[Federal Register Volume 63, Number 57 (Wednesday, March 25, 1998)]
[Notices]
[Pages 14482-14497]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-7652]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any

[[Page 14483]]

amendments issued, or proposed to be issued, under a new provision of 
section 189 of the Act. This provision grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 2, 1998, through March 13, 1998. The 
last biweekly notice was published on March 11, 1998 (63 FR 11913).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By April 24, 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

[[Page 14484]]

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. 50-325, Brunswick 
Steam Electric Plant, Unit 1, Brunswick County, North Carolina

    Date of amendment request: February 23, 1998.
    Description of amendment request: The amendment request proposes 
changes to the Brunswick Steam Electric Plant Unit 1 Technical 
Specifications (TS) in support of Cycle 12 operation, including a 
change to the Minimum Critical Power Ratio safety limit (safety limit 
MCPR) to a value equivalent to the generic safety limit MCPR for 
General Electric type GE-13 fuel. The request would additionally remove 
a footnote limiting the stated value for the safety limit MCPR to a 
specific fuel cycle and reference to an NRC safety evaluation 
documenting acceptance of methods used for determining the current 
cycle safety limit MCPR. The amendment request is provided both in the 
format of the current TS as well as improved Standard Technical 
Specifications (iSTS). The Brunswick licensee applied for conversion to 
ISTS on November 1, 1996, as supplemented on October 13, 1997, and 
February 26, 1998, and that application is currently undergoing NRC 
staff review. For iSTS, the licensee has proposed two safety limits 
MCPR, one pertaining to two-recirculation loop operation and the other 
to single-recirculation loop operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed license amendment establishes a revised safety 
limit MCPR value of 1.09 [two-recirculation loop and 1.10 for 
single-recirculation loop operation] for use during Unit 1 Cycle 12 
operation. General Electric (GE) has determined that both generic 
and plant-specific evaluations [two-loop operation] yield the same 
calculated safety limit MCPR value. Additionally, a document 
referenced by the Technical Specification 6.9.3.2 of methodologies 
used in determining core operating limits is being removed.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established, consistent with NRC[-] 
approved methods, to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable.
    The probability of an evaluated accident is not increased by 
revising the safety limit MCPR value to 1.09 [two-loop/1.10 single-
loop]. The change does not require any physical plant modifications 
or physically affect any plant components. Therefore, no individual 
precursors of an accident are affected.
    The proposed license amendment establishes a revised safety 
limit MCPR that ensures the fuel is protected during normal 
operation and during any plant transients or anticipated operational 
occurrences. Specifically, the reload analysis demonstrates that a 
safety limit MCPR value of 1.09 [two-loop/1.10 single-loop] ensures 
that less than 0.1 percent of the fuel rods will experience boiling 
transition during any plant operation if the limit is not violated.
    The methods for calculating the safety limit MCPR have been 
approved by the NRC and are described in GE's reload licensing 
methodology topical report NEDE-24011, ``General Electric Standard 
Application for Reactor Fuel (GESTAR II).'' Based on (1) the 
determination of the new safety limit MCPR value using conservative 
approved methods, and (2) the operability of plant systems designed 
to mitigate the consequences of accidents not having been changed; 
the consequences of an accident previously evaluated have not been 
increased.
    Additionally, removal of the footnote on the safety limit MCPR 
value in Technical Specification 2.1.2 and removal of reference 
``c'' from the document list in Technical Specification 6.9.3.2 will 
not increase the probability or consequences of accidents previously 
evaluated. The footnote on the safety limit MCPR value in Technical 
Specification 2.1.2 and reference ``c'' in Technical Specification 
6.9.3.2 were associated with the safety limit MCPR value of 1.10 for 
Unit 1 Cycle 11 operation. Since the current safety limit MCPR value 
of 1.10 applies only to Unit 1 Cycle 11 operation, the footnote on 
the safety limit MCPR value in Technical Specification 2.1.2 and the 
reference ``c'' in Technical Specification 6.9.3.2 are no longer 
needed and should be deleted. Thus, removal of the footnote on the 
safety limit MCPR value in Technical Specification 2.1.2 and removal 
of reference ``c'' from Technical Specification 6.9.3.2 is an 
administrative change that has no effect on the probability or 
consequences of accidents previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed license amendment involves a revision of the 
safety limit MCPR from 1.10 to 1.09 [two-loop/1.10 single-loop] 
based on the results of both cycle-specific and generic analyses, 
removal of the footnote on the safety limit MCPR value in Technical 
Specification 2.1.2, and the removal of a document reference listed 
in Technical Specification 6.9.3.2 describing the methods used only 
during Unit 1 Cycle 11 to determine core operating limits. Creation 
of the possibility of a new or different kind of accident would 
require the creation of one or more new precursors of that accident. 
New accident precursors may be created by modifications of the plant 
configuration, including changes in allowable modes of operation. 
This proposed license amendment does not involve any modifications 
of the plant configuration or changes in the allowable modes of 
operation. Therefore, no new precursors of an accident are created 
and no new or different kinds of accidents are created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    As previously stated, the methods for calculating the safety 
limit MCPR have been previously approved by the NRC and are 
described in GE's reload licensing methodology topical report NEDE-
24011. Use of these methods ensures that the resulting safety limit 
MCPR satisfies the fuel design safety criteria that less than 0.1 
percent of the fuel rods experience boiling transition if the safety 
limit is not violated. Based on the assurance that the fuel design 
safety criteria will be met, the proposed license 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

[[Page 14485]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Pao-Tsin Kuo (Acting).

