of accidents 
will not be greater than that previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from previously evaluated accidents?
    The addition of post-accident monitoring instrumentation to the CR-
3 ITS and ITS Bases is to ensure instrumentation is available for use 
by the operators for performing manual actions, or to verify automatic 
actions have occurred, which are required to mitigate the effects of a 
design basis accident. The instrumentation is used for monitoring by 
the operators only after an accident occurs, performs no automatic 
functions, and there are no credible failures of this instrumentation 
which could initiate a new or different kind of accident. Therefore, 
the possibility of a new or different kind of accident occurring as a 
result of this passive instrumentation is not created.
    3. Involve a significant reduction in a margin of safety?
    The addition of post-accident monitoring instrumentation to the CR-
3 ITS and ITS Bases is to ensure instrumentation is available for use 
by the operators for performing manual actions, or to verify automatic 
actions have occurred, which are required to mitigate the effects of a 
design basis accident. The instrumentation is used for monitoring by 
the operators only after an accident occurs, and performs no automatic 
functions. The availability and use of this instrumentation ensures 
that the prescribed manual operator actions for mitigating the 
consequences of an accident will be implemented when necessary, and 
that the operator has sufficient information to verify required 
automatic actions have occurred when necessary. These required manual 
and automatic actions are necessary to preserve the margin of safety as 
defined in the CR-3 ITS and ITS Bases. The availability and use of this 
instrumentation provides assurance that the existing margin of safety 
will be maintained, and assumptions related to the margin of safety 
during mitigation of design basis accidents will be preserved. 
Therefore, the existing margin of safety will not be reduced.
    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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.
    GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey.
    Date of amendment request: May 5, 1998.

[[Page 56251]]

    Description of amendment request: This request is to change the 
licensing basis to allow for a small amount of containment overpressure 
to ensure sufficient net positive suction head for the Emergency Core 
Cooling System pumps under post Loss of Cooling Accident (LOCA) 
conditions.
    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 to the licensing basis does not ``Involve a 
significant increase in the probability or consequences of an accident 
previously evaluated * * * ''. As the strainers have no function until 
after the design basis LOCA occurs, the design of the strainer cannot 
affect the probability of a Large Break LOCA.
    The requested change to raise the assumed containment overpressure 
for suction strainer design to 1.25 psig is less than that which is 
already used in LOCA analyses for offsite releases. Therefore, this 
change will not increase the offsite consequences of any previously 
analyzed accident. The frequency of a design basis LOCA occurrence at 
the Oyster Creek Nuclear Generating Station is conservatively estimated 
at 5.67  x  10-4 per year. The frequency of a design basis 
LOCA with a loss of containment overpressure is conservatively 
estimated at 2.46  x  10-7 per year.
    Since the frequency of the design basis LOCA coincident with a loss 
of containment overpressure is insignificant (2.46  x  
10-7), the requested increase does not significantly impact 
the probability of exceeding the existing design bases. The core damage 
frequency increase due to the request for overpressure is mitigated, in 
part, by the current procedural requirement to flood containment 
following the design basis LOCA, thereby obviating the need for over 
pressure in the long term. The risk evaluation, performed in support of 
the request for over pressure, indicated a non-risk significant change 
in the core damage frequency.
    The proposed change to the licensing bases does not ``Create the 
possibility of a new or different kind of accident from any accident 
previously evaluated * * *''. Both the new and existing strainers are 
passive. They function solely to prevent debris from entering the 
suction of the core and containment spray pumps. The only significant 
difference is that the new strainers can remove more debris without 
clogging. The slight amount of containment overpressure does not affect 
the operation of the strainers, and improves the ability of the core 
spray and containment spray systems to continue operation. Therefore, 
no new or different kind of accident is created or possible.
    The proposed change to the licensing bases does not ``Involve a 
significant reduction in a margin of safety * * *.'' The modification 
increases the amount of debris that can be removed while maintaining 
core spray system operation. The requested change takes credit for 1.25 
psig of wetwell overpressure. However, as the requested change is 
bounded by existing calculations for offsite release, no significant 
reduction in the margin of safety can occur. Additionally, as 
demonstrated in Attachment III, the probability of a LOCA with a loss 
of containment overpressure is not significant.
    Guidance has been provided in ``Final Procedures and Standards on 
No Significant Hazards Considerations,'' Final Rule, 51 FR 7744, for 
the application of standards to license change requests for 
determination of the existence of significant hazards considerations. 
This document provided examples of amendments which are and are not 
considered likely to involve significant hazards considerations.
    Based on the above evaluation and the review of 51 FR 7744, this 
proposed change to the licensing basis of the Oyster Creek Nuclear 
Generating Station does not involve irreversible changes, a significant 
relaxation of the criteria used to establish safety limits, a 
significant relaxation of the bases for the limiting safety system 
settings, or a significant relaxation of the bases for the limiting 
conditions for operations. Therefore, based on the guidance provided in 
the Federal Register and the criteria established in 10 CFR 50.92(c), 
the proposed change does not constitute a significant hazard.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for Licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: 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: September 9, 1998.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) by: (1) Changing the TS 
Definitions 1.24, ``Core Operating Limits Report,'' 1.27, ``Engineering 
Safety Feature Response Time,'' and 1.31, ``Radiological Effluent 
Monitoring and Offsite Dose Calculation Manual (REMODCM)''; (2) 
changing TS 3.0.2, ``Limiting Condition For Operation,'' by adding a 
new TS 3.0.6 to the Limiting Condition For Operation TS section; (3) 
changing TS 4.0.5, ``Surveillance Requirements''; (4) changing the mode 
applicability of TS 3.2.3, ``Total Unrodded Integrated Radial Peaking 
Factor--FrT''; (5) changing TS 3.3.2.1, 
``Engineered Safety Features Actuation System Instrumentation,'' by 
modifying TS Table 4.3-2 Table Notation (1) which it references; (6) 
changing TS 3.4.1.1, ``Reactor Coolant System--Coolant Loops and 
Coolant Circulation Startup and Power Operation'; and (7) changing TS 
3.4.11, ``Reactor Coolant System--Reactor Coolant System Vents.'' The 
associated TS Bases sections would also be updated to reflect the 
proposed changes. The proposed changes would resolve identified 
compliance issues.
    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.
Technical Specification Definitions
    The minor editorial and non-technical changes to correct reference, 
spelling and terminology errors contained in the definitions will not 
result in any technical changes to the Millstone Unit No. 2 Technical 
Specifications. The proposed changes will have no adverse effect on 
plant operation. Therefore, the proposed change will not result in a 
significant increase in the probability or consequences of an accident 
previously evaluated.
Technical Specification 3.0.6
    The new Technical Specification, 3.0.6, will provide guidance on 
returning inoperable equipment to service, under administrative 
control, to demonstrate operability of that

[[Page 56252]]

equipment, or the operability or other equipment. Various Technical 
Specification Actions require inoperable equipment to be removed from 
service, such as maintaining a containment isolation valve closed or 
tripping/bypassing a failed instrument channel. An exception to these 
required actions is necessary to allow the performance of testing to 
demonstrate the operability of the equipment being returned to service. 
Specifically, this Technical Specification addresses the situation 
where the inoperable equipment has been repaired, tested to the extent 
possible, and believed to be capable of performing its function. At 
this point, a presumption of the operability of the equipment is 
reasonable, and is supported by experience. Therefore, it is acceptable 
to place the equipment in service for testing under administrative 
control. Administrative controls will be used to ensure the time the 
equipment is returned to service is consistent with the Action 
Statements and is limited to the time necessary to perform the 
surveillance requirements.
    This specification will also allow the inoperable equipment to be 
placed in a condition different from that required by the action 
statement to demonstrate the operability of other equipment. An example 
would be during the performance of an operability test on one reactor 
protection channel while another channel associated with the same 
function is inoperable. In this situation only one of the channels 
could be in the tripped condition, otherwise a reactor trip would be 
initiated. This is already permitted for reactor protection channels by 
Technical Specifications 3.3.1.1, ``Instrumentation--Reactor Protective 
Instrumentation,'' Action 2, and for engineered safety features 
channels by 3.3.2.1, ``Instrumentation--Engineered Safety Feature 
Actuation System Instrumentation,'' Action 2.
    This provision is provided only to perform surveillance 
requirements to prove operability, and not to provide time to perform 
any other preventive or corrective maintenance. The testing will be 
performed consistent with the current Technical Specification Action 
Statement and will be limited to the time necessary to perform the 
surveillance requirement. The proposed changes will have no adverse 
effect on plant operations. Therefore, the proposed change will not 
result in a significant increase in the probability or consequences of 
an accident previously evaluated.
Technical Specification 4.0.5
    The proposed changes will revise Technical Specification 4.0.5.a 
and Bases 3/4.4.10, ``Structural Integrity,'' by removing the phrase 
``(g), except where specific written relief has been granted by the 
Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).'' The 
changes to Technical Specifications clarify that all applicable 
requirements in 10 CFR 50.55a apply. The changes relate to inservice 
inspection (ISI) and inservice testing (IST) requirements which are 
specified in 10 CFR 50.55a, ``Codes and Standards.'' The ISI and IST 
requirements are given in 10 CFR 50.55a, which the licensee documents 
via its 10 year interval program requirements. Upon finding a Code 
requirement impractical because of limitations in the design (including 
prohibitive dose rates), construction, or system configurations, NNECO 
[Northeast Nuclear Energy Company] would be required to prepare the 
determination describing the impractical condition(s) and the 
applicable code requirements that cannot be met in accordance with 10 
CFR 50.55a, paragraphs (f)(5)(iii) and (iv), and (g)(5)(iii) and (iv) 
if within the first 12 months of a new interval. For example, 10 CFR 
50.55a(f)(5)(iv), and (g)(5)(iv) allow a licensee up to a full year 
after the beginning of an updated interval to inform the NRC of the new 
Code requirements which cannot be met and to request relief. If an 
impracticality is identified after the first 12 months, the guidance 
contained in NUREG-1482 will be followed. This will eliminate 
inconsistencies between the Technical Specifications and the 
regulations. There will be no adverse effect on plant operations. 
Therefore, the proposed changes will not result in a significant 
increase in the probability or consequences of an accident previously 
evaluated.
Technical Specification 3.2.3
    The proposed change will change the mode of applicability for 
Technical Specification 3.2.3 from Mode 1 to Mode 1 with thermal power 
>20%. Data from the incore detectors are used for determining the 
measured radial peaking factors to verify compliance with Technical 
Specification 3.2.3. However, the accuracy of the neutron flux 
information from the incore detectors is not reliable below 20% power. 
The proposed change acknowledges this limitation of the incore 
detectors by changing the applicability of this specification to power 
levels where the data from the incore detectors is reliable. This will 
have no adverse effect on plant operations since the current Technical 
Specification surveillance requirements do not require the verification 
of this limit until prior to operation above 70% following each fuel 
loading, prior to 31 days accumulated operation in Mode 1, or if the 
azimuthal power tilt limit is exceeded (Technical Specification 3.2.4 
which is applicable in Mode 1 above 50% power). Therefore, the proposed 
change has no impact on the initial conditions, with respect to power 
distribution, assumed in the accident analysis. Thus, the proposed 
change will not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
Technical Specification 3.3.2.1
    The proposed change will add an exception to Technical 
Specification 4.0.4 that will allow the channel functional test of the 
automatic actuation logic associated with ESF [engineered safety 
feature] actuations for safety injection, containment spray, 
containment isolation, main steam line isolation, enclosure building 
filtration, and containment sump recirculation to be delayed during 
plant startup until the actuation blocks are removed. This will allow 
entry into Mode 3 where plant conditions (sufficient pressurizer and 
steam generator pressure) can be established that will automatically 
remove the blocks of these ESF actuations. The channel functional test 
of the automatic actuation logic, using the ATI [Automatic Testing 
Insertor] circuit, will then be performed. In addition, the channel 
functional tests of the automatic actuation logic must be performed 
prior to entering Mode 2.
    The exception to Technical Specification 4.0.4 allows a mode change 
with equipment that is inoperable only because conditions [cannot] be 
established to perform the SR [surveillance requirement] until after 
the mode is entered. All other equipment operability requirements must 
be met. Even though operability of the automatic actuation logic for 
the affected ESF actuations cannot be verified prior to entering Mode 
3, this equipment is still expected to be operable. The ESFAS 
[engineered safety feature actuation system] will continue to function 
as before. Therefore, the proposed change will not result in a 
significant increase in the probability or consequences of an accident 
previously evaluated.
Technical Specification 3.4.1.1
    The Flow Dependent Setpoint Selector Switch was installed to allow 
power operation with less than four reactor coolant pumps (RCPs) in 
operation by changing the reactor trip setpoints for the variable high 
power, Reactor Coolant System (RCS) low flow,

