[Federal Register Volume 61, Number 60 (Wednesday, March 27, 1996)]
[Notices]
[Pages 13521-13540]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10327]



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


NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 4, 1996, through March 15, 1996. The 
last biweekly notice was published on March 13, 1996.

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 Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, 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 26, 1996, 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

[[Page 13522]]
Board will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: February 1, 1996
    Description of amendments request: The proposed amendment would (1) 
revise Technical Specifications (TS) Sections 3/4.1.1.1, 6.9.1.9, and 
6.9.1.10 to relocate the shutdown margin (reactor trip breakers open) 
to the Core Operating Limits Report (COLR); (2) revise TS 3/4.3.2 
(Tables 3.3-3 and 3.3-4), to specify an additional restriction for the 
allowed low pressurizer pressure trip setpoint when reducing reactor 
coolant system (RCS) pressure in Mode 3; (3) revise TS Section 2.2.1 
(Table 2.2-1) to make it consistent with the footnote in TS Tables 3.3-
3 and 3.3-4; and (4) revise TS Sections 3/4.5.2 and 3/4.5.3 to specify 
an additional restriction to require that two emergency core cooling 
system (ECCS) subsystems be operable in Mode 3 whenever the RCS cold 
leg temperature is equal to or above 485 degrees F. In addition, the 
Table of Contents and the Bases would be revised to be consistent with 
these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the Updated Final Safety Analysis Report (UFSAR). The proposed 
changes to TS Tables 2.2-1, 3.3-3, and 3.3-4 to add additional 
restrictions to the pressurizer pressure - low trip setpoint 
requirements are more conservative than the current Technical 
Specifications and will reflect the updated Mode 3 steam line break 
safety analyses assumptions. The proposed changes to TS sections 3/
4.5.2 and 3/4.5.3 to add additional restrictions to the requirement 
to have two ECCS Subsystems operable are also more conservative than 
the current Technical Specifications and will reflect the updated 
Mode 3 steam line break safety analyses assumptions. Since these 
changes are more restrictive, they would not contribute to the 
initiation of any accident, nor would they increase the consequences 
of an accident, but

[[Page 13523]]
they would enhance the plant response to a steam line break in Mode 
3 to reduce consequences. The proposed changes to relocate the 
shutdown margin - reactor trip breakers open to the COLR will have 
no effect on the initiation or consequences of an accident. The 
shutdown margin-reactor trip breakers open, which would be 
determined using NRC approved analytical methods, as required by the 
proposed changes, would ensure that the probability and consequences 
of an accident would not increase. The changes to the titles of TS 
3/4.5.2 and 3/4.5.3, and to the Table of Contents, are editorial and 
have no effect on the operation of the plant or on any structures, 
systems or components.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes do not create the possibility of an 
accident of a new or different kind. The proposed changes to TS 
Tables 2.2-1, 3.3-3, and 3.3-4, and TS section 3/4.5.2 and 3/4.5.3, 
to add additional restrictions to the pressurizer pressure - low 
trip setpoint requirement and add additional restrictions to the 
requirement to have two ECCS Subsystems operable are more 
conservative than the current Technical Specifications and will 
reflect the updated Mode 3 steam line break safety analyses 
assumptions. Since these changes are more restrictive, and therefore 
bounded by the current TS, they would not contribute to the 
initiation of any kind of new or different accident. The proposed 
changes to relocate the shutdown margin -reactor trip breakers open 
to the COLR will have no effect on the possibility of a new or 
different kind of accident. The shutdown margin-reactor trip 
breakers open, which would be determined using NRC approved 
analytical methods as required by the proposed changes, would ensure 
that there would be no possibility of a new or different kind of 
accident from any accident previously evaluated. The changes to the 
titles of TS 3/4.5.2 and 3/4.5.3, and to the Table of Contents, are 
editorial and have no effect on the operation of the plant or on any 
structures, systems or components.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed TS changes do not involve a reduction in any margin 
of safety. The proposed changes to TS Tables 2.2-1, 3.3-3, and 3.3-
4, and TS section 3/4.5.2 and 3/4.5.3, to add additional 
restrictions to the pressurizer pressure - low trip setpoint 
requirement and add additional restrictions to the requirement to 
have two ECCS Subsystems operable are more conservative than the 
current Technical Specifications and will reflect the updated Mode 3 
steam line break safety analyses assumptions. Since these changes 
are more restrictive, they do not involve a reduction in any margin 
of safety as currently established by the existing TS. The proposed 
changes to relocate the shutdown margin - reactor trip breakers open 
to the COLR will have no effect on any margin of safety. The 
shutdown margin - reactor trip breakers open would be determined 
using NRC approved analytical methods as required by the proposed 
changes, thus ensuring that there would be no reduction in any 
margin of safety. The changes to the titles of TS 3/4.5.2 and 3/
4.5.3, and to the Table of Contents, are editorial and have no 
effect on the operation of the plant or on any structures, systems 
or components.
    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