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: May 16, 1997.
    Description of amendment request: The proposed changes would 
replace the existing Technical Specification (TS) 4.6.2.3 a.2 cooling 
water flow rate of 1425 gpm with a new value of 1300 gpm.
    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.
    Cooling water flow to the Containment Fan Coolers is provided by 
the Emergency Service Water (ESW) System, and Emergency Service 
Water is not an initiating system in any FSAR [Final Safety Analysis 
Report] Chapter 15 analyses. Revising the minimum cooling water flow 
to the Containment Fan Coolers will not increase the probability of 
initiating any previously evaluated accident, because Containment 
Fan Cooler performance and integrity will not be adversely affected. 
The heat removal capacity of the Containment Fan Coolers will be 
maintained consistent with the assumptions used in the existing HNP 
[Harris Nuclear Plant] containment analyses, and, therefore, 
containment integrity should not be challenged.
    Therefore, there would be no 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.
    The proposed amendment will not create any new accident 
scenarios, because the change does not introduce any new single 
failures, adverse equipment or material interactions, or release 
paths.
    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.
    Although the proposed amendment replaces the TS 4.6.2.3 a.2 
cooling water flow rate of 1425 gpm with a lower flow rate of 1300 
gpm, a cooling water flow rate of greater than or equal to 1300 gpm 
maintains adequate heat removal capacity as required by existing HNP 
containment analyses. The Bases for TS 4.6.2.3 a.2 is to ensure that 
adequate heat removal capacity is available, when the Containment 
Fan Coolers are operated in conjunction with the Containment Spray 
Systems, during post-LOCA [Loss-of-Coolant Accident] conditions to 
prevent the pressure inside containment from exceeding its design 
rating.

    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).

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit No. 2, Shippingport, Pennsylvania

    Date of amendment request: October 22, 1997
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) by reducing the reactor 
coolant system (RCS) specific activity limits in accordance with 
Generic Letter 95-05. The definition of DOSE EQUIVALENT I-131 would be 
replaced with the Improved Standard TS definition wording in the first 
sentence and an equation added based on dose conversion factors derived 
from International Commission on Radiation Protection (ICRP) ICRP-30. 
TS 3.4.8, Specific Activity, would be revised by reducing the DOSE 
EQUIVALENT I-131 limit from 1.0 [micro] Ci[curies]/gram to 0.35 
[micro]Ci[curies]/gram. Item 4.a in TS Table 4.4-12, Primary Coolant 
Specific Activity Sample and Analysis Program, TS Figure 3.4-1, and the 
Bases for TS 3/4.4.8 would be modified to reflect the reduced DOSE 
EQUIVALENT I-131 limit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change reduces the reactor coolant system (RCS) 
specific activity limits of Specification 3.4.8 from 1.0 [micro]Ci/
gram to 0.35 [micro]Ci/gram and lowers the graph in Figure 3.4-1 by 
39 [micro]Ci/gram following the guidance provided in Generic Letter 
(GL) 95-05. This reduces the RCS acvitity allowed to leak to the 
secondary side when the plant is operating so that additional margin 
is available to support a higher allowable accident-induced leakage 
value as justified by analysis.
    The proposed changes to Specification 3.4.8 and the definition 
of DOSE EQUIVALENT I-131 ensure these requirements are consistent 
with the latest analyses.
    These changes implement the more restrictive RCS activity limits 
in accordance with applicable analyses and GL 95-05 to ensure the 
regulations are satisfied. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not alter the configuration of the 
plant or affect the operation with the reduced specific activity 
limit. By reducing the specific activity limit, the limit would be 
reached sooner to initiate evaluation of the out of limit condition. 
The proposed changes will not result in any additional challenges to 
the main steam system or the reactor coolant system pressure 
boundary. Consequently, no new failure modes are introduced as a 
result of the proposed changes. As a result, the main steam line 
break, steam generator tube rupture and loss of coolant accident 
analyses remain bounding. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change reduces the RCS specific activity limit to 
0.35 [micro]Ci/gram along with lowering the Figure 3.4-1 limits by 
39 [micro]Ci/gram. Reduction of the RCS specific activity limits 
allows an increase in the limit for the projected SG [steam 
generator] leakage following SG tube inspection and repair in 
accordance with the voltage-based SG tube alternate repair criteria 
(ARC). This follows the guidance provided in GL 95-05 and 
effectively takes margin available in the specific activity limits 
and applies it to the projected SG leakage for the ARC. This has 
been determined to be an acceptable means for accepting higher

[[Page 14486]]

projected leakage rates while still meeting the applicable limits of 
10 CFR [Part] 100 and GDC [General Design Criterion] 19 with respect 
to offsite and control room doses.
    The capability for monitoring the specific activity and 
complying with the required actions remains unchanged. In addition, 
there is no resultant change in dose consequences. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety.

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

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

    Date of amendment request: March 3, 1998.
    Description of amendment request: The licensee proposed to revise 
Section 6.2.3.2 of the units' Technical Specifications. Currently, this 
section prescribes that the Catawba Safety Review Group (SRG) be 
composed of at least five individuals and at least three of these shall 
have a bachelor's degree in engineering or related science and at least 
2 years professional level experience in his/her field, at least 1 year 
of which experience shall be in the nuclear field. The licensee 
proposed to revise this section to provide the option of replacing one 
of the three degreed individuals with one with at least 15 years of 
professional level experience in his/her field, at least 10 years of 
which experience shall be in the nuclear field, at least 3 years of 
which nuclear experience shall be supervisory/managerial experience in 
engineering, and shall hold or have held a Senior Reactor Operator 
license. The licensee also proposed to editorially revise this section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below.

    1. Would the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed amendment would only change administrative 
requirements related to personnel qualifications for one of the five 
SRG [Safety Review Group] positions. The SRG is an oversight group, 
and the individual who meets the new qualification requirements 
would be expected to perform at the same level of quality as an 
individual who meets the current qualification requirements. 
Changing qualification requirements for an individual who primarily 
performs an oversight function will not have any direct effect on 
the design or operation of any plant structures, systems, or 
components. No previously analyzed accidents were initiated by the 
functions of the SRG, and the SRG was not a factor in the 
consequences of previously analyzed accidents. Therefore, the 
proposed change would have no impact on the consequences or 
probabilities of any previously evaluated accidents.
    2. Would the change create the possibility of a new or 
difference kind of accident from any accident previously evaluated?
    No. The proposed change would not lead to any hardware or 
operating procedure change. Hence, no new equipment failure modes or 
accidents from those previously evaluated will be created.
    3. Would the change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is associated with confidence in the design 
and operation of the plant. The proposed change to the Technical 
Specifications does not involve any change to plant design or 
operation. Thus, the margin of safety previously analyzed and 
evaluated is maintained.