[[Page 56253]]

and thermal margin low pressure (TM/LP) reactor trips. Millstone Unit 
No. 2 is not currently licensed to operate with less than four RCPs in 
operation. Therefore, this switch should be maintained in the four pump 
position.
    The use of the switch position to ensure compliance with Technical 
Specification 3.4.1.1 provides an indirect verification of LCO 
[limiting condition for operation] compliance since the loss of an RCP 
will result in a reactor trip when in the four pump position. The 
proposed change will replace the method used for LCO verification with 
one that is more consistent with the LCO. Verification of switch 
position is performed as a prerequisite prior to reactor startup 
(entering Mode 2). It is not necessary to verify the switch position 
every 12 hours as currently required. The position of this switch is 
important to the operability of the associated Reactor Protection 
System (RPS) trips variable high power, RCS low flow, and TM/LP). The 
operability of these RPS trips and associated setpoints is already 
covered by Technical Specifications 2.2.1, ``Reactor Trip Setpoints,'' 
and 3.3.1.1, ``Reactor Protective Instrumentation.''
    It is not necessary to verify the position of this switch fifteen 
minutes prior to reactor criticality since the switch position is 
verified prior to a reactor startup, and is not expected to be changed 
during power operation. If surveillance testing or maintenance 
activities are to be performed which may require the switch to be in 
other than the four pump position, the affected RPS channels will 
already have been removed from service (declared inoperable and placed 
in the tripped or bypassed condition) prior to commencing the 
activities. In addition, a light (``PUMP SETPOINT ERROR'') on each of 
the RPS Calibration and Indication Panels will illuminate if the switch 
is not in the four pump position.
    It is also not necessary to verify compliance with the requirements 
of Technical Specification 3.4.1.1 within fifteen minutes prior to 
reactor criticality since this condition is verified prior to a reactor 
startup, and the RPS will initiate a reactor trip if less than four 
RCPs are in operation.
    The proposed change will replace SR 4.4.1.1, verification of the 
Flow Dependent Setpoint Selector Switch position, with a verification 
check of the required RCS loops. This verification is more consistent 
with the Limiting Condition for Operation (LCO). This will not change 
the requirement that both RCS loops be operable and operating in Modes 
1 and 2. The Technical Specification will continue to assure that the 
initial condition, with respect to RCS loops in service, in the 
accident analysis is applicable. Therefore, the proposed change will 
not result in a significant increase in the probability or consequences 
of an accident previously evaluated.
Technical Specification 3.4.11
    The proposed change to modify the wording of SR 4.4.11.3 will not 
affect the operability requirements of the RCS Vent System. This change 
will provide operational flexibility to use a series of overlapping 
tests to verify flow through sections of the vent system, such that 
when completed, flow will be verified through all parts of the vent 
system. This will minimize potential contamination of the area 
surrounding the sparger and will eliminate the need to establish solid 
water conditions in the RCS.
    The proposed surveillance requirement will still verify the ability 
of the vent valves to operate. This will provide reasonable assurance 
of system operability and availability if needed to mitigate the 
consequences of design basis accidents. Therefore, the proposed change 
will not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes have no adverse effect on any of the design 
basis accidents previously evaluated or on any equipment important to 
safety. Therefore, the license amendment request does not impact the 
probability of an accident previously evaluate 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 reference, spelling, and 
terminology errors in various Technical Specification Definitions; add 
a new Technical Specification, 3.0.6; modify Technical Specification 
4.0.5 to remove an inconsistency between the Technical Specification 
and the regulations; change the applicability of Technical 
Specification 3.2.3; add an exception to Technical Specification 4.0.4 
to Technical Specification 3.3.2.1; modify the wording of a 
surveillance requirement associated with RCS Technical Specification 
3.4.1.1; and modify the wording of a surveillance requirement 
associated with the RCS Vent System, Technical Specification 3.4.11 to 
provide operational flexibility in the performance of the test. These 
changes will have no adverse effect on equipment important to safety. 
The equipment will continue to function as assumed in the design basis 
accident analysis. Therefore, there will be no significant reduction of 
the margin of safety as defined in the Bases for the Technical 
Specifications affected by these proposed changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Project Director: William M. Dean.

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

    Date of amendment request: September 14, 1998.
    Description of amendment request: The proposed amendment to the 
Limerick Generating Station (LGS), Unit 2, Technical Specifications 
(TS) would revise TS Table 4.4.6.1.3-1, ``Reactor Vessel Material 
Surveillance Program--Withdrawal Schedule.'' This table provides the 
schedule for withdrawing the reactor pressure vessel material 
surveillance program capsules. This proposed TS change involves 
revising the schedule for withdrawing the first surveillance capsule 
from 8 Effective Full Power years (EFPY) to 15 EFPY, and the second 
surveillance capsule from 20 EFPY to 30 EFPY.

[[Page 56254]]