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

    Date of amendment request: February 15, 1996
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.7 to add operability requirements 
for the Keowee Hydro units during periods of commercial power 
generation. These requirements are based on lake level and power level 
of the Keowee Hydro units. Also, two surveillance requirements would be 
added to TS 4.6 to (1) address periodic testing of the circuitry that 
was added by the modification approved in NRC's SER dated August 15, 
1995, and (2) add a load rejection surveillance to ensure that the 
response of the Keowee Hydro units is bounded by the design criteria 
used to develop the Keowee operating restrictions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) [Does not] involve a significant increase in the probability 
or consequences of an accident previously evaluated:
    Each accident analysis addressed within the Oconee Final Safety 
Analysis Report (FSAR) has been examined with respect to the change 
proposed within this amendment request. The probability of any 
Design Basis Accident (DBA) is not significantly increased by this 
change. In addition, the consequences of the accidents are within 
the bounds of the FSAR analyses.
    The design basis of the auxiliary electrical systems is to 
supply the required engineered safeguards (ES) loads of one unit and 
the safe shutdown loads of the other two units. The systems are 
arranged so that no single failure will jeopardize plant safety. The 
addition of the operability requirement and surveillances for the 
Keowee Hydro units will ensure that the electrical systems can meet 
their design basis.
    (2) [Does not] create the possibility of a new or different kind 
of accident from any kind of accident previously evaluated:
    Addition of the operability requirement and surveillances will 
not create a new or different kind of accident. The addition of the 
circuitry which is covered by the operability requirement and 
surveillances has been reviewed and approved by the NRC. Therefore, 
operation of ONS [Oconee Nuclear Station] in accordance with this 
Technical Specification amendment will not create any failure modes 
not bounded by previously evaluated accidents. Consequently, this 
change will not create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated.
    (3) [Does not] involve a significant reduction in a margin of 
safety:
    The design basis of auxiliary electrical systems is to supply 
the required ES loads of one Unit and safe shutdown loads of the 
other two units. The ability of the Keowee Hydro units to provide 
emergency power following an accident during a period of Keowee 
Hydro commercial power generation was reviewed and approved by the 
NRC in [an] SER dated August 15, 1995. Therefore, there will be no 
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 location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: February 20, 1996
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) 3.1.5, 3.1.10, and 4.1. The TS 
changes would: (1) reduce the frequency for the concentrated boric acid 
storage tank boron concentration surveillance, (2) delete the chemical 
and radiochemical surveillance requirements for the reactor