    Based on this analysis, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: February 7, 1997.
    Description of amendment request: The proposed amendment, if 
approved, would revise Technical Specification (TS) as delineated 
below:
    1. 4160 Volt Tie From Unit 2.
    TS sections 3.7.2.b & d to delete reference to the optional use of 
the 4160 volt tie from the unit 2 transformer.
    2. Emergency Load Sequence and Power Transfer.
    a. The testing required by Section 4.5.1.1.b of the TS would be 
considered satisfactory if the pumps have started and valves have 
completed travel. The need to evidence the successful starting of pumps 
and fans and the complete travel of valves by observation of control 
board component operating lights will be deleted. Neither would a 
second means of verification, such as: the station computer or control 
board indicating lights initiated by separate limit switch contacts be 
required.
    b. Section 4.5.1.2.b would be revised in the same manner as 
4.5.1.1.b above.
    3. Reactor Building Cooling and Isolation System.
    a. Section 4.5.3.1.a.1 of the TS would be revised to delete the 
need to simultaneously test start a spray pump using a Reactor Building 
30-psi high pressure test signal while testing the emergency loading 
sequence.
    The proposed change also eliminates the need to evidence the 
successful starting of the spray pumps by observation of the control 
board indicating lights or the use of the station computer for Sections 
4.5.3.1.a.1 and 4.5.3.1.b.2.
    4. Instrument Surveillance Requirements.
    Table 4.1-1 of the TS would be revised to delete the strong motion 
accelerometer and its quarterly battery check surveillance requirement.
    5. Air Intake Tunnel (AIT) Fire Protection Systems.
    Section 5.5 of the TS would be deleted. The description of the 
equipment contained in Section 5.5 would be transferred to the Final 
Safety Analysis Report (FSAR).
    6. Hydrogen Recombiner System.
    The Bases for Section 4.4.4 TS would be changed to reflect a 
reduction in the time interval for operation of the hydrogen recombiner 
following a loss of cooling accident (LOCA) from 9.8 to 9 days.
    7. Various editorial and typographical errors would be corrected.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The revised TS eliminate overly prescriptive 
requirements for evidencing component performance, the requirement 
for redundant diesel block loading tests,

[[Page 14487]]

instrumentation from SR [surveillance requirement] tables having no 
associated LCO [limiting condition for operation], AIT fire 
protection systems descriptive text, and correct previous 
typographical errors. Several of the proposed revisions involve 
changes which are consistent with NUREG-1430, the Revised Standard 
Technical Specifications (RSTS) for B&W plants. The reliability of 
systems and components depended upon to prevent or mitigate the 
consequences of accidents previously evaluated is not degraded by 
the proposed changes because assurance of system and equipment 
availability is maintained by surveillance testing program 
requirements.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The revised 
surveillance requirements create no new failure modes. Verification 
of equipment operation continues to be required by plant procedures. 
Elimination of the AIT fire protection system descriptive text from 
the TSs would not create a new or different kind of accident since 
the change has no effect on surveillance methodology and frequency 
requirements. They are maintained in the Fire Protection Program.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety because no operating limits are affected.

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

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone 
Nuclear Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: March 3, 1998.
    Description of amendment request: The proposed revision to the 
Millstone Unit 3 licensing basis would eliminate the requirement to 
have the recirculation spray system directly inject into the reactor 
coolant system following a design basis accident.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Northeast Nuclear Energy Company (NNECO) has reviewed the 
proposed revision in accordance with 10CFR50.92 and has concluded 
that the revision does not involve a significant hazards 
consideration (SHC). The basis for this conclusion is that the three 
criteria of 10CFR50.92(c) are not satisfied. The proposed revision 
does not involve an SHC because the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The change to the Emergency Operating Procedures (EOP) to 
eliminate the use of Recirculation Spray System (RSS) direct 
injection during cold and hot leg recirculation does not effect the 
probability of any accident. The elimination of the requirement to 
have RSS directly [inject] into the reactor coolant system did not 
increase the consequences of the previously evaluated accidents. 
These consequences were evaluated based on very conservative 
assumptions concerning the containment pressure after the design 
basis Loss of Coolant Accident (LOCA), containment integrated 
leakage rates, and the fraction of the sprayed volume. None of these 
assumptions were affected by the elimination of the direct cold-leg 
injection.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The modification to the RSS did not create the possibility of a 
new or different accident from those previously analyzed. The change 
involved elimination of the direct injection flow path from the 
design basis of the system but did not involve physical 
modifications to the system itself. The operability of the affected 
valves within the direct injection alignments remained unchanged and 
these paths were still available to the operators for contingencies 
beyond the design basis. The EOPs provided clear and explicit 
guidance to that effect.
    Therefore, the proposed revision does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    In considering the impact on the margin of safety as defined in 
the bases of the Technical Specifications, the impact of the change 
on the design basis analysis of the fission product barriers must be 
evaluated.
    The minimum Emergency Core Cooling System flow requirement for 
long-term core cooling is that the modified alignment deliver 
sufficient flow to satisfy the inventory lost to the boil off in the 
vessel due to the decay heat and the extended boiling from hot metal 
in the downcomer and the lower plenum. The analysis determined that 
these requirements were being met.
    The elimination of the direct injection resulted in a flow 
reduction through the RSS heat exchanger, from approximately 4000 
gpm [gallons per minute] to 1200 gpm, thus reducing the rate of the 
heat transfer from the containment to the service water system. The 
design basis of the containment heat removal systems (circa 1986) is 
that the containment pressure will decrease to subatmospheric within 
one hour after the Design Basis Accident to compensate for the 
reduction in heat removal from the containment, a smaller allowable 
RSS pump degradation was assumed in the revised containment 
analysis. The original RSS pump performance curve was based on a 10 
percent reduction in developed head from the design curve. For the 
modification, a 5 percent reduction was used. The results of the 
analysis show that with these changes the design basis of 
maintaining subatmospheric containment pressure was met.
    Based on the above, elimination of the direct injection did not 
reduce the margin of safety because there was no violation of the 
acceptance limits and no weakening of the protective boundaries.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve an SHC.