    A revision to TS Surveillance Requirement (SR) 4.4.6.1.4 is also 
proposed. This revision will remove the reference to flux wire removal 
and analysis that was originally required following the first cycle of 
operation. TS SR 4.4.6.1.4 will be changed to refer to the flux wires 
that are located within the surveillance capsules, which will be 
removed and analyzed in accordance with the surveillance capsule 
removal schedule, located in Table 4.4.6.1.3-1.
    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 Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes do not increase the probability of occurrence 
of an accident previously evaluated in the safety analysis report and 
do not affect any accident initiators as described in the Safety 
Analysis Report (SAR). The change revises the withdrawal schedule for 
the reactor vessel material surveillance capsules. The capsules are not 
an initiator of any previously analyzed accident nor does the 
withdrawal schedule of the surveillance capsules affect the probability 
or consequences of any previously analyzed accident.
    The proposed changes will not affect the Pressure-Temperature (P-T) 
limits as specified in LGS TS Figure 3.4.6.1-1 and Updated Final Safety 
Analysis Report (UFSAR) Figure 5.3-4. P-T limits are imposed on the 
reactor coolant system to ensure that adequate safety margins exist 
during normal operation, anticipated operational occurrences, and 
system hydrostatic tests. The P-T limits are related to the 
RTNDT [reference temperatures], as described in ASME Section 
III, Appendix G. Changes in the fracture toughness properties of RPV 
[reactor pressure vessel] beltline materials, resulting from neutron 
irradiation and the thermal environment, are monitored by a 
surveillance program in compliance with the requirements of 10 CFR 50 
Appendix H. The effect of neutron fluence on the shift in the 
RTNDT is predicted by methods given in Regulatory Guide 
1.99, Rev.2.
    As detailed in Attachment 3 [of the September 14, 1998, submittal], 
for LGS, Unit 2, the combination of low expected RTNDT shift 
for the plate material due to low predicted fluence and excellent 
material chemistry; Supplemental Surveillance Program (SSP) data on 
similar material; and the inherent margin in the P-T curve 
calculations, with the withdrawal schedule of the first surveillance 
capsule modified from 8 EFPY to 15 EFPY and the second surveillance 
capsule modified from 20 EFPY to 30 EFPY, will result in more credible 
sets of surveillance data, while ensuring the continued safe operation 
of LGS, Unit 2.
    The current LGS P-T limits were established based on adjusted 
reference temperatures developed in accordance with the procedures 
prescribed in Regulatory Guide 1.99, Revision 2, Regulatory Position 1, 
``Surveillance Data Not Available.'' Calculation of adjusted reference 
temperature by these procedures includes a conservative base fluence 
estimate; power rerate adjustment of a 110% fluence multiplier from 
startup, instead of a 105% fluence multiplier since 2R03 [third 
refueling outage]; and a margin term to ensure conservative, upper-
bound values are used for the calculation of the P-T limits. Revision 
of the first capsule withdrawal schedule will not affect the P-T limits 
because they will continue to be established in accordance with 
Regulatory Position 1 guidance. Also, as indicated in Attachment 3, it 
is also appropriate to extend the withdrawal of the LGS, Unit 2, second 
capsule. The current schedule specifies withdrawal of the second 
capsule at 20 EFPY. Based upon the information provided in Attachment 3 
supporting withdrawal of the first capsule at 15 EFPY, there will be an 
insignificant shift in material properties at 20 EFPY, after only an 
additional exposure of 5 EFPY. It is appropriate to extend this 
schedule to 30 EFPY which meets the intent of ASTM E185-82, such that 
the withdrawal of the second capsule occurs before the accumulated 
neutron fluence of the capsule corresponds to the approximate EOL [end 
of life] fluence at the reactor pressure vessel inner wall location, 
and provides consistency with the LGS, Unit 1, withdrawal schedule.
    In accordance with the guidance stipulated in Regulatory Guide 
1.99, ``Radiation Embrittlement of Reactor Vessel Materials,'' Revision 
2, Regulatory Position 2, ``Surveillance Data Available,'' the 
collection of two (2) or more sets of credible surveillance data is 
necessary to empirically calculate the adjusted reference temperature 
(ART). Each surveillance capsule constitutes one set of credible 
surveillance data. This calculated ART can be used to revise the P-T 
curves (TS Figure 3.4.6.1-1). Without two (2) or more sets of credible 
data, the ART must be calculated and the P-T curves revised, based upon 
the calculational methodologies as provided in the Regulatory Guide 
1.99, Revision 2, Regulatory Position 1, ``Surveillance Data Not 
Available.'' These methodologies use plant specific chemistry and 
fluence values to determine a calculated shift in RTNDT. A 
``margin'' term is then added, to obtain conservative, upper-bound 
values of adjusted reference temperature.
    The existing LGS, Unit 2, P-T curves are based upon the Regulatory 
Position 1 methodology, and are currently valid up to 10 EFPY. With 
first capsule removal at either 8 or 15 EFPY, the existing P-T curves 
will require a revision, prior to reaching 10 EFPY, based upon the 
calculational methodologies as contained in the Regulatory Guide 1.99, 
Revision 2, Regulatory Position 1, ``Surveillance Data Not Available.'' 
Therefore, the Technical Specification revision to the first capsule 
withdrawal schedule, as supported by this Safety Evaluation [supporting 
information described in attachments 1 and 3 of the September 14, 1998, 
submittal], results in no impact to the calculational methodologies 
that will be used for the P-T curve revision that will be necessary to 
extend the curves beyond 10 EFPY.
    The fluence data as determined from the surveillance capsule flux 
wires at 15 EFPY will provide an accurate indication of neutron 
fluence. In accordance with Regulatory Guide 1.99, Revision 2, 
Regulatory Position 1 methodology, data from these flux wires will 
permit an adjustment of TS Figure 3.4.6.1-1 in accordance with TS SR 
4.4.6.1.3, if required, and will meet the requirements of 10 CFR 50, 
Appendix H, and ASTM E-185.
    The proposed changes will not affect any plant safety limits or 
limiting conditions of operation. The proposed changes will not affect 
reactor pressure vessel performance as it involves no physical changes 
and LGS P-T limits will remain conservative in accordance with 
Regulatory Guide 1.99, Revision 2, guidance. The proposed changes will 
not cause the reactor pressure vessel or interfacing systems to be 
operated outside of their design or testing limits.
    The proposed changes do not increase the probability of the 
occurrence of a malfunction, or consequences of a malfunction, of 
equipment important to safety previously evaluated in the SAR. The 
proposed changes do not involve any physical changes to equipment 
important to safety. The potential for reactor vessel failure will be 
adequately assessed by the proposed withdrawal schedule. In addition, 
the results from

[[Page 56255]]

the Supplemental Surveillance Program (SSP) will provide industry data 
that bounds the materials used in the LGS vessel until the data from 
the first LGS capsule is available. The proposed change provides the 
same level of confidence in the integrity of the vessel. The P-T curves 
are currently controlled by the TS and are determined using the 
conservative methodology delineated in Regulatory Guide 1.99. 
Therefore, the possibility of failure of the reactor vessel is not 
increased. The current P-T limit curves are inherently conservative and 
will continue to be adhered to.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed changes do not create the possibility of a different 
type of accident than any previously evaluated in the SAR. The proposed 
changes are a revision of the withdrawal schedule for the first reactor 
pressure vessel material surveillance capsule from 8 EFPY to 15 EFPY, 
and for the second capsule from 20 EFPY to 30 EFPY. The proposed 
changes do not involve a physical modification of the design of plant 
structures, systems, or components. The proposed changes will not 
impact the manner in which the plant is operated as plant operating and 
testing procedures will not be affected by the change. No new accident 
types or failure modes will be introduced as a result of the proposed 
change.
    LGS's current P-T limits were established based on adjusted 
reference temperatures developed in accordance with the procedures 
prescribed in Regulatory Guide 1.99, Revision 2, Regulatory Position 1, 
``Surveillance Data Not Available.'' Calculation of adjusted reference 
temperature by these procedures includes a conservative base fluence 
estimate; power rerate adjustment of a 110% fluence multiplier from 
startup, instead of a 105% fluence multiplier since 2R03; and a margin 
term to ensure conservative, upper-bound values are used for the 
calculation of the P-T limits. Revision of the first capsule withdrawal 
schedule will not affect the P-T limits because they will continue to 
be established in accordance with the guidance of Regulatory Position 1 
of Regulatory Guide 1.99. Also, as specified in Attachment 3, it is 
appropriate to extend the withdrawal of the LGS, Unit 2, second 
capsule. The current schedule specifies withdrawal of the second 
capsule at 20 EFPY. Based upon the information provided in Attachment 3 
supporting withdrawal of the first capsule at 15 EFPY, there will be an 
insignificant shift in material properties at 20 EFPY, after only an 
additional exposure of 5 EFPY. It is appropriate to extend this 
schedule to 30 EFPY which meets the intent of ASTM E185-82, such that 
the withdrawal of the second capsule occurs before the accumulated 
neutron fluence of the capsule corresponds to the approximate EOL 
fluence at the reactor inner wall location, and provides consistency 
with the LGS, Unit 1, withdrawal schedule.
    The existing LGS, Unit 2, P-T curves are based upon the Regulatory 
Position 1 methodology, and are currently valid up to 10 EFPY. With 
first capsule removal at either 8 or 15 EFPY, the existing P-T curves 
will require a revision, prior to reaching 10 EFPY, based upon the 
calculational methodologies as contained in the Regulatory Guide 1.99, 
Revision 2, Regulatory Position 1, ``Surveillance Data Not Available.'' 
Therefore, the proposed TS revision to the first capsule withdrawal 
schedule results in no impact to the calculational methodologies that 
will be used for the P-T curve revision that will be necessary to 
extend the curves beyond 10 EFPY.
    The fluence data as determined from the surveillance capsule flux 
wires at 15 EFPY will provide an accurate indication of neutron 
fluence. In accordance with Regulatory Guide 1.99, Revision 2, 
Regulatory Position 1 methodology, data from these flux wires will 
permit an adjustment of TS Figure 3.4.6.1-1 in accordance with TS SR 
4.4.6.1.3, if required, and will meet the requirements of 10 CFR 50, 
Appendix H, and ASTM E-185.
    The potential for reactor vessel failure will be adequately 
assessed by the proposed withdrawal schedule. In addition, the results 
from the SSP will provide industry data that bounds the materials used 
in the LGS vessel, until the data from the first LGS capsule is 
available. The proposed changes provide the same level of confidence in 
the integrity of the vessel . The P-T curves are currently controlled 
by the TS and are determined using the conservative methodology in 
Regulatory Guide 1.99. Therefore, the possibility of failure of the 
reactor vessel is not increased. The current P-T limit curves are 
inherently conservative and will continue to be adhered to.
    Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes to the TS do not reduce the margin of safety 
as defined in the Bases for any TS. The proposed changes will not 
affect any safety limits, limiting safety system settings, or limiting 
conditions of operation. The proposed changes do not represent a change 
in initial conditions, system response time, or in any other parameter 
affecting the course of an accident analysis supporting the Bases of 
any TS. The proposed changes do not involve revision of the P-T limits, 
but rather a revision of the withdrawal schedule for the surveillance 
capsules. The current P-T limits were established based on the adjusted 
reference temperatures for reactor pressure vessel beltline materials 
calculated in accordance with the guidance stipulated in Regulatory 
Position 1 of Regulatory Guide 1.99, Revision 2. P-T limits will 
continue to be revised as necessary for changes in adjusted reference 
temperature due to changes in fluence according to Regulatory Position 
1 until two (2) or more credible surveillance data sets becomes 
available. When two (2) or more credible surveillance data sets become 
available, P-T limits will be revised as prescribed by Regulatory 
Position 2 of Regulatory Guide 1.99, Revision 2, or other NRC approved 
guidance.
    The current P-T limit curves are inherently conservative and 
provide sufficient margin to ensure the integrity of the reactor 
vessel. The changes do not adversely affect these curves. The fluence 
data as determined from the surveillance capsule flux wires at 15 EFPY 
will provide an accurate indication of neutron fluence. In accordance 
with Regulatory Guide 1.99, Revision 2, Regulatory Position 1 
methodology, data from these flux wires will permit an adjustment of TS 
Figure 3.4.6.1-1 in accordance with TS SR 4.4.6.1.3, if required, and 
will meet the requirements of 10 CFR 50, Appendix H, and ASTM E-185.
    Therefore, the proposed TS changes do not involve a reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

[[Page 56256]]