[[Page 13524]]
coolant for Sr<approx-eq><SUP>189 and Sr<approx-eq><SUP>190, gross beta 
activity, gross alpha activity, dissolved gas concentration in the 
reactor coolant, and gross beta activity in the steam generator 
feedwater, and (3) relocate the surveillance requirements for tritium, 
chloride, fluoride and oxygen to the Selected Licensee Commitments 
(SLC) Manual. The proposed changes would also delete some temperature 
and pressure requirements on control rod 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. The licensee has determined that operation of the 
facility in accordance with the proposed amendments would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    Each accident analysis addressed within the Oconee Final Safety 
Analysis Report (FSAR) has been examined with respect to the 
proposed amendment request. The probability of any Design Basis 
Accident (DBA) is not significantly increased by the proposed 
amendment due to the fact that the identified cause in the FSAR 
accidents is not impacted. In addition, the consequences of the 
accidents are within the bounds of the FSAR analyses since the 
proposed amendment does not change the accident analysis methods or 
assumptions described in the FSAR.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    The proposed amendment revises and eliminates several of the RCS 
[Reactor Coolant System] chemistry Technical Specification 
surveillance requirements. The changes in the surveillance 
requirements do not alter the plant safety features or the method of 
operation at ONS [Oconee Nuclear Station]. Therefore, operation of 
ONS in accordance with the proposed Technical Specification will not 
create any failure modes not bounded by previously evaluated 
accidents.
    (3) Involve a significant reduction in a margin of safety.
    The proposed amendment does not impact the mitigation of any of 
the accidents analyzed in the FSAR. Therefore, there is not a 
significant reduction in the margin of safety associated with the 
proposed amendment.
    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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: February 22, 1996
    Description of amendment request: The licensee has proposed to 
increase the safety function lift setpoint tolerances for the safety 
and relief valves that are listed in Surveillance Requirement 3.4.4.1 
(Page 3.4-10) of the Technical Specifications TSs) for the Grand Gulf 
Nuclear Station, Unit 1. The tolerances would be increased from the 
current plus/minus 1 percent of the safety function (i.e., safety 
relief valve) lift setpoint to plus/minus 3 percent.
    The frequency of verifying these setpoints would not be changed by 
this amendment request. Also, the other surveillance requirements in 
the TSs on these valves and the number of these valves required to be 
operable are not being changed by this amendment request.
    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 (NSHC) in Attachment 2 to its application of February 22, 
1996.
    In its application, the licensee stated that it has used the NRC 
staff's safety evaluation report (SER), NEDC 31753-P-A, issued in the 
NRC letter of March 8, 1993, which evaluated General Electric (GE) 
topical report NEDC-31753P, ``BWROG [BWR Owners' Group] In-Service 
Pressure Relief Technical Specification Revision Licensing Topical 
Report,'' dated February 1990.
    The licensee's NSHC analysis is presented below:
    Entergy Operations, Inc. is proposing that the Operating License 
for Grand Gulf Nuclear Station (GGNS) be amended to increase the 
tolerance of the safety function lift setpoints [from plus/minus 1%] 
to plus/minus 3%. The GGNS Inservice Testing (IST) program controls 
the frequency of safety relief valve (S/RV) testing as required by 
the GGNS Operating License; therefore, this proposal will also 
incorporate changes [concerning the setpoint tolerances] to 
applicable IST procedures. GGNS will incorporate the recommendations 
of the NEDC-31753-P-A [NRC staff's] SER, by resetting the safety 
function [S/RV] lift setpoints for all tested valves to within plus/
minus 1% of the design lift setpoint and increasing the test sample 
size by two valves for each valve found outside the plus/minus 3% 
safety function lift setpoint. S/RV test sample population will be 
determined based upon the currently licensed ASME [American Society 
of Mechanical Engineers] Boiler and Pressure Vessel Code.
    The commission has provided standards for determining whether a 
no significant hazards consideration exists as stated in 
10CFR50.92(c). A proposed amendment to an operating license involves 
no significant hazards if the 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.
    Entergy Operations, Inc. has evaluated the no significant 
hazards considerations in its request for a license amendment. In 
accordance with 10CFR50.91(a), Entergy Operations, Inc. is providing 
the following analysis of the proposed amendment against the three 
standards in 10CFR50.92(c):
    a. No significant increase in the probability or consequences of 
an accident previously evaluated results from this change.
    The GGNS safety design bases for the S/RVs are:
    ) Prevent overpressurization of the nuclear system that could 
lead to failure of the reactor coolant pressure boundary,
    ) Provide automatic depressurization for small breaks in the 
nuclear system,
    ) Permit verification of operability,
    ) Withstand adverse combinations of loadings and forces during 
abnormal, accident, or special event conditions.
    The most limiting vessel overpressurization event is a closure 
of all main steam isolation valves with a high flux scram. This 
event was analyzed for GGNS using the minimum number of S/RVs 
required by the GGNS Operating License. The safety function lift 
setpoint tolerance used in the analysis bounds the proposed plus/
minus 3% setpoint tolerance. The analysis indicates that the S/RVs 
are capable of maintaining adequate margin below the Operating 
License Reactor Coolant System Pressure of 1325 psig.
    Anticipated operational transients can also challenge the 
operation of the S/RVs, for instance, Generator Load Reject without 
Bypass. Analyses have been performed on the limiting events that 
bound other pressure transient events using safety function limit 
setpoint tolerances that bound the proposed plus/minus 3% tolerance 
request. Fuel operating limits are based on the results of these 
analyses; therefore, adequate fuel thermal margin is maintained.
    Plant transients and events that require the use of automatic 
depressurization and the low-low set feature utilize the relief mode 
of S/RV operation. This proposed change does not affect the relief 
mode of S/RV operation.
    The verification of valve operability will still be performed in 
accordance with the GGNS Inservice Testing Program, and S/RV safety 
mode operability will be verified prior to reinstallation. Analysis 
of the loads placed on each S/RV sub-system (discharge piping, 
spargers and associated components) verifies that adequate margin 
exists to ensure that the