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

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of amendment requests: November 2, 1995, as supplemented by 
letter dated January 9, 1998. The January 9, 1998, submittal supersedes 
the staff's proposed no significant hazards consideration determination 
evaluation for the requested changes that was published on April 10, 
1996 (61 FR 15995).
    Description of amendment requests: In the November 2, 1995, letter, 
the

[[Page 14488]]

licensee proposed to revise Technical Specification (TS) 3.8.1, ``AC 
Sources--Operating,'' to extend the offsite circuit completion time and 
to extend the allowed outage time for an emergency diesel generator. 
The January 9, 1998, letter modifies the original request to (1) 
further extend the offsite completion time and allowed outage time for 
an emergency diesel generator, and (2) add a new TS 5.5.2.14, 
``Configuration Risk Management Program,'' that ensures a 
proceduralized probabilistic risk assessment-informed process is in 
place that assesses the overall impact of plant maintenance on plant 
risk.
    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 Emergency Diesel Generators (EDGs) are backup alternating 
current power sources design to power essential safety systems in 
the event of a loss of offsite power. EDGs are not accident 
initiators in any accident previously evaluated. Therefore, this 
change does not involve an increase in the probability of an 
accident previously evaluated.
    The EDGs provide backup power to components that mitigate the 
consequences of accidents. The proposed changes to the Completion 
Times do not affect any of the assumptions used in the deterministic 
safety analysis.
    To fully evaluate the effect of the EDG Completion Time 
extension, Probabilistic Safety Analysis (PSA) methods were 
utilized. The results of these analyses show no significant increase 
in the core damage frequency. As a result, there would be no 
significant increase in the consequences of accidents previously 
evaluated.
    The Configuration Risk Management Program is an Administrative 
Program that assesses risk based on plant status. Adding the 
requirement to implement this program for Technical Specification 
3.8.1 does not affect the probability or the consequences of an 
accident.
    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 alter the design, configuration, 
or method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the Limiting Conditions for 
Operation or their Bases that are used in the deterministic analyses 
to establish the 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.
    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 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: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of amendment requests: December 19, 1997.
    Description of amendment requests: The licensee proposed to revise 
Technical Specification (TS) 3.4.9, ``Pressurizer,'' to reduce the 
allowable pressurizer water volume for pressurizer operability.
    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 limiting events impacted by this Technical Specification 
change have been reanalyzed. These events are the Chemical and 
Volume Control System (CVCS) Malfunction and CVCS Malfunction With a 
Concurrent Single Failure of an Active Component, Inadvertent 
Operation of the Emergency Core Cooling System (ECCS) During Power 
Operation (Including Single Failure of an Active Component), and 
Feedwater System Pipe Breaks. The probability of these events is not 
changed by the restriction of the pressurizer level to 57%. An 
operator action time of 15 minutes has been identified for the CVCS 
malfunction and inadvertent ECCS operation events. Based on the 
availability of operator alarms and indications and operator 
Simulator training, 15 minute operator action is sufficient to 
recognize and mitigate the inadvertent CVCS or ECCS operation. 
Therefore, this change will not involve an increase in the 
probability or consequences of any previously evaluated accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This amendment request does not involve any change to plant 
equipment or operation. All the events identified in Chapter 15 of 
the Updated Final Safety Analysis Report (UFSAR) were evaluated to 
determine the impact of the change in pressurizer level. In addition 
to the normally analyzed Inadvertent Operation of the ECCS During 
Power Operation event a concurrent single failure of an active 
component was considered in this evaluation. The analysis of this 
event with single failure of an active component produced 
consequences that are bounded by the CVCS malfunction with single 
failure of an active component. No new or different kind of accident 
will be created as a result of this Technical Specification change. 
Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This amendment request does not change the manner in which 
safety limits, limiting safety settings, or limiting conditions for 
operation are determined. There are no changes to the acceptance 
criteria for these events as a result of the proposed reduction in 
the maximum pressurizer water level. This change does not reduce a 
margin of safety since it lowers allowed pressurizer operational 
level to 57%. An operator action time of 15 minutes has been 
identified for the CVCS malfunction and inadvertent ECCS operation 
events. Based on the availability of operator alarms and 
indications, and demonstrated operator response in Simulator 
training, 15 minute operator action has been demonstrated to be 
adequate to recognize and mitigate the inadvertent CVCS or ECCS 
operation. Therefore, this proposed change does not involve a 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

[[Page 14489]]

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of amendment requests: January 2, 1998.
    Description of amendment requests: The licensee proposed to revise 
Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW) 
System,'' to indicate the turbine driven AFW pump is operable when 
running in the manual mode to support plant startups, shutdowns, and 
testing.
    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.
    Probabilistic analyses have been performed in support of 
declaring P140 operable when the pump is manually actuated and 
operating.
    The results show that, considering P-140 to be in test for an 
entire year, the core damage risk of a Main Steam Line Break/
Feedwater Line Break (MSLB/FWLB) slightly increases (4.3E-8/yr) 
while the risk due to other initiating events decreases (3E-7/yr). 
The net core damage impact of P-140 in test for an entire year is a 
Core Damage Frequency (CDF) decrease of 2E-7/yr. Having P140 
operating instead of being in standby increases its reliability. 
This increased reliability reduces the risk due to other initiating 
events, such as loss of main feedwater, medium and small Loss of 
Coolant Accidents (LOCAs), Steam Generator Tube Rupture (SGTR), and 
Loss of Offsite Power (LOP), which require Auxiliary Feedwater (AFW) 
and which occur with much greater frequency than MSLB/FWLB. With the 
overall CDF reduction a result of considering P140 being in a test 
configuration for an entire year, the actual cumulative risk 
incurred is the weighted fraction that P140 is in the test 
configuration over a year period. Based on past experience, the pump 
is running in manual approximately 500 minutes/year, which results 
in an annual net cumulative CDF reduction on the order of 2E-10/yr 
due to running P140 in the manual mode.
    Therefore, the operation of the facility in accordance with this 
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.
    This change does not involve a plant hardware modification or 
allow the operation of any plant equipment in any way other than 
originally designed. This change only affects the administrative 
tracking of the turbine-driven AFW pump when the steam driven AFW 
pump is operating in the manual mode.
    Therefore, the operation of the facility in accordance with this 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Pump history shows the pump is run approximately 500 minutes per 
year. In all cases except for the one postulated scenario of the 
Main Steam Isolation Signal followed by an Emergency Feedwater 
Actuation Signal the turbine-driven AFW pump is not susceptible to 
being tripped. Also, this postulated scenario does not affect the 
capability of the motor-driven AFW pumps.
    Even though there is a small increase in the CDF from the AFW 
steam driven pump operating in manual mode based on the possibility 
of a MSLB/FWLB, also considering other initiating events results in 
an annual net cumulative CDF reduction on the order of 2E-10/yr due 
to P140 running in the manual mode.
    Therefore, the operation of the facility in accordance with this 
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 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: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: August 20, 1997, as supplemented by 
letters dated September 18, 1997 and October 31, 1997.
    Description of amendment request: The proposed change would revise 
the Vermont Yankee Technical Specifications Section 6.0, 
``Administrative Controls,'' to add and revise reference to NRC-
approved methodologies which will be used to generate the cycle-
specific thermal operating limits in the Vermont Yankee Core Operating 
Limits Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change will not involve any significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The change updates the Technical Specifications to include an 
NRC approved method reference to allow calculation of thermal limits 
with a revised method. It does not affect plant operation and will 
not weaken or degrade the facility.
    2. The proposed change will not create the possibility of a new 
or different kind of accident since the change is administrative. No 
physical alterations of the plant, setpoint changes, or operating 
conditions are proposed.
    3. The proposed change will not involve a significant reduction 
in a margin of safety. The change involves an update to the 
Administrative Controls in Section 6.0 of the Technical 
Specifications by adding a reference to NRC approved methods. This 
administrative change does not alter plant safety margins.