    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: April 16, 1998.
    Description of amendment request: This application for amendment to 
the Indian Point 3 Technical Specifications (TSs) proposes to modify a 
testing requirement for the emergency diesel generators (EDGs).
    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 proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
analyzed?
    Response:
    No. The three Emergency Diesel Generators (EDG) at Indian Point 3 
are designed to provide a source of power to support a safe and orderly 
plant shutdown in the event that all other normal and standby sources 
of power are not available, such as during a postulated Loss of Offsite 
Power (LOOP). The probability of such events occurring is not affected 
by the proposed amendment. Any two of the three EDGs are capable of 
supplying the minimum power requirements for emergency safeguards 
equipment that mitigate the consequences of postulated design basis 
accident conditions. Periodic preventive maintenance and surveillance 
testing are performed to provide assurance that the operability of all 
three EDGs is maintained. In the event that an inoperable EDG is 
identified, both the existing specification and the proposed change 
provide for actions that verify the operability of the remaining 2 
EDGs. Operability of 2 EDGs ensures that sufficient emergency power is 
available, if needed, to mitigate the consequences of postulated 
accidents. Therefore, the proposed license amendment does not involve a 
significant increase in the probability or consequences of an accident 
previously analyzed.
    (2) Does the proposed license amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    Response:
    No. The proposed license amendment does not involve any physical 
changes to plant systems or component setpoints. Also, there are no 
changes to the way in which systems or equipment are operated. The 
proposed change will continue to require that the operability of the 
remaining two EDGs be verified if one of the three EDGs is found to be 
inoperable. The proposed change to allow the use of a common cause 
failure evaluation, as an alternative to testing, to accomplish the 
operability verification can benefit overall EDG reliability by 
eliminating unnecessary EDG starts. Therefore, the proposed license 
amendment does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction in 
a margin of safety?
    Response:
    No. Important performance requirements for the EDGs include 
electrical output capacity, elapsed time to start and reach rated 
output, and fuel storage supply to support a minimum period of 
operation. The proposed amendment does not change EDG performance 
requirements. The existing specification allows a period of 24 hours in 
which to verify the operability of the remaining 2 EDGs if one of the 
three EDGs is found inoperable. The proposed amendment does not change 
the 24-hour time limit. Operability verification, either by testing or 
evaluation, within 24 hours provides assurance that this source of 
emergency power is available if needed. Therefore, the proposed 
amendment does not involve a significant reduction in a margin of 
safety. Also, this verification method has been approved for use with 
the current Standard Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director

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

    Date of amendment request: April 16, 1998, as supplemented August 
20, 1998.
    Description of amendment request: This application for amendment to 
Table 4.1-1 of the Indian Point 3 Technical Specifications (TSs) 
proposes to change surveillance frequency requirements for the various 
instrument channels to accommodate a 24-month operating cycle. The 
proposed amendment also revises Section 6 of the TSs to reflect updated 
analyses.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Does the proposed license amendment involve a significant increase 
in the probability or consequences of an accident previously analyzed?
    Response:
    No. The proposed license amendment to extend the calibration 
surveillance frequency of the following instrument channels is being 
made to support plant operation with a 24-month fuel cycle:
    (a) Pressurizer Water Level
    (b) Accumulator Level and Pressure
    (c) Reactor Coolant System Subcooling Margin Monitor
    (d) Core Exit Thermocouples
    (e) Reactor Vessel Level Indication System
    Changing the calibration intervals for these instrument channels 
neither directly nor indirectly affects the initiation or probability 
of any previously analyzed accident. The changes do not affect the 
integrity of any of the principal barriers against radiation release 
(fuel cladding, reactor vessel, and containment building). The ability 
of the plant to mitigate the consequences of any previously analyzed 
accidents is not adversely affected. Evaluation of the proposed change 
to the surveillance interval demonstrates that licensing basis safety 
analyses acceptance criteria and Indian Point 3 Emergency Operating 
Procedure (EOP) criteria continue to be met.
    Item (a) provides an input to the Reactor Protection System (RPS) 
to initiate a reactor trip if the measured parameters exceed specified 
values. Item (b) is used by control room operators to ensure that the 
accident mitigation capability of the accumulators is maintained within 
specified limits. Items (c), (d), and (e)

[[Page 56257]]

provide post-accident information to control room operators to support 
recovery efforts. Item (d) is also used to monitor core performance for 
fuel management activities.
    The proposed new surveillance frequency for these instrument 
channels was evaluated using the guidance of Generic Letter 91-04. The 
basis for the changes includes a quantitative evaluation of instrument 
drift. Also, loop accuracy/setpoint calculations were updated to 
accommodate the extended surveillance period. Analyses and evaluations 
completed to assess the proposed increase in the surveillance interval 
demonstrate that the effectiveness of these instruments in fulfilling 
their respective functions is maintained. Channel checks required to be 
performed each shift or each day, according to Technical Specifications 
for the subject channels, will continue to be performed to provide 
assurance of instrument channel operability. Therefore, the proposed 
amendment does not involve a significant increase in the probability or 
consequences of any previously analyzed accident.
    Does the proposed license amendment create the possibility of a new 
or different kind of accident from any accident previously evaluated?
    Response:
    No. The increased calibration surveillance intervals for the above 
listed instrument channels were justified based on evaluation of past 
equipment performance and do not require any plant hardware changes or 
changes in normal system operation. Changing the calibration intervals 
for these channels neither directly nor indirectly has any means of 
creating the possibility of a new or different kind of accident. 
Certain alarm and EOP setpoint changes will be made consistent with the 
revised uncertainty calculations for the subject channels. These new 
setpoints and related operator responses support existing accident 
mitigation strategies and do not create the possibility of a new or 
different kind of accident from any previously analyzed. Therefore, 
there are no new failure modes introduced as a result of extending 
these surveillance intervals, and the proposed amendment does not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Response:
    No. Pressurizer water level instrumentation provides input to the 
reactor protection system and to the pressurizer water level control 
system. Pressurizer water level, as indicated by the selected control 
channel, is used to establish the initial condition pressurizer water 
level assumption for certain UFSAR [Updated Final Safety Analysis 
Report] Chapter 14 safety analyses. The proposed change to the 
calibration surveillance interval was evaluated using the criteria of 
95% probability/95% confidence level for process sensor drift. The loop 
accuracy/setpoint calculations were updated for the level channels to 
demonstrate the acceptability of the proposed increase in the 
surveillance interval. There are no changes required to the limiting 
safety system setting (LSSS) stated in the Technical Specifications for 
these channels. The LSSS for high pressurizer water level will remain 
at [less than or equal to] 92% of span. The margin of safety between 
the specified LSSS value required by Technical Specifications and the 
safety limit used in the UFSAR Chapter 14 safety analyses is unchanged.
    The instrument channels for accumulator pressure and level do not 
provide input to the reactor protection system or the engineered safety 
features system. These instruments provide alarms and indication to 
control room operators to maintain accumulator cover gas pressure and 
water volume within specified limits. They are also used for 
establishing initial condition accumulator pressure and level 
assumptions for certain UFSAR Chapter 14 safety analyses. Accordingly, 
the process sensor drift analysis was performed using the criteria of 
95% probability/75% confidence level.
    The remaining three instrument channels addressed by this proposed 
license change are used to provide indication of adequate core cooling 
following certain hypothetical accident conditions. These instrument 
channels are not associated with any margin of safety specified by the 
Technical Specifications, and they are not factors in any UFSAR Chapter 
14 safety analyses. However, they are factored into the calculations of 
pertinent setpoints used in alarm response procedures and EOPs. The 
updated drift and uncertainty calculations and evaluations for these 
instrument channels demonstrate that applicable accuracy requirements 
for Indian Point 3 are satisfied with the proposed new surveillance 
intervals. The instrument channels will remain effective to support 
plant operator implementation of the Emergency Operating Procedures, 
which are consistent with the Westinghouse Owners' Group Emergency 
Response Guidelines.
    Changing the calibration interval for these channels does not 
affect margin of safety for previously analyzed accidents. Also, the 
evaluation of related changes to UFSAR Chapter 14 safety analyses input 
assumptions has demonstrated that licensing basis safety analysis 
acceptance criteria and EOP criteria continue to be met, and previously 
existing margins based on these pertinent acceptance criteria continue 
to be maintained.
    Therefore, 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 50.92(c) are 
satisfied. The staff has also reviewed the licensee's proposed change 
to reflect updated safety analyses in Section 6 of the TSs and it 
appears that the three standards of 50.92(c) are satisfied for these 
changes as well. Therefore, the NRC staff proposes to determine that 
the amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of amendment request: September 17, 1998.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3/4.8.2, ``Electrical Power Sources--
Shutdown,'' for the AC distribution system and the 125-volt and 28-volt 
DC distribution systems. Specifically, the amendments would change the 
Applicability and Action Statements, if less than the complement of 
equipment and busses are operable, to eliminate the need to establish 
containment integrity and to add the action to suspend core 
alterations, positive reactivity additions, and movement of irradiated 
fuel assemblies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will not involve a significant increase in the probability or

[[Page 56258]]