[[Page 13525]]
overpressurization system can perform its designed function.
    The negative tolerance of the safety function lift setpoint 
remains above the highest setpoint of the S/RV relief mode, and 
therefore normal vessel pressure. This margin provides reasonable 
assurance that inadvertent opening of an S/RV will not occur during 
power operations.
    GGNS will replace each S/RV removed for IST program testing with 
an S/RV that has been reset to within plus/minus 1% of the designed 
safety function lift setpoint. During each refueling outage, at 
least six of the installed S/RVs will be tested for safety lift 
setpoint in accordance with the current IST program plant 
procedures. This sample population is in agreement with the current 
ASME Boiler and Pressure Vessel Code requirements for the GGNS IST 
program, and is more restrictive than the ANSI/ASME OM-1-1981 
requirement upon which the setpoint tolerance was based. For S/RV 
setpoint testing ([the] as-found [setpoint]), additional valves will 
be tested if the as-found setpoint is outside plus/minus 3% of its 
designed safety function lift setpoint. Sample expansion will be 
consistent with the NEDC 31753-P-A SER requirement of two additional 
valves per valve failure.
    The GGNS UFSAR currently requires at least fifty percent of the 
installed valves to be removed and tested during each refueling 
outage. GGNS FSAR Questions & Responses <greek-i> 211.49 discusses 
the bases for this requirement. The concern regarded the performance 
of S/RVs installed in operating plants at the time of GGNS 
construction and licensing, and that new plants should have 
significantly better performing S/RVs. The fifty percent requirement 
provides a very conservative margin of testing to demonstrate that 
no common cause of S/RV failure occurs within any one operating 
cycle. The minimum testing of six valves proposed for each outage, 
with additional testing for each failure from the initial test 
population, provides reasonable assurance that no common cause 
failure is occurring without early detection. [The minimum testing 
of six valves is in agreement with the current ASME Code 
requirements and is consistent with the current industry practices 
that was accepted in the NRC staff's safety evaluation report, NEDC 
31753-P-A.]
    One of the major factors in the requirement of additional 
testing population beyond ASME Boiler and Pressure Vessel Code is 
many of the older plants were experiencing failures with multiple 
stage pilot operated S/RVs. The safety function of this type of S/RV 
requires operation of a pilot valve that is susceptible to excessive 
leakage and corrosive bonding to cylinder walls; thereby preventing 
proper safety function operation. The GGNS Dikkers S/RVs are direct 
acting, and do not require the operation of a pilot valve for the 
safety function. The Dikkers S/RV Instruction Manual recommends ``to 
replace part of the installed valves each maintenance stop 
(refueling outage)'', and does not prescribe any particular [number 
of valves to be tested].
    Therefore, no significant increase in the probability or 
consequences of an accident previously evaluated results from this 
proposed change.
    b. This change would not create the possibility of a new or 
different kind of accident from any previously analyzed.
    The plant specific analyses verify that each S/RV will still 
perform the intended function of preventing overpressurization of 
the nuclear system. The vessel will have adequate margin below the 
Operating License Reactor Coolant System Pressure of 1325 psig, and 
plant system response will not deviate from the expected sequence of 
events. Each system, structure, and component that communicates with 
the reactor vessel has been verified to be within its design and 
operational margin, and no unanticipated plant transients will occur 
as a result of the safety lift function setpoint tolerance change.
    The negative tolerance of the safety function lift setpoint 
remains above the highest setpoint of the S/RV relief mode, and 
therefore normal vessel pressure. This margin provides reasonable 
assurance that inadvertent opening of an S/RV will not occur during 
power operations.
    This proposed change does not add any new systems, structures or 
supports, nor does it introduce new S/RV operating modes.
    Therefore, this change would not create the possibility of a new 
or different kind of accident from any previously analyzed.
    c. This change would not involve a significant reduction in the 
margin of safety.
    The increase in the S/RV safety function lift tolerance has been 
analyzed for bounding limiting events and accident conditions. [The 
safety function lift setpoint tolerance used in the analysis bounds 
the proposed plus/minus 3% setpoint tolerance.] No condition exists 
that reduces the margin of safety on the reactor coolant pressure 
boundary or any system, structure or component that is required to 
operate during vessel overpressurization events. Fuel operating 
limits are based on the results of these analyses; therefore, 
adequate fuel thermal margin is maintained.
    [The negative tolerance of the safety function lift setpoint 
remains above the highest setpoint of the S/RV relief mode, and 
therefore normal vessel pressure. This margin provides reasonable 
assurance that inadvertent opening of an S/RV will not occur during 
power operations.]
    Therefore, this change would not involve a significant reduction 
in the margin of safety.
    Based on the above evaluation, Entergy Operations, Inc. has 
concluded that operation in accordance with the proposed amendment 
involves no significant hazards considerations.
    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: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: February 22, 1996
    Description of amendment request: The amendment proposes to delete 
a specification which requires a thorough inspection of the Emergency 
Diesel Generator (EDG) every 24 months during shutdown. In addition 
this Technical Specification proposes to delete the phrase ``in any 
thirty day period'' from a specification concerning Allowed Outage time 
(AOT).
    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:
    GPU Nuclear has determined that this [technical specification 
change request] TSCR poses no significant hazard as defined by the 
NRC in 10 CFR 50.92.
    1. State the basis for the determination that the proposed 
activity will or will not increase the probability of occurrence of 
the consequences of an accident.
    The proposed activity deletes the requirement to inspect EDGs 
during shut down from the Technical Specifications. It further 
modifies the operability of a single EDG for a limited and defined 
period of time. These changes do not affect the design or 
performance of the EDGs or their ability to perform their design 
function. Analysis using PRA techniques indicates the changes do not 
significantly increase the probability or consequences of an 
accident.
    2. State the basis for the determination that the activity does 
or does not create a possibility of an accident or malfunction of a 
different type than any previously identified in the SAR.
    The EDGs are not the source of any accident described in the 
SAR. These changes do not modify the design or performance of the 
EDGs and do not affect plant functions or actions. Therefore, the 
proposed change does not create the possibility of an accident or 
malfunction of a different type than those previously identified.
    3. State the basis for the determination that the margin of 
safety is not reduced. The proposed changes are designed to improve 
EDG reliability and availability during shutdown periods by 
providing flexibility in the scheduling and performance of 
maintenance. The surveillance intervals are unchanged and 
operability requirements are only modified to an acceptable degree. 
The proposed activity does not alter the basis of