    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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, DC 20037-1128.
    NRC Project Director: Cecil O. Thomas, Director.

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

    Date of amendment request: February 4, 1998.
    Description of amendment request: The amendment would revise 
Technical Specification 3.2.4, quadrant power tilt ratio (QPTR), and 
associated Bases, to clarify the required actions for the limiting 
condition for operation (LCO) and other changes consistent with the 
technical specification conversion application submitted by letter 
dated May 15, 1997.
    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.

[[Page 14490]]

1. Requirements for Determining QPTR

    The Action to calculate QPTR once per hour until THERMAL POWER 
was reduced to less than 50% RATED THERMAL POWER (RTP) when QPTR 
exceeds the LCO requirements would be deleted and replaced by a new 
requirement to determine QPTR at least once per 12 hours.
    The proposed change involves only the compensatory measures to 
be taken should the QPTR be outside its limit. The frequency with 
which QPTR is calculated is not assumed in the initiating events for 
any accident previously evaluated. In addition, the change does not 
involve any new operating activities or hardware change. Therefore, 
the proposed change would not significantly increase the probability 
of an accident previously evaluated.
    Once THERMAL POWER has been reduced appropriately in proportion 
to the amount that QPTR exceeds 1.00, any additional change would be 
sufficiently slow that a 12-hour interval for recalculating QPTR 
will provide an adequate level of protection. Therefore, the 
proposed change will not significantly increase the consequences of 
any accident previously evaluated.

2. Completion Time for Resetting the Power Range Neutron Flux-High Trip 
Setpoints

    The proposed change to allow 72 hours for resetting the Power 
Range Neutron Flux-High trip setpoints involves only the 
compensatory measures to be taken should the QPTR be outside its 
limit. These compensatory measures are not assumed in the initiating 
events for any accident previously evaluated. The proposed actions 
recognize that the required reduction in power (3% for each 1% of 
indicated QPTR in excess of 1.00) provide adequate margin for fuel 
design limits so that consequences of assumed accidents would not be 
significantly affected. Therefore, the proposed change will not 
adversely affect the probability or consequences of any accident 
previously evaluated. Further, by permitting more time to perform 
resetting the trip setpoints, the chances of a transient may be 
reduced.

3. Delete(tion) of the Actions (a.3., a.4.) for verifying QPTR to be 
restored within 24 hours and for identifying and correcting the cause 
of the out-of-limit condition prior to increasing THERMAL POWER

    The proposed changes would delete current Actions a.3. and a.4. 
and add new Actions for QPTR out of limit including requirements for 
measuring FQ(Z) and F N delta H prior to and 
following a return to power and performing safety analyses to verify 
safety requirements are met prior to increasing power above the 
limits of Action a.1. The proposed changes involve only the 
compensatory measures to be taken should the QPTR be outside its 
limit. These compensatory measures are not assumed in the initiating 
events for any accident previously evaluated. Therefore, the 
proposed change will not affect the probability or consequences of 
any accident previously evaluated.

4. Deletion of the Actions for QPTR in excess of 1.09

    The proposed change would delete the required Actions for QPTR 
in excess of 1.09 and Actions for QPTR in excess of 1.02 are 
followed for all instances where QPTR exceeds 1.02. The proposed 
change involves only the compensatory measures to be taken should 
the QPTR be outside its limit. These compensatory measures are not 
assumed in the initiating events for any accident previously 
evaluated. The proposed actions recognize that the required 
reduction in power (3% for each 1% of indicated QPTR in excess of 
1.00) provide adequate margin for fuel design limits so that 
consequences of assumed accidents would not be significantly 
affected. Therefore, the proposed change will not affect the 
probability or consequences of any accident previously analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

1. Requirements for Determining QPTR

    The proposed change for calculating QPTR once every 12 hours 
does not involve a physical alteration to the plant or change the 
method by which any safety-related system performs its function. The 
manner in which the plant would be operated would not be altered. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any previously evaluated.

2. Completion Time for Resetting the Power Range Neutron Flux-High Trip 
Setpoints

    The proposed change to allow 72 hours for resetting the Power 
Range Neutron Flux-High trip setpoints does not involve a permanent 
physical alteration to the plant; no new or different kinds of 
equipment will be installed. The change would not alter the manner 
in which the plant would be operated only the timing of actions that 
provide potential mitigation of accidents. Thus, the change would 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.

3. Delete the Actions (a.3., a.4.) for verifying QPTR to be restored 
within 24 hours and for identifying and correcting the cause of the 
out-of-limit condition prior to increasing THERMAL POWER

    The proposed changes would delete current Actions a.3, and a.4. 
and add new Actions for QPTR out-of-limit including requirements for 
measuring FQ(Z) and F N delta H prior to and 
following a return to power and performing safety analyses to verify 
safety requirements are met prior to increasing power above the 
limits of Action a.1. The proposed changes do not involve a physical 
alteration to the plant; no new or different kinds of equipment 
would be installed. The changes would not alter the manner in which 
the plant would be operated only the timing of actions that provide 
potential mitigation of accidents. Thus, the changes would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.