consequences of an accident previously evaluated.
    In Modes 1 through 4 [power operation through hot shutdown], a 
Design Basis Accident would cause the release of radioactive material 
into the containment. Release of that radioactive material to the 
environment is prevented during operation in Modes 1 through 4 by 
maintaining containment integrity. In Modes 5 and 6 [cold shutdown and 
refueling] the probability and consequences of this event are lower 
because of the reduced reactor coolant pressure and temperature 
limitations of these modes.
    A minimum complement of electrical power sources and distribution 
systems is established in Modes 5 and 6 to assure that adequate 
electrical power is available to mitigate the consequences of a fuel 
handling accident. Because of the lack of containment pressurization 
potential during a fuel handling accident, less stringent requirements 
are needed to isolate containment from the outside atmosphere. These 
requirements are applied during refueling operations by Technical 
Specification 3.9.4, Refueling Operations, Containment Building 
Penetrations. Technical Specification 3.9.4 is applicable in Mode 6 and 
establishes containment closure vice containment integrity during 
refueling operation (core alterations and movement of irradiated fuel 
within containment).
    In Mode 5, fuel handling is generally limited to placement of new 
fuel prior to core off load or movement of irradiated fuel within the 
spent fuel pool. Because the Spent Fuel Pool is not located within 
containment, establishment of either containment integrity or 
containment closure would not help to mitigate the consequences of a 
fuel handling accident in that area. Mitigation of a fuel handling 
accident is accomplished through Technical Specification 3.9.12, 
Refueling Operations, Fuel Handling Area Ventilation System, which 
requires that the Fuel Handling Area Ventilation system be operable 
whenever irradiated fuel is present in the storage pool. This insures 
that all radioactive material released from the rupture of an 
irradiated fuel assembly would be filtered through filtration equipment 
prior to discharge to the atmosphere.
    With the number of energized A.C. or D.C. power distribution 
systems less than the required, sufficient power may not be available 
to recover from a fuel handling accident. Consequently, the Action 
statements require immediate suspension of all operations involving 
core alterations, positive reactivity changes, and movement of 
irradiated fuel assemblies. This precludes the possibility of a fuel 
handling accident and the need for containment integrity.
    Based upon the above, the proposed change will not increase the 
probability or consequences of an accident previously analyzed.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes do not require any change in the configuration 
or operation of the plant. Specifically, no new hardware is being added 
to the plant as part of the proposed change, no existing equipment is 
being modified, and no significant changes in operations are being 
introduced. Therefore, these changes will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Will not involve a significant reduction in a margin of safety.
    The proposed change will not alter any assumptions, initial 
conditions, or results of any accident analyses. The proposed 
additional Applicability will ensure proper operation of the Fuel 
Handling Area Ventilation system during movement of irradiated fuel in 
the spent fuel pool. The proposed ACTIONS, to be taken in the event 
that the LCO [limiting condition for operation] is not met, will 
preclude the conditions that would lead to the need for establishing 
containment integrity. The change will, therefore, not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: Robert A. Capra.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of amendment request: September 29, 1998.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3/4.9.4, ``Refueling Operations, 
Containment Building Penetrations,'' to permit the use of equivalent 
methods to obtain containment closure during refueling operations. 
Specifically, the proposed changes would allow the installation of an 
outage equipment door or other closure devices that are capable of 
providing access for temporary services needed to support maintenance 
activities within containment.
    In addition to the above changes, the terminology for the 
Containment Equipment Hatch inside door used in LCO 3.9.4.a is being 
changed. The term ``Containment Equipment Door'' is being changed to 
``Containment Equipment Hatch Inside Door'' to bring it into agreement 
with the terminology used in Salem design documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    In Modes 1 through 4 [power operations through hot shutdown], a 
Design Basis Accident would cause the release of radioactive material 
into the containment. The release of radioactive material from the 
containment to the environment is prevented during operation in Modes 1 
through 4 by maintaining CONTAINMENT INTEGRITY. In Mode 5 and 6 [cold 
shutdown and refueling] the requirements to prevent releases from the 
containment to the environment from postulated accidents are less 
stringent because of the reduced reactor coolant pressure and 
temperature limitations of these modes. In all cases, the containment 
serves as a passive barrier to mitigate the consequences of accidents 
analyzed. The containment is not considered to be a contributor to the 
probability of those accidents. Therefore, this change, which will 
permit the use of equivalent methods for establishing containment 
closure during refueling operations, will not increase the probability 
of an accident previously analyzed.
    During refueling operations, a release of radioactive material to 
the containment could occur as the result of a fuel handling accident. 
Actions are taken to mitigate the consequences of a fuel handling 
accident inside containment during refueling operations through 
application of technical specification requirements for Refueling 
Cavity water level, minimum decay time prior to CORE ALTERATIONS, and 
Containment Building Penetrations.
    Because of the lack of containment pressurization potential and the 
reduced

[[Page 56259]]

source term during a fuel handling accident, less stringent 
requirements are needed to isolate containment from the outside 
atmosphere. These requirements are applied during refueling operations 
by Technical Specification 3.9.4, Refueling Operations, Containment 
Building Penetrations. Technical Specification 3.9.4 is applicable in 
Mode 6 and establishes containment closure vice CONTAINMENT INTEGRITY 
during CORE ALTERATIONS and movement of irradiated fuel within 
containment. Containment closure means that all potential release paths 
are closed or capable of being closed to provide an atmospheric 
pressure, ventilation barrier. Since there is no potential for 
containment pressurization, establishment of a pressure tight boundary 
is not required.
    As a part of the containment closure requirements of Technical 
Specification 3.9.4, the Containment Equipment Hatch inside door must 
be installed with a minimum of four bolts. In addition, each 
penetration providing direct access from the containment atmosphere to 
the outside atmosphere must be closed by either an isolation valve, a 
blind flange, or a manual valve, or must be capable of being closed by 
an OPERABLE automatic containment isolation valve.
    The proposed changes will modify Technical Specification 3/4.9.4 to 
permit the use of an equivalent closure device as an alternative to 
installation of the inner door with a minimum of four bolts to provide 
containment closure for the Containment Equipment Hatch. The proposed 
change will also modify Technical Specification 3.9.4 to permit the use 
of an equivalent method for containment closure for containment 
penetrations providing direct access from the containment to the 
outside atmosphere as an alternate method to closure by an isolation 
valve, blind flange, or manual valve. Any alternate method used will be 
designed, fabricated, installed, tested, and utilized in accordance 
with established procedures to ensure that it is capable of providing 
containment closure during a fuel handling accident to prevent the 
release of fission product radioactivity to the environment. Because 
the proposed technical specifications must provide equivalent 
containment closure, these changes will not increase the consequences 
of an accident previously evaluated.
    Based upon the above, the proposed changes do not increase the 
probability or the consequences of an accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes do not require any change in the operation of 
the plant. The proposed changes will permit the use of an equivalent 
method to achieve containment closure for the Containment Equipment 
Hatch or for individual containment penetrations that provide direct 
access to the outside atmosphere. However, any equivalent method used 
will be designed, fabricated, installed, tested, and utilized in 
accordance with established procedures to ensure that the closure 
method meets design requirements.
    Based upon the above, these changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will not involve a significant reduction in a margin of safety.
    The proposed change will not affect the existing analysis that 
forms the basis for the Technical Specifications, and does not violate 
Technical Specification and Updated Final Safety Analysis Report 
(UFSAR) requirements. The proposed change will not affect any design or 
functional requirements of the containment, the Containment Equipment 
Hatch, or containment penetrations or any conditions or assumptions of 
the applicable safety analyses.
    Based upon the above, the proposed changes will not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: Robert A. Capra.

Southern California Edison Company, et al., Docket No. 50-362, San 
Onofre Nuclear Generating Station, Unit No. 3, San Diego County, 
California

    Date of amendment request: September 22, 1998.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) to change the parameter used 
to establish and remove the bypasses for high reactor power trips. The 
parameter would be changed from the current ``THERMAL POWER'' to 
logarithmic power. This amendment was processed on San Onofre Nuclear 
Generating Station (SONGS) Unit 2 under emergency circumstances to 
allow resumption of power operations, and is being processed under 
normal notice circumstances on SONGS Unit 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change to Technical Specification (TS) 3.3.1 does not 
adversely impact structure, system, or component design or operation in 
a manner which would result in a change in the frequency of occurrence 
of accident initiation. The reactor trip bypass and automatic enable 
functions are not accident initiators. Consequently, the proposed TS 
change will not significantly increase the probability of accidents 
previously evaluated. Clarifying the input process variable of the 
operating bypasses and automatic bypass removals of the affected 
reactor trips does not alter the setpoint nor the manner of operation 
of the operating bypasses and automatic bypass removals. Therefore, the 
consequences of previously evaluated accidents remain unchanged.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    No new or different accidents result from clarifying the input 
process variable of the operating bypasses and automatic bypass 
removals of the affected reactor trips. The results of previously 
performed accident analyses remain valid. Therefore, this amendment 
request does not create the possibility of a new or different kind of 
accident.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change does not alter the setpoint nor the manner of 
operation of the operating bypasses and automatic bypass removals of 
the affected reactor trips. The change merely replaces the 
identification of the input process variable with the appropriate 
identification of power. Therefore, this amendment request does not 
involve a significant reduction in any margin of safety.

[[Page 56260]]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: August 31, 1998.
    Description of amendment request: The proposed amendment would 
revise the cold overpressure mitigation curves in Technical 
Specification (TS) Figure 3.4-4. This change would account for the TS 
maximum allowable power-operated relief valve setpoint changes 
associated with the new Model Delta 94 steam generator operating 
parameters.
    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 current pressurizer maximum allowable Power Operated Relief 
Valve (PORV) setpoints, provided by the Cold Overpressure Mitigation 
System (COMS) curves (Figure 3.4-4) of Technical Specification 3.4.9.3, 
are nonconservative for application with the new Delta 94 Replacement 
Steam Generators. The South Texas Project Cold Overpressure Event has 
been re-analyzed as a result of changed operating parameters due to 
installation of new Delta 94 Steam Generators. The re-analysis 
determined that maximum allowable PORV setpoint required decreases to 
ensure that the Cold Overpressure Mitigation System (COMS) continued to 
provide design basis low temperature overpressure protection with Delta 
94 Steam Generators. New COMS curves have been developed and are to be 
incorporated into Technical Specification 3.4.9.3 by this change 
request. Since the proposed COMS curves result in maximum allowable 
PORV setpoint decreases to account for the changed Delta 94 Steam 
Generator operating parameters, these curves are more conservative than 
the existing COMS curves utilized for Model E Steam Generators. 
Therefore, application of these proposed COMS curves for a unit with 
Model E or Delta 94 Steam Generators ensures compliance with the 
original design basis of the Cold Overpressure Mitigation System for 
the South Texas Project.
    This proposed change is based on a re-analysis which accounts for 
changed operating parameters associated with the Delta 94 Replacement 
Steam Generators. Reflecting actual operating parameters and adjusting 
the maximum allowable PORV setpoints, as necessary, in the conservative 
direction has no adverse effect on the probability or consequences of 
an accident previously evaluated. 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 accident previously evaluated.
    The proposed PORV maximum allowable setpoint changes do no create 
any new operating conditions or modes. The proposed change only revises 
the maximum allowable PORV setpoint curves for the Cold Overpressure 
Mitigation System to account for the revised operating parameters 
associated with Delta 94 Steam Generators. The actions of this system 
continue to be performed in accordance with existing requirements, 
which are sufficient to ensure plant safety is maintained.
    The proposed change is the result of a re-analysis of a previously 
evaluated accident. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change reflects the revised operating parameters 
associated with the new Delta 94 Steam Generators. The revised COMS 
curves are the result of a re-analysis of the COMS analysis performed 
to ensure the margin of safety is not reduced with Delta 94 Steam 
Generators. 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: August 31, 1998.
    Description of amendment request: The proposed amendment would 
modify Technical Specification Surveillance Requirement 3.6.1.3.4 to 
permit removal of the inclined fuel transfer system primary containment 
blind flange while primary containment integrity is required.
    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 permits removal of the blind flange on the 
Inclined Fuel Transfer System (IFTS) when primary containment 
operability is required in Modes 1, 2 and 3. This will permit operation 
of IFTS when the plant is operating. This aspect of the containment 
structure does not directly interface with the reactor coolant pressure 
boundary. The removal of this blind flange does not involve 
modifications to plant systems or design parameters that could 
contribute to the initiation of any accidents previously evaluated. 
Operation of IFTS is unrelated to the operation of the reactor, and 
there is no aspect of IFTS operation that could lead to or contribute 
to the probability of occurrence of an accident previously evaluated. 
Removal of the blind flange and operation of IFTS does not result in 
changes to procedures that could impact the probability of occurrence 
of an accident.
    With respect to consequences, the function of the containment is to