[[Page 13526]]
any technical specification that is related to the establishment or 
maintenance of a nuclear safety margin. Therefore, the margin of 
safety is not significantly reduced by this action.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: February 23, 1996
    Description of amendment request: The proposed change to the 
Technical Specifications would allow the implementation of 10 CFR 50, 
Appendix J, Option B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    GPU Nuclear has determined that this TSCR [technical 
specification change request] involves no significant hazards 
considerations as defined by NRC in 10 CFR 50.92.
    The major changes from the existing Oyster Creek Technical 
Specifications requested in accordance with the Option B 
requirements:
    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 or 
malfunction of equipment important to safety as previously evaluated 
in the Safety Analysis Report.
    The proposed change implements Option B of 10 CFR 50, Appendix J 
on performance based containment leakage testing. The proposed 
change does not involve a change to the plant design or operation. 
Therefore, the proposed change does not affect any of the parameters 
or conditions that contribute to initiation of any of the analyzed 
accidents or malfunctions. The proposed change does request an 
allowable extension of containment testing. Therefore, a 
hypothetical leak could remain undetected for a greater period of 
time. This slight increase in risk has been determined to be 
insignificant as:
    Type A Testing
    NUREG 1493 determined that the effect of containment leakage on 
overall accident risk is small as risk is dominated by accident 
sequences that result in the failure or bypass of the containment. 
Industry wide PCILRTs have demonstrated that only a small fraction 
of the leaks discovered during testing exceeded acceptance criteria, 
and that the leak rate has been only marginally above the acceptable 
limit. Only 3% of all leaks can be detected only by PCILRT, 
therefore, only 3% of the theoretical leaks are affected by the 
extension to the Type A test interval. Experience at Oyster Creek 
agrees with the industry wide data in that the majority of the 
detected leakage from the primary containment is found through Type 
B and C testing.
    NUREG 1493 found that these observations, together with the 
insensitivity of reactor accident risk to the containment leakage 
rate, demonstrates that increasing the Type A leakage test intervals 
would have a minimal impact on public risk.
    Type B and C Testing
    Penetrations are designed to ensure reliability of the 
containment isolation function. Type B penetrations use a double 
passive seal (e.g. o-ring, gasket) and Type C penetrations use a 
double isolation valve design to ensure reliability of the isolation 
function. Because valves perform the isolation function actively, 
they are more likely to fail on demand (e.g. failure to completely 
close on demand). To address this failure mode, Type C valves are 
subjected to increased design constraints and testing to ensure both 
acceptable leak rates and stroke times. The proposed change does not 
alter the installation, operation, operating environment, or testing 
method of these valves. Therefore, the proposed change does not 
introduce any new component failure modes, nor does it affect the 
probability of occurrence of any existing evaluated failure mode.
    The failure of any single penetration barrier (isolation valve 
or passive seal) does not cause penetration failure. Therefore, a 
double failure would have to occur to cause a failure of the 
penetration and affect containment. Additionally, the proposed 
change does not change the acceptance criteria for acceptable 
leakage testing.
    The proposed change does not alter plant design or operation, 
nor does it alter the allowable maximum leakage rate limit. Thus, 
the proposed change does not affect the probability of occurrence 
nor the consequences of any evaluated accident or malfunction of 
equipment important to safety.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of an accident or 
malfunction different from any accident or malfunction previously 
evaluated.
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
initiation of any accidents. This change only involves the reduction 
in Type A, B and C test frequencies, and the Type A test pressure.
    Type A Testing
    The only changes proposed to the Type A testing are to frequency 
and test pressure. As the proposed test pressure is grater than the 
existing test pressure, no new type of accident or malfunction is 
created, and the increase in pressure provides an additional margin 
of safety. The increase in pressure provides an additional margin of 
safety. The increase in surveillance interval cannot introduce any 
new type of accident or malfunction.
    The PCILRT is presently performed at 20 psig. Performance of the 
PCILRT at P<INF>a (35 PSIG) will provide a more direct leak rate for 
analysis.P<INF>a is the design pressure of the torus (the drywell 
design pressure is 44 psig, but the torus is non isolable form the 
drywell. Therefore, P<INF>a will not create the possibility of the 
failure of the torus due to overpressurization. No new accident 
modes can be created by extending the test intervals. No safety 
related functions or components are altered as a result of this 
change. Therefore, no new accident or malfunction different form 
those evaluated in the Safety Analysis Report can result due to the 
increase in test pressure or increase in surveillance interval.
    Type B and C Testing
    The proposed change only deals with the frequency of performing 
Type B and C testing. It does not change what components are tested 
or the method of testing. There is no proposed change to the design 
or operation of the plant. Therefore, no new accident or malfunction 
different form those evaluated in the Safety Analysis Report can 
result due to the increase in test pressure or increase in 
surveillance interval.
    3. Operation of the facility in accordance with the proposed 
amendment would not decrease the margin of safety as defined in the 
bases of the Technical Specifications.
    Type A Testing
    Except for the method of defining the test frequency and 
pressure at which the PCILRT is performed, the methods for 
performing the actual test are not changed. However, the proposed 
change can increase the probability that an increase in leakage 
could go undetected for an extended period of time. NUREG 1493 has 
determined that under several different accident scenarios, the 
increased risk of radioactivity release from containment is 
negligible with the implementation of these proposed changes.
    Type B and C Testing
    The proposed change only affects the frequency of Type B and C 
testing. The methods for performing the actual test are not changed. 
The design or operation of Type B and C components are not changed. 
The proposed change will result in a longer interval between tests 
of good performing Type B and C components.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to containment 
leakage rate. The containment isolation system is designed to limit 
leakage to L<INF>a, which is defined by the Oyster Creek Technical 
Specifications to be 1.0 percent by weight of the containment air at 
35 psig per 24 hours. The limitation on