4. Deletion of the Actions for QPTR in excess of 1.09

    The proposed change would delete the required Actions for QPTR 
in excess of 1.09 and Actions for QPTR in excess of 1.02 are 
followed for all instances where QPTR exceeds 1.02. The proposed 
change does not involve a physical alteration to the plant or 
changes in the way in which the plant is operated. The proposed 
change involves only the compensatory measures to be taken should 
QPTR be outside its limit. The assumptions of the accident analyses 
are unaffected by the proposed change. No new permutations or event 
initiators are introduced by the proposed alternate methods of 
dealing with QPTRs in excess of 1.09. Therefore, there is no 
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.

1. Requirements for Determining QPTR

    The proposed change for calculating QPTR once every 12 hours 
does not change any accident analysis assumptions, initial 
conditions or results. The proposed change will continue to ensure 
that the plant is maintained in a safe condition while QPTR is in 
excess of its limit. Additionally, calculating QPTR once per 12 
hours as opposed to every hour while QPTR is in excess of its limit 
would avoid the diversion of personnel resources from corrective 
actions with regard to meeting the LCO. Therefore, the proposed 
change will not involve a significant reduction in any margin of 
safety.

2. Completion Time for Resetting the Power Range Neutron Flux-High Trip 
Setpoints

    The proposed change to allow 72 hours for resetting the Power 
Range Neutron Flux-High trip setpoints will continue to ensure that 
the plant is maintained in a safe condition within the envelope of 
the safety analyses while QPTR is in excess of its limit. The 
proposed actions recognize that the required reduction in power (3% 
for each 1% of indicated QPTR in excess of 1.00) provide adequate 
margin for fuel design limits so that consequences of assumed 
accidents would not be significantly affected. Therefore, the 
proposed change will not involve a significant reduction in any 
margin of safety.

3. Delete the Actions (a.3., a.4.) for verifying QPTR to be restored 
within 24 hours and for identifying and correcting the cause of the 
out-of-limit condition prior to increasing THERMAL POWER

    The proposed changes would delete current Actions a.3. and a.4 
and add new Actions for QPTR out-of-limit including requirements for 
measuring FQ(Z) and F N delta H prior to and 
following a return to power and performing safety analyses to verify 
safety requirements are met prior to increasing power above the 
limits of Action a.1. The proposed changes will continue to ensure 
that the plant is maintained in a safe condition within the envelope 
of the safety analysis while QPTR is in excess of its limit. 
Therefore, the proposed changes will not involve a significant 
reduction in any margin of safety.

4. Deletion of the Actions for QPTR in excess of 1.09

    The proposed change would delete the required Actions for QPTR 
in excess of 1.09

[[Page 14491]]

and Action for QPTR in excess of 1.02 are followed for all instances 
where QPTR exceeds 1.02. The proposed change will continue to ensure 
that the plant is maintained in a safe condition within the envelope 
of the safety analyses while QPTR is in excess of its limit. While 
different actions are taken in response to a QPTR in excess of 1.09, 
the proposed change will assure that accident analyses assumptions 
continue to be met. Therefore, the proposed changes will not involve 
a significant reduction in any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: February 4, 1998.
    Description of amendment request: The amendment would revise the 
technical specifications to (1) create separate functional units for 
the analog and digital portions of the engineered safety features 
actuation system (ESFAS) function associated with starting the turbine-
driven auxiliary feedwater pump on a loss of offsite power, and (2) add 
a table notation to clarify that the testing of the time delay relays 
for the 4 kV undervoltage, loss of voltage and grid degraded voltage 
portion of the ESFAS is performed as part of the channel calibration.
    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.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The recognition that different 
OPERABILITY and surveillance requirements apply to analog vs. 
digital circuitry does not impact any previously analyzed accidents. 
The clarification that testing of the time delay relays is performed 
as part of the CHANNEL CALIBRATION does not impact any previously 
analyzed events. The proposed change will not affect any of the 
analysis assumptions for any of the accidents previously evaluated. 
The proposed change does not alter the current method or procedures 
for meeting the surveillance requirements in Table 4.3-2. The 
proposed change will not affect the probability of any event 
initiators nor will the proposed change affect the ability of any 
safety-related equipment to perform its intended function. There 
will be no degradation in the performance of nor an increase in the 
number of challenges imposed on safety-related equipment assumed to 
function during an accident situation. 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.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The separation of analog and digital portions of 
Functional Unit 6.f or the clarification of testing of the time 
delay relays will not impact the normal method of plant operation.
    The OPERABILITY requirements, ACTION Statement, and surveillance 
requirements for the analog portion, new Functional Unit 6.f.1), are 
identical to those of Functional Unit 8.a, while the requirements 
for the digital portion, new Functional Unit 6.f.2), are consistent 
with the current technical specifications, other than the new ACTION 
Statement 30 provisions that defer to the TDAFW pump Specification 
3.7.1.2 requirements and the performance of a TADOT during 
appropriate plant conditions. These changes do not change any ESFAS 
design standard and are appropriate for digital functions such as 
this.
    Testing of the time delay relays has been performed as part of 
the 18 month CHANNEL CALIBRATION. The tolerancesfor the time delay 
relays are sufficient to account for relay drift encountered during 
the 18 month surveillance testing. The calculated tolerances for the 
time delay setpoints have been evaluated to insure that safety-
related systems, subsystems and components would not be adversely 
affect[ed] by the drift within the permissible tolerance band.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this change. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the acceptance criteria for 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection functions. There will be no impact 
on any margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

Previously Published Notices of Consideration of Issuance of 
Amendments toFacility 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.