[[Page 56261]]

mitigate the radiological consequences of a loss of coolant accident 
(LOCA) or other postulated events that could result in radiation 
release from the fuel inside containment. The pressure and temperature 
transient resulting from a design basis loss of coolant accident (LOCA) 
is considered the primary challenge to the integrity of the 
containment. While the proposed change does not change the plant 
design, it does permit alteration of the containment boundary for the 
IFTS penetration. Altering the containment boundary in this case 
(removing the blind flange) results in some IFTS components possibly 
seeing a containment pressure rise should a LOCA occur. The thermal and 
mechanical load requirements do not appreciably change as a result of 
such a small pressure increase (peak post-accident pressure 
(Pa) of 7.8 psig). The IFTS components will be more than 
adequate and capable of withstanding the Design Basis LOCA and 
associated loads prior to implementation of this amendment. Therefore, 
they are considered an acceptable barrier to prevent uncontrolled 
release of post-accident fission products for this proposed change.
    The proposed change required examination of two potential leakage 
pathways. The larger is the transfer tube itself, the other, much 
smaller one, is the drain piping. It is clear that the gate valve at 
the bottom of the transfer tube is always water sealed and maintained 
so by the submergence of the water in the transfer tube and in the Fuel 
Handling Building Fuel Transfer Pool. The height of this water seal is 
greater than that necessary to prevent leakage from the bottom of the 
transfer tube during accidents that result in the calculated peak post-
accident pressure (Pa). The potential leakage pathway from 
the drain piping which attaches to the transfer tube will be isolated 
if required, via administrative controls on the drain piping isolation 
valve. Additionally, the drain piping isolation valve will be added to 
the Primary Containment Leakage Rate Testing Program (Specification 
5.5.12) to ensure that leakage past this valve will be maintained 
consistent with the leakage rate assumptions of the accident analysis. 
Due to the test methodology, the portion of the large transfer tube 
piping outboard of the blind flange (the portion of the tube which 
becomes exposed to containment air during the draining portion of the 
IFTS operation) will also be part of the leakage rate test boundary and 
will therefore also be tested with air. Therefore, no unidentified 
leakage paths will exist from the piping and components that are 
outboard of the blind flange, and the leakage rate assumptions of the 
accident analysis will be maintained.
    Therefore, the proposed change does not result in a significant 
increase in the probability or the consequences of previously evaluated 
accidents.
    (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 consists of the removal of a passive component 
which is not part of the primary reactor coolant pressure boundary nor 
involved in the operation or shutdown of the reactor. Being passive, 
its presence or absence does not affect any of the parameters or 
conditions that could contribute to the initiation of any incidents or 
accidents that are created from loss of coolant or positive reactivity. 
Re-aligning the boundary of the primary containment to include portions 
of the IFTS is also passive in nature and therefore has no influence 
on, nor does it contribute to the possibility of a new or different 
kind of incident, accident or malfunction from those previously 
analyzed. Furthermore, operation of IFTS is unrelated to the operation 
of the reactor and there is no mishap in the process that can lead or 
contribute to the possibility of losing any coolant in the reactor or 
introducing the chance for positive or negative reactivity or other 
accidents different from and not bounded by those previously evaluated.
    Therefore, the proposed change does not result in creating the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) The proposed change will not involve a significant reduction in 
the margin of safety.
    The proposed change involves the re-alignment of the primary 
containment boundary by removing the blind flange which is a passive 
component. The margin of safety that has the potential of being 
impacted by the proposed change involves the dose consequences of 
postulated accidents which are directly related to potential leakage 
through the primary containment boundary. The potential leakage 
pathways due to the proposed change have been reviewed, and leakage can 
only occur from the administratively controlled IFTS transfer tube 
drain piping. An individual will be designated to provide timely 
isolation of this drain piping during the durations of time when this 
proposed change is in effect. The conservatively calculated dose which 
might be received by the designated individual while isolating the 
drain piping is less than or equal to 1.9 rem, well within the 
guidelines of General Design Criterion 19. Furthermore, the drain 
piping isolation valve will be added into the Primary Containment 
Leakage Rate Testing Program (Specification 5.5.12) to ensure that 
leakage from the piping and components located outboard of the blind 
flange will be maintained consistent with the leakage rate assumptions 
of the accident analysis. Therefore, the dose consequences of an event 
would be unchanged, and the associated margin of safety would also be 
unchanged.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Stuart A. Richards.

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

    Date of amendment request: September 3, 1998.
    Description of amendment request: The proposed amendment would 
permit an Emergency Diesel Generator (EDG) Technical Specification (TS) 
Action Completion Time of up to 14 days for a Division 1 or 2 EDG and 
allow performance of the EDG 24-hour TS surveillance requirement test 
in modes 1 and 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 Technical Specification changes do not significantly 
increase the probability of occurrence of a previously evaluated 
accident because the standby

[[Page 56262]]

Emergency Diesel Generators (EDGs), including the High Pressure Core 
Spray diesel generator, are not initiators of previously evaluated 
accidents. The EDGs mitigate the consequences of previously evaluated 
accidents involving a loss of offsite power. The proposed changes to 
the Technical Specification Action Completion Times do not affect any 
of the assumptions used in the deterministic or Probabilistic Safety 
Analysis (PSA).
    The proposed Technical Specification changes will continue to 
ensure the EDGs perform their function when called upon. Extending the 
Technical Specification Completion Times to 14 days and allowing the 
performance of the EDG 24-hour run test in either Modes 1 or 2 does not 
affect the design of the EDGs, the operational characteristics of the 
EDGs, the interfaces between the EDGs and other plant systems, the 
function, or the reliability of the EDGs. Thus, the EDGs will be 
capable of performing their accident mitigation function and there is 
no impact to the radiological consequences of any accident analysis.
    To fully evaluate the effect of the EDG Completion Time extension, 
PSA methods and deterministic analysis were utilized. The results of 
this analysis show no significant increase in the Core Damage 
Frequency. The proposed changes remain bounded by the Core Damage 
Frequency identified in the Individual Plant Examination.
    The Configuration Risk Management Program (CRMP) is an 
administrative program that assesses risk based on plant status. Adding 
the requirement to implement the CRMP for Technical Specification 3.8.1 
requires the consideration of other measures to mitigate consequences 
of an accident occurring while an EDG is inoperable.
    The proposed change will not alter the operation of any plant 
equipment assumed to function in response to an analyzed event or 
otherwise increase its failure probability. Therefore, this change does 
not involve a significant increase in the probability or consequences 
of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change does not change the design, configuration, or 
method of operation of the plant. The proposed activity involves a 
change to the allowed plant mode for the performance of specific 
Technical Specification surveillance requirements. No physical or 
operational changes to the EDGs or supporting systems are made by this 
activity. Since the proposed changes do not involve a change to the 
plant design or operation, no new system interactions are created by 
this change. The proposed Technical Specification changes do not 
produce any parameters or conditions that could contribute to the 
initiation of accidents different from those already evaluated in the 
Updated Safety Analysis Report.
    The proposed changes only address the methods used to ensure EDG 
reliability. Thus, the proposed Technical Specification change does not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes do not affect the Limiting Conditions for 
Operation or their Bases that are used in the deterministic analysis to 
establish any margin of safety. PSA evaluations were used to evaluate 
these changes, and these evaluations determined that the changes are 
either risk neutral or risk beneficial. The proposed activity involves 
changes to certain Completion Times and to the allowed plant mode for 
the performance of specific Technical Specification Surveillance 
Requirements. The proposed change remains bounded by the existing 
Surveillance Requirement Completion Times and therefore has no impact 
to the margins of safety.
    The proposed change does not involve a change to the plant design 
or operation, and thus does not affect the design of the EDGs, the 
operational characteristics of the EDGs, the interfaces between the 
EDGs, and other plant systems, or the function or reliability of the 
EDGs. Because EDG performance and reliability will continue to be 
ensured by the proposed Technical Specification changes, the proposed 
changes do not result in a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Stuart A. Richards.

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

    Date of amendment request: September 9, 1998.
    Description of amendment request: The proposed license amendment 
concerns hydrostatic (water) testing of containment isolation valves in 
the Feedwater System lines. The proposed technical specification change 
stipulates that water leakage from the feedwater motor-operated 
containment isolation valves will be added into the Primary Coolant 
Sources Outside of Containment Program (Technical Specification 5.5.2), 
and therefore the feedwater check valves do not need to be included in 
the hydrostatic test program addressed by Surveillance Requirement 
3.6.1.3.11. The proposed testing change is based on design and 
licensing basis changes being implemented to improve functioning of the 
Feedwater Leakage Control System.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    (1) This proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    It is proposed that water leakage from the Feedwater motor-operated 
containment isolation valves will be added into the Primary Coolant 
Sources Outside Containment Program (Technical Specification 5.5.2), 
and therefore the Feedwater lines do not need to also be included in 
the hydrostatic test program addressed by Surveillance Requirement 
3.6.1.3.11. The proposed testing change is based on design/licensing 
basis changes being implemented to improve functioning of the Feedwater 
Leakage Control System. The proposed design change will provide 
Feedwater Leakage Control System seal water directly to the bonnets and 
seats of the motor operated gate valves in the Feedwater lines, and 
allow for power to the valves to be provided from redundant power 
supplies.
    The proposed changes do not increase the probability of occurrence 
of an accident previously evaluated because the Feedwater Leakage 
Control System is not an initiator of a previously evaluated accident. 
The Feedwater Leakage Control System is used to