[[Page 13527]]
containment leakage rate is designed to ensure the total leakage 
volume will not exceed the value assumed in the accident analyses at 
the peak accident pressure (P<INF>a). The margin of safety for the 
offsite dose consequences of postulated accidents directly related 
to the containment leakage rate is maintained by meeting the 1.0 
L<INF>a acceptance criteria. The L<INF>a value is not being modified 
by this proposed Technical Specification change request.
    Therefore, the margin of safety as defined in the bases for the 
Technical Specification will not be reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment requests: February 22, 1996 (AEP:NRC:0659AA)
    Description of amendment requests: The proposed amendments would 
revise the technical specifications to remove the requirement that the 
Operations Superintendent must hold or have held a Senior Operator 
License at Cook Nuclear Plant, or a similar reactor. In addition, a 
mid-level operations manager will only be required to hold a Senior 
Operator License if the Operations Superintendent does not hold one.
    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:
    Per 10 CFR 50.92, this proposed change does not involve a 
significant hazards consideration because the change does not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety.
    Criterion 1
    The amendment request does not involve a significant increase in 
the probability or consequences of [an] accident previously 
evaluated because the proposed change to the Technical Specification 
does not affect the assumptions, parameters, or results of any UFSAR 
[updated final safety analysis report] accident analysis. The 
proposed amendment does not modify any existing equipment. It is 
concluded that the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Criterion 2
    The proposed change does not involve physical changes to the 
plant or changes in plant operating configuration. The proposed 
change updates the requirements for the Operations Superintendent. 
Thus, it is concluded that the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Criterion 3
    The proposed change updates the requirements for Operations 
Superintendent. There is no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook 
Nuclear Plant, Unit No. 2, Berrien County, Michigan