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

    Date of amendment request: January 26, 1998.
    Brief description of amendment request: The proposed amendment 
would change the SSES Technical Specifications facility staff 
requirements to allow an individual who does not hold a current senior 
reactor operator (SRO) license to hold the position of Manager-Nuclear 
Operations (MNO) and require an individual serving in the capacity of 
the Operations Supervisor-Nuclear to hold a current SRO license

[[Page 14492]]

and report directly to the MNO and be responsible for directing the 
licensed activities of licensed operators.
    Date of publication of individual notice in Federal Register: 
February 24, 1998 (63 FR 9270).
    Expiration date of individual notice: March 26, 1998.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

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

    Date of amendment request: February 25, 1998, TXX-98050.
    Description of amendment request: The proposed amendment would be a 
temporary change to the Technical Specifications to remove the 
requirement to demonstrate the load shedding feature of MCC XEB4-3 as 
part of Surveillance Requirements (SRs) 4.8.1.1.2f.4)a) and 
4.8.1.1.2f.6)a) until the plant startup subsequent to the next 
refueling outage or until an outage of greater than 24 hours in 
duration for each respective unit. This temporary change is requested 
as a result of the failure to confirm the load shedding feature of MCC 
XEB4-3 during the last performance of these SRs for the Unit 1 and Unit 
2 train B diesel generators (DGs).
    Date of individual notice in the Federal Register: March 9, 1998, 
(63 FR 11458).
    Expiration date of individual notice: April 8, 1998.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR 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 amendments: March 18, 1997, as supplemented 
by letters dated July 28, 1997, and September 9, 1997.
    Brief description of amendments: The amendments revise the 
operating licenses to reflect approval of Amendment 42 to the Palo 
Verde Nuclear Generating Station Physical Security Plan. The amendments 
revise the methods used to search materials, packages, and personnel 
prior to their entry into the protected area, as described in the 
security plan.
    Date of issuance: March 4, 1998.
    Effective date: March 4, 1998.
    Amendment No.: Unit 1-115; Unit 2-108; Unit 3-87.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the operating licenses.
    Date of initial notice in Federal Register: October 8, 1997 (62 FR 
52580).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 4, 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, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: December 17, 1997, as 
supplemented by letters dated February 6, 1998 and March 12, 1998.
    Brief description of amendment: The proposed change would revise 
Technical Specifications Section 5.6.5, ``Core Operating Limits 
Report.'' The revisions add reference to an additional approved 
methodology for correlating departure from nucleate boiling (DNB) 
ratios. The added methodology is the Siemens Power Corporation Topical 
Report, EMF-92-153(P)(A), ``HTP: Departure from Nucleate Boiling 
Correlation for High Thermal Performance Fuel.''
    Date of issuance: March 16, 1998.
    Effective date: March 16, 1998.
    Amendment No. 178.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4309). The February 6 and March 12, 1998 submittals provided clarifying 
information that did not affect the initial determination of no 
significant hazards considerations. The Commission's related evaluation 
of the amendment is contained in a Safety Evaluation dated March 16, 
1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

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: October 27, 1997.
    Brief description of amendments: The amendments would change the 
Dresden and Quad Cities Technical Specifications (TS) to clarify the 
applicability, action and surveillance requirements for the Standby 
Liquid Control System (SLCS). The changes would make the current TS 
requirements for the SLCS consistent with the Improved Standard 
Technical Specifications (ISTS) contained in NUREG-1433, ``Standard 
Technical Specifications General Electric Plants, BWR/4.''
    Date of issuance: March 6, 1998.

[[Page 14493]]

    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 167, 162, and 180, 178.
    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: January 14, 1998 (63 FR 
2277).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 6, 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.

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

    Date of application for amendment: October 2, 1996, as supplemented 
July 31, 1997.
    Brief description of amendment: The amendment revises Figures 
3.1.A-1, 3.1.A-2 and 3.1.A-3, Section 3.1.B and its Bases, Figures 
3.1.B-1 and 3.1.B-2, and the Bases of Section 4.3 and Figure 4.3-1 of 
the Technical Specifications to incorporate the revised Indian Point 
Unit 2 Heatup and Cooldown Limit Curves for Normal Operation.
    Date of issuance: February 27, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 195.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1996 (61 
FR 58901).
    The July 31, 1997, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: December 17, 1997Brief 
description of amendments: The amendments revise Section 6.9.1.9 of the 
Technical Specifications to reference updated or recently approved 
topical reports, which contain methodologies used to calculate cycle-
specific limits contained in the Core Operating Limits Report.
    Date of issuance: March 2, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-163; Unit 2-155.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4310).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of application for amendment: July 21, 1997, as supplemented 
February 18, 1998.
    Brief description of amendment: Technical Specification Change 
Request concerning Emergency Feedwater Surveillance Testing. This 
request is to make several changes to the ANO-2 Technical 
Specifications including extension of the emergency feedwater (EFW) 
pump surveillance testing frequency, a reduction in the minimum steam 
generator pressure required to perform the surveillance testing on the 
turbine-driven EFW pump, and a modification to the EFW pump testing 
requirements.
    Date of issuance: March 12, 1998.
    Effective date: March 12, 1998.
    Amendment No.: 188.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43367).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of application for amendment: September 23, 1997, as 
supplemented by letters dated February 27 and March 4, 1998.
    Brief description of amendment: The amendment changes the Reactor 
Protective System (RPS) and Engineering Safety Actuation System (ESFAS) 
trip set point and allowable values for steam generator low pressure. 
The amendment also relocates the RPS and ESFAS response time tables 
from the Technical Specifications to the Safety Analysis Report as 
described in NRC Generic Letter 93-08, ``Relocation of Technical 
Specification Tables of Instrument Response Time Limits,'' dated 
December 29, 1993.
    Date of issuance: March 12, 1998.
    Effective date: March 12, 1998.
    Amendment No.: 189.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: January 28, 1998, (63 
FR 4311).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of application for amendment: September 23, 1997, as 
supplemented by letters dated February 27 and March 4, 1998.
    Brief description of amendment: The amendment reduces the minimum 
required reactor coolant system flow rate in TS 3.2.5 until the ANO-2 
steam generators are replaced. The reduced reactor coolant system flow 
requirement will account for plugging of up to approximately 30 percent 
of the tubes in the existing steam generators at ANO-2.
    Date of issuance: March 12, 1998.
    Effective date: March 12, 1998.
    Amendment No.: 190.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications/license.
    Date of initial notice in Federal Register: January 28, 1998, (63 
FR 4312).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 1998.

[[Page 14494]]

    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of application for amendment: December 5, 1997, as 
supplemented December 11, 1997, January 9, February 12 and 19, 1998.
    Brief description of amendment: To revise the Final Safety Analysis 
Report (FSAR) and the Improved Technical Specification Bases to reflect 
the modified reactor building fan recirculation system fan cooler 
starting logic.
    Date of issuance: March 9, 1998.
    Effective date: March 9, 1998.
    Amendment No.: 165.
    Facility Operating License No. DPR-31: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 15, 1998 (63 FR 
2423). The supplemental letters dated December 11, 1997, January 9, 
February 12 and 19, 1998, did not change the original no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida.
    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.