[[Page 56263]]

mitigate the consequences of an event that has already been initiated 
due to some other cause, specifically a design basis Loss of Coolant 
Accident (LOCA). Therefore, changes to the design and testing on the 
Feedwater Leakage Control System have no impact on the probability of 
occurrence of an accident previously evaluated. The Feedwater Leakage 
Control System is a manually initiated system, and the probability of 
an inadvertent initiation remains unchanged from that previously 
reviewed, so the possibility of a loss of feedwater transient is not 
increased.
    The proposed changes do not significantly increase the radiological 
consequences of an accident previously evaluated, because the Feedwater 
lines will continue to be isolated following a LOCA either inside or 
outside of containment. For a line break outside of containment, the 
check valves will provide the necessary short-term closure function to 
prevent significant loss of reactor coolant inventory, as currently 
stated in Updated Safety Analysis Report (USAR) Section 
6.2.4.2.2.1.a.1. The third (gate) valves in the Feedwater line will 
also be available to provide the long-term, high integrity leakage 
protection. The check valves Code Class 1 closure function will be 
verified at an appropriate frequency by performance of an exercise 
closed (EC) test comprised of a visual inspection of the internals of 
the valves, in accordance with the Inservice Testing Program. The 
radiological consequences of such a line break outside of containment 
event are not significant, as there is no postulated fuel damage.
    For a line break inside of containment (a design basis LOCA event), 
the majority of the currently reviewed and accepted licensing basis is 
being maintained. Design changes are being implemented to improve the 
functioning of the Feedwater Leakage Control System. The redundant 
subsystems will be piped to the bonnets of the third, high integrity 
valves in the Feedwater lines (the gate valves) to provide a more rapid 
and effective seal on the stem, bonnet and flexible wedge seats. Water 
leakage from the stem, bonnets and seats of the gate valves will be 
addressed through controls imposed by Technical Specification 5.5.2, 
``Primary Coolant Sources Outside Containment.'' The doses from such 
water leakage are accounted for in the radiological dose calculations. 
Since the leakage from the Feedwater lines is accounted for by the 
Primary Coolant Sources Outside Containment Program, there is no need 
to include the water test results of the Feedwater lines into the 
Surveillance Requirement 3.6.1.3.11 leak test totals.
    The branch lines off of the Feedwater lines will also be addressed 
either through the Primary Coolant Sources Outside Containment Program 
(Technical Specification 5.5.2) or through additional Appendix J air 
leak rate test requirements (Technical Specification Surveillance 
Requirement 3.6.1.1.1 and Specification 5.5.12, ``Primary Containment 
Leakage Rate Testing Program''). The new test methods for these lines 
do not impact the existing radiological dose calculations, since the 
existing leakage limits of the leak rate test programs are not changed 
by the proposal.
    The design changes associated with the Feedwater Leakage Control 
System will continue to satisfy licensing/design criteria for this 
piping to an equivalent degree as the current design. The minor 
exception is where the two Feedwater Leakage Control subsystems tie in 
to the bonnets of the gate valves, and this constitutes only a 
separation issue. Since the Feedwater Leakage Control System piping at 
this juncture is Code Class 2, break excluded, and protected from pipe 
whips and jet impingements, it is considered to be acceptable.
    Addition of the provisions for an alternate power supply to be 
provided to the gate valves (if necessary following a LOCA event) will 
improve the probability of closure of these high integrity valves 
without creating an electrical separation concern. A separation concern 
will not be created since the supply circuitry from the alternate power 
source will be a permanent modification, and physical and electrical 
separation between electrical divisions will be maintained by employing 
two features:
    1. Normally open, fused disconnect switches at both ends of the 
circuit, and
    2. Fuses normally stored out of the circuit.
    Based on the discussions above, it is concluded that neither the 
probability nor the consequences of previously evaluated accidents are 
significantly increased as a result of the proposed changes to the 
Technical Specifications and to the licensing bases for the Feedwater 
penetrations.
    (2) This proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The Feedwater Leakage Control System was developed specifically to 
mitigate the consequences of a design basis LOCA inside the 
containment. The system itself and the proposed changes do not produce 
parameters or conditions that could contribute to the initiation of 
accidents different than those already evaluated in the Updated Safety 
Analysis Report. The proposed changes are intended to improve the 
functioning of the Feedwater Leakage Control System should it be called 
upon following a LOCA. The changes affect mitigation of that previously 
evaluated event.
    In other plant conditions, including normal operation, the system 
is not activated and cannot induce events. Thus, the proposed amendment 
does not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) This proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes only affect the methods used to ensure 
Feedwater Leakage Control System performance and reliability, and 
clarification of the licensing/design basis of the system. The new 
proposed Note in Surveillance Requirement 3.6.1.3.11 clarifies that the 
water leakage from the Feedwater lines does not need to be counted in 
two separate leak test programs. The Primary Coolant Sources Outside 
Containment Program (Technical Specification 5.5.2) will ensure that 
leakage from the Feedwater lines is minimized, and accounted for in an 
appropriate fashion in the radiological does calculations. Leak rate 
testing on the branch lines off of the Feedwater lines will also be 
controlled and limited by existing acceptance criteria for plant 
programs that protect the assumptions of the radiological dose 
calculations. Therefore, the margin of safety provided in the Perry 
Nuclear Power Plant dose calculations will remain unchanged.
    The majority of the existing licensing basis, and therefore the 
margins of safety, are maintained by this proposal. The items that are 
changed are done so to improve the reliability of the system or for an 
administrative clarification. The Feedwater Leakage Control System 
Technical Specification itself (Technical Specification 3.6.1.8) does 
not need revision. The design changes will maintain the existing 
licensing/design criteria, with the minor exception of divisional 
separation at the point that the two divisions have to be piped into 
the bonnets of the third (gate) valve. Since the piping at this 
junction point is Code Class 2, break excluded, and protected from pipe 
whips and jet impingements, it is considered to be acceptable. It will 
not lead to a significant reduction in a margin of safety. The manually 
initiated divisional cross-tie will not create an electrical separation 
concern. The alternate power supply provision will be a permanent

[[Page 56264]]

modification, and physical and electrical separation between electrical 
divisions will be maintained.
    Based on the above discussions, the proposed license amendment is 
concluded to not result in a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Stuart A. Richards.

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: September 17, 1998.
    Description of amendment request: The proposed amendments would 
allow a revision to the Oconee Updated Final Safety Analysis Report 
that addresses potential plant conditions that could occur during 
engineered safeguards functional tests of the emergency electrical 
system. These tests are planned to be performed on Unit 3 in November 
1998, with Unit 3 in the cold shutdown condition, and Units 1 and 2 
operating at power. If an actual loss-of-coolant accident with loss of 
offsite power were to occur on Unit 1 or 2, simultaneously with test 
initiation on Unit 3, the Emergency Power System would be placed in a 
condition outside the present design basis. This involves an unreviewed 
safety question that requires NRC approval before implementation of the 
tests.
    Date of publication of individual notice in Federal Register: 
September 30, 1998 (63 FR 52304).
    Expiration date of individual notice: October 30, 1998.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

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

    Date of amendment request: September 19, 1998.
    Description of amendment request: The amendment would revise 
Section 5.4.8 of the Oyster Creek Nuclear Generating Station Updated 
Final Safety Analysis Report (UFSAR) such that it incorporates the use 
of a freeze seal as a temporary part of the reactor coolant pressure 
boundary.
    Date of publication of individual notice in Federal Register: 
September 30, 1998 (63 FR 52307).
    Expiration date of individual notice: October 30, 1998.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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 amendment requests: January 29, 1997, as supplemented 
February 11, 12, March 7, 10, 11, 19, 20, April 29, June 30, and July 
10, 1997, June 20, June 22, July 24 and September 15, 1998.
    Brief description of amendment request: The proposed amendments 
would change the design basis of the cooling water system emergency 
intake line flow capacity. The licensee determined through testing that 
the emergency intake line flow capacity was less than the design value 
stated in the Updated Safety Analysis Report. The proposed changes 
reflect the use of operator actions to control cooling water system 
flow following a seismic event. The proposed changes also reclassify 
the intake canal for use during a seismic event, which would be an 
additional source of cooling water during a seismic event.
    Date of publication of individual notice in Federal Register: 
October 1, 1998 (63 FR 52772).
    Expiration date of individual notice: November 2, 1998.
    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 amendment request: October 14, 1997, as supplemented July 
23, 1998.
    Description of amendment request: The amendment would update the 
Technical Specifications to provide for installation of additional 
racks to increase spent fuel storage capacity, and to correct the 
maximum exposure dependent, infinite lattice multiplication factor for 
fuel bundles.
    Date of publication of individual notice in Federal Register: 
August 24, 1998 (63 FR 45096).
    Expiration date of individual notice: September 23, 1998.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: August 8, 1997, as supplemented by 
letters dated March 9, May 6, July 6, July 31, September 4, and 
September 11, 1998, and advanced information related to the application 
submitted April 17, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to accommodate an increase in the 
maximum licensed thermal power level from 2558 megawatts thermal (MWt) 
to 2736 MWt.
    Date of publication of individual notice in Federal Register: 
October 6, 1998 (63 FR 53730).
    Expiration date of individual notice: November 5, 1998.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.

[[Page 56265]]

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

    Date of amendment request: September 4, 1998.
    Description of amendment request: The amendment would revise the 
Technical Specifications to reflect an increase in the spent fuel 
storage capacity.
    Date of publication of individual notice in Federal Register: 
October 1, 1998. (63 FR 52774)
    Expiration date of individual notice: November 2, 1998.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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.

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

    Date of application for amendment: June 13, 1995, as supplemented 
by letters dated August 16, 1995, June 9, 1998, and September 6, 1998.
    Brief description of amendment: These amendments revise TS 3.5.1, 
``Safety Injection Tanks (SITs)--Operating,'' and TS 3.5.2, ``Safety 
Injection Tanks--Shutdown,'' to extend the allowed outage times for the 
SITs.
    Date of issuance: October 2, 1998.
    Effective date: October 2, 1998.
    Amendment No.: 118.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendment revised the Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54715)
    The June 9, 1998, and September 6, 1998, letters provided 
additional clarifying information and do not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.