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

    Date of application for amendments: May 14, 1997, as supplemented 
by letter dated October 9, 1997 (published in Federal Register as May 
15, 1997).
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant (DCPP) Unit Nos. 1 and 2 to revise the surveillance frequencies 
from at least once every 18 months to at least once per refueling 
interval (nominally 24 months) including (1) reactor coolant system 
total flow rate, (2) instrumentation for radiation monitoring, (3) 
instrumentation and controls for remote shutdown, (4) instrumentation 
for accident monitoring, and (5) several miscellaneous TS.
    Date of issuance: February 27, 1998.
    Effective date: February 27, 1998, to be implemented within 90 days 
of the date of issuance.
    Amendment Nos.: Unit 1-123; Unit 2-121.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40855).
    The October 9, 1997, supplemental letter provided additional 
clarifying information and did not change the staff's initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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: January 2, 1997, as supplemented 
November 13, 1997.
    Brief description of amendment: The amendment changes the Technical 
Specifications by extending the surveillance interval for the 
functional testing of certain Inservice Inspection American Society of 
Mechanical Engineers Code Class 1, 2, and 3 pumps and valves from once 
a month to once a quarter.
    Date of issuance: March 2, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 178.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14468).
    The November 13, 1997, submittal contained clarifying information 
that did not change the staff's proposed finding of no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.

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

    Date of application for amendment: December 14, 1995, as 
supplemented September 26, 1997.
    Brief description of amendment: The amendment changes the James A. 
FitzPatrick Technical Specifications (TSs) to incorporate the inservice 
testing requirements of Section XI of the American Society of 
Mechanical Engineers Boiler and Pressure Vessel Code. The amendment 
supplements Amendment No. 241, dated December 2, 1997, by issuing seven 
TS pages inadvertently omitted from Amendment No. 241.
    Date of issuance: February 27, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 242.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1635).
    The September 26, 1997, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendments: December 15, 1997.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) to adopt Option B, of 10 CFR Part 50, 
Appendix J, ``Primary Reactor Containment Leakage Testing for Water-
Cooled Power Reactors,'' to implement a performance-based approach for 
Type B and C testing. Additionally, the wording in the TSs would be 
modified for the previous adoption of Option B on Type A testing and a 
section added on the primary

[[Page 14495]]

containment leakage rate testing program.
    Date of issuance: February 27, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos: 207 and 188.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 FR 
2281).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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 17, 1997.
    Brief description of amendments: The amendments extended the 
surveillance interval of the containment spray nozzle air flow test to 
ten years from five years.
    Date of issuance: March 11, 1998.
    Effective date: March 11, 1998.
    Amendment Nos.: Unit 1--Amendment No. 94; Unit 2--Amendment No. 81.
    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 Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 11, 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.

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

    Date of application for amendment: December 23, 1997.
    Brief description of amendment: This amendment revised Technical 
Specification 3.8.1, ``A.C. Sources--Operating,'' consistent with the 
recommendations in NRC Generic Letter 94-01, ``Removal of Accelerated 
Testing and Special Reporting Requirements for Emergency Diesel 
Generators.''
    Date of issuance: March 12, 1998.
    Effective date: March 12, 1998.
    Amendment No.: 92.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4326).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.

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

    Date of application for amendment: July 11, 1997, as supplemented 
November 21, December 22, 1997, and February 6, 1998.
    Brief description of amendment: The amendment revised Technical 
Specifications 3.7/4.7 and their associated Bases to incorporate Option 
B of Appendix J to 10 CFR 50, and editorial changes to TS Table 4.7.2
    Date of Issuance: February 26, 1998.
    Effective date: February 26, 1998, with full implementation within 
30 days.
    Amendment No.: 152.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: (62 FR 45465). The 
November 21, December 22, 1997, and February 6, 1998, letters did not 
change the initial proposed no significant hazards determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated February 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of application for amendment: November 20, 1997.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.10 and its associated Bases to eliminate the use 
of battery charger AB for meeting the requirement of the TS.
    Date of issuance: March 5, 1998.
    Effective Date: This license amendment is effective as of its date 
of issuance, to be implemented within 30 days.
    Amendment No.: 153
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68319).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of application for amendment: August 22, 1997, as supplemented 
by letter dated September 18 and October 31, 1997.
    Brief description of amendment: The amendment revises the Technical 
Specifications to address the new low pressure CO2 suppression system 
for the East and West Switchgear Rooms and more clearly describes the 
separation of the two rooms.
    Date of Issuance: March 6, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 154.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 8, 1997 (62 FR 
52590). Information provided by letter dated October 31, 1997, did not 
affect the original no significant hazards consideration.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 6, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following

[[Page 14496]]

amendments. The Commission has determined for each of these amendments 
that the application for the amendment complies with the standards and 
requirements of the Atomic Energy Act of 1954, as amended (the Act), 
and the Commission's rules and regulations. The Commission has made 
appropriate findings as required by the Act and the Commission's rules 
and regulations in 10 CFR Chapter I, which are set forth in the license 
amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By April 24, 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

[[Page 14497]]

requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendments: February 5, 1998, as 
supplemented February 12, March 3 and 5, 1998.
    Brief description of amendments: The amendments revised the 
surveillance requirements in Technical Specification (TS) 4.6.1.2 
(Requirement a). The change to the referenced TS adds a footnote 
stating that the requirement for Type A testing will not apply to 
certain instrument line penetrations.
    Date of issuance: March 10, 1998.
    Effective date: Both units, as of the date of issuance.
    Amendment Nos.: 173 and 146.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No. On February 5, 1998, the staff issued a Notice of 
Enforcement Discretion, which was immediately effective and remained in 
effect until this amendment was issued.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances, consultation with the State of Pennsylvania, 
and final no significant hazards consideration determination are 
contained in a Safety Evaluation dated March 10, 1998.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

    Dated at Rockville, Maryland, this 18th day of March 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-7652 Filed 3-24-98; 8:45 am]
BILLING CODE 7590-01-P