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

    Date of application for amendment: June 12, 1997, as supplemented 
by letter dated August 27, 1998. The August 27, 1998, supplemental 
letter provided clarifying information only, and did not change the 
initial no significant hazards consideration determination.
    Brief description of amendment: This amendment changes the 
description of the Harris Nuclear Plant Operations organization in TS 
6.0, ``Administrative Controls.''
    Date of issuance: October 7, 1998.
    Effective date: October 7, 1998.
    Amendment No: 83.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40847).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 7, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 
and 2, Will County, Illinois

    Date of application for amendments: December 30, 1997.
    Brief description of amendments: The amendments change the 
Technical Specifications for the condensate storage tank (CST) level 
and the automatic auxiliary feedwater pump switchover from the suction 
of the CST to the essential service water system.
    Date of issuance: October 6, 1998.
    Effective date: October 6, 1998.
    Amendment Nos.: 104; 104 & 96; 96.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998. (63 
FR 9596)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 6, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

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

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

    Date of application for amendments: May 18, 1998.
    Brief description of amendments: The amendments will change several 
Technical Specification (TS) values to reflect design values. These TS 
values affect (1) 125/250 volts direct current (Vdc) electrolyte 
temperature; (2) control rod drive accumulator pressure; (3) standby 
liquid control solution temperature; (4) ultimate heat sink minimum 
water level; (5) shutdown

[[Page 56266]]

suppression chamber level (Quad Cities only); and (6) a degraded 
voltage setpoint (Quad Cities only).
    Date of issuance: October 8, 1998.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: Dresden 169 & 164; Quad Cities 181 & 179.
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 17, 1998 (63 FR 
33105).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina

    Date of application for amendment: August 6, 1998.
    Brief description of amendment: The amendment deletes Surveillance 
Requirement 4.8.1.1.2.i.2, regarding diesel fuel oil system pressure 
testing, from the Technical Specifications on the basis that the staff 
had previously approved alternative surveillance based on Code Case N-
498-1 of the American Society of Mechanical Engineers.
    Date of issuance: September 28, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment No.: 164.
    Facility Operating License No. NPF-52: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1998 (63 FR 
43962).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of application for amendments: October 22, 1996, as 
supplemented by letters dated March 19, July 6, and September 15, 1998.
    Brief description of amendments: The amendments allow continued 
plant operation at elevated Containment Lower Compartment temperatures 
between 125  deg.F and 135  deg.F for a period not to exceed 72 
cumulative hours per calendar year.
    Date of issuance: September 28, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-183; Unit 2-165.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6574).
    The March 19, July 6, and September 15, 1998, submittals provided 
clarifying information that did not change the scope of the October 22, 
1996, 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 September 28, 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.

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

    Date of application of amendment: July 21, 1998.
    Brief description of amendment: The amendment permits an 
alternative to the requirement to perform Control Rod Drive scram time 
testing with the reactor pressurized prior to resuming power operation. 
The change permits: (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.
    Date of Issuance: October 1, 1998.
    Effective date: October 21, 1998, to be implemented within 30 days.
    Amendment No.: 198.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 12, 1998 (63 FR 
43204).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated October 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

    Date of application for amendment: March 23, 1998, as supplemented 
June 30, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.1.2, to incorporate new pressure/temperature 
limits for reactor vessel pressurization heatup, cooldown, and 
inservice leak and hydrostatic test.
    Date of issuance: October 5, 1998.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 208.
    Facility Operating License No. DPR-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19970). The June 30, 1998, submittal provided additional information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

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

    Date of application for amendment: April 15, 1998.
    Brief description of amendment: The amendment revises the Technical 
Specifications by updating the existing pressure-temperature curves 
with new curves with values from 18 to 32 effective full power years. 
Applicable surveillance requirements are also revised to reflect 
operation with the new curves.
    Date of issuance: October 1, 1998.
    Effective date: October 1, 1998.
    Amendment No.: 224.
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25110).

[[Page 56267]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, IA 52401.

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

    Date of application for amendment: May 20, 1998, as supplemented 
July 17 and August 6, 1998.
    Brief description of amendment: The amendment provides for 
automatic operation of a new emergency reserve auxiliary transformer to 
provide power to the plant 4.16-kV buses from the offsite 138-kV 
transmission network.
    Date of issuance: October 1, 1998.
    Effective date: October 1, 1998.
    Amendment No.: 116.
    Facility Operating License No. NPF-62: The amendment authorized 
revision of the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: June 4, 1998 (63 FR 
30519).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: May 23, 1997, as supplemented 
by letter dated September 11, 1998. The September 11, 1998, letter 
provided the typed TS pages that did not change the Nuclear Regulatory 
Commission staff's proposed no significant hazards consideration 
determination.
    Brief description of amendments: The proposed amendments would 
revise the Technical Specifications TSs to exclude the Main Steam 
Isolation Valves leakage from the total Type B and Type C local leak 
rate test results.
    Date of issuance: October 1, 1998.
    Effective date: The amendments are effective as of the date of 
issuance, and are to be implemented within 30 days from the date of 
their issuance.
    Amendments Nos.: 223 and 227.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 2, 1998 (62 FR 
35852).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

PECO Energy Company, Public Service Electric and Gas Company Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: May 1, 1998, as supplemented by 
letter dated September 11, 1998.
    Brief description of amendments: These amendments revise the 
technical specifications to delete the requirements for functional 
testing of safety relief valves during each unit startup.
    Date of issuance: October 5, 1998.
    Effective date: As of the date of issuance and is to be 
implemented, Unit 2, prior to October 1998 refueling outage and Unit 3, 
prior to October 1999 refueling outage.
    Amendments Nos.: 224 and 228.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 29, 1998 (63 FR 
40559).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

PECO Energy Company, Public Service Electric and Gas Company Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: March 20, 1998, as supplemented 
by letters dated June 26, August 11, and September 14, 1998. The August 
11 an September 14 letters provided clarifying information that did not 
change the initial proposed no significant hazards consideration 
determination.
    Brief description of amendments: These amendments would revise the 
Technical Specifications to permit incorporation of end-of-cycle 
recirculation pump trip systems.
    Date of issuance: October 5, 1998.
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendments Nos.: 225 and 229.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 29, 1998 (63 FR 
40558).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: March 24, 1997, as supplemented 
September 4, 1998.
    Brief description of amendments: These amendments approve the 
deletion of the Drywell and Suppression Chamber Purge System 
operational time limit, removal of a footnote regarding 1-inch and 2-
inch valves, and the addition of a surveillance requirement ensuring 
the purge system large supply and exhaust valves are closed as 
required.
    Date of issuance: October 1, 1998.
    Effective date: Units 1 and 2, As of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 130 and 91.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30643).
    The September 4, 1998, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

[[Page 56268]]

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

    Date of application for amendment: November 13, 1997.
    Brief description of amendment: The amendment changes the Technical 
Specifications by increasing the minimum test frequency for main 
turbine stop valves.
    Date of issuance: October 5, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 182.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38203).
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.

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

    Date of application for amendment: May 13, 1998.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 3/4.10.8, ``Inservice Leak and Hydrostatic 
Testing,'' to delete the requirement for an operable High Drywell 
Pressure trip function. Specifically, TS 3.10.8.a is being revised to 
remove the reference to the Secondary Containment Isolation Actuation 
Instrumentation trip function 2.b.
    Date of issuance: October 1, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 112.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 1, 1998 (63 FR 
35994).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of amendment request: December 31, 1997, as supplemented June 
30, August 6, August 18, and August 27, 1998.
    Brief description of amendments: The amendments revised TS 2.1 
(Safety Limits), 2.2 (Limiting Safety System Settings), and 3/4.2.5 
(Departure from Nucleate Boiling Parameters) by including alternate 
operating criteria to allow continued plant operation with a reduced 
measured reactor coolant system flow rate, if necessary.
    Date of issuance: September 29, 1998.
    Effective date: September 29, 1998.
    Amendment Nos.: Unit 1--Amendment No. 97; Unit 2--Amendment No. 84.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4325).
    The additional information contained in the supplemental letters 
dated June 30, August 6, August 18, and August 27, 1998, were 
clarifying in nature and thus, within the scope of the initial notice 
and did not affect the staff's proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

    Date of application for amendment: April 18, 1997, as supplemented 
by letters dated October 10, 1997, and February 27 and September 8, 
1998.
    Brief description of amendment: This amendment revises TS Section 
3/4.7.6, ``Plant Systems--Control Room Emergency Ventilation System,'' 
and the associated bases. Action statements have been added related to 
the availability of the station vent normal range radiation monitoring 
instrumentation. The bases have been modified consistent with these 
changes.
    Date of issuance: October 5, 1998.
    Effective date: October 5, 1998.
    Amendment No.: 227.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30646). The supplemental information submitted by letters dated October 
10, 1997, and September 8, 1998, did not affect the proposed no 
significant hazards consideration. However, the supplemental letter 
dated February 27, 1998, included a new analysis of the issue of no 
significant hazards consideration. Based on this, the Commission issued 
a new proposed finding that the amendment involves no significant 
hazards consideration (63 FR 25117). The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.

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

    Date of application for amendment: August 8, 1997, as supplemented 
by letters dated December 16, 1997, January 20, 1998, March 4, 1998, 
March 17, 1998, June 29, 1998, and July 28, 1998.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7-2 to specify that the lift setting tolerance for 
the main steam line safety valves is +3/-1 percent as-found and +/-1 
percent as-left. The amendment also revised TS Table 2.2-1 to reduce 
the sensor error for the pressurizer pressure-high trip.
    Date of issuance: October 2, 1998.
    Effective date: October 2, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 128.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66144).
    The December 16, 1997, January 20, 1998, March 4, 1998, March 17, 
1998, June 29, 1998, and July 28, 1998, supplemental letters provided 
additional clarifying information and did not change the initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated October 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Elmer Ellis Library, 
University of Missouri, Columbia, Missouri 65201-5149.

[[Page 56269]]

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

    Date of application for amendment: May 14, 1998, supplemented July 
3, August 27, and October 1, 1998.
    Brief description of amendment: This amendment redefines the 
pressure boundary for Westinghouse mechanical hybrid expansion joints 
(HEJs) in sleeved steam generator tubes. TS 4.2 b, ``Steam Generator 
Tubes,'' is changed to incorporate a length criterion to allow tubes 
with degraded HEJ sleeves to remain in service if a minimum length of 
the HEJ is free of flaws.
    Date of issuance: October 2, 1998.
    Effective date: October 2, 1998.
    Amendment No.: 139.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 3, 1998 (63 FR 
30269).
    The July 3, August 27, and October 1, 1998 submittals provided 
clarifying information within the scope of the original Federal 
Register notice and did not change the staff's initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001

    Dated at Rockville, Maryland, this 14th day of October 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director Division of Reactor Projects--III/IV, Office of Nuclear 
Reactor Regulation.
[FR Doc. 98-28069 Filed 10-20-98; 8:45 am]
BILLING CODE 7590-01-P