    Date of amendment request: March 12, 1996 (AEP:NRC:1248)
    Description of amendment request: The proposed amendment would 
remove the technical specifications related to shutdown and control rod 
position indication while in modes 3, 4, and 5. The change would make 
the Unit 2 technical specifications consistent with the Unit 1 
technical specifications and the Standard Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Per 10 CFR 50.92, this proposed change does not involve a 
significant hazards consideration because the change does not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety.
    Criterion 1
    The boron concentration in the reactor coolant system will be 
high enough to assure adequate SDM in modes 3, 4, and 5. The 
calculation to obtain the required boron concentration takes into 
account the position of the rods. Shutdown margin is assumed as an 
initial condition in the safety analysis. The safety analysis 
establishes a SDM that ensures specified acceptable fuel design 
limits are not exceeded. As long as the SDM is satisfied, no change 
in the probability or consequences of an accident previously 
evaluated will result from the proposed deletion of the ``position 
indicator - shutdown'' specification. It is noted that this change 
is consistent with the new ISTS approved by the NRC as NUREG-1431, 
Rev. 1.
    Criterion 2
    The ability to insert the control and shutdown rods provided by 
the rod control system is not affected by the OPERABILITY status of 
the ARPI system. As mentioned previously, the reactor coolant system 
boron concentration will be high enough to assure adequate SDM is 
maintained. Therefore, it is concluded that the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Criterion 3
    The margin of safety requirements are not affected by the 
removal of this T/S. The required SDM which is an initial condition 
in the safety analysis, is unaffected since the reactor coolant 
system boron concentration is increased to address the potential 
``all rods out'' configuration. Based on these considerations, it is 
concluded that the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: November 29, 1995
    Description of amendment request: The proposed amendment would

[[Page 13528]]
modify the Technical Specifications to remove the requirement for 
additional pressure relief by a residual heat removal (RHR) spring 
relief valve during low temperature overpressure protection (LTOP) 
conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change to delete Technical Specification 3.4.D.3b 
has been evaluated against the standards of 10 CFR 50.92 and has 
been determined not to involve a significant hazards consideration. 
This proposed change does not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously analyzed. The Power Operative 
Relief Valves (PORVs) remain operable to mitigate any LTOP event. 
Thus, this change does not result in an increase in the probability 
or consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated. Removing the RHR spring relief valve as 
an additional relief requirement does not create the possibility of a 
new or different kind of accident since the proposal involves neither a 
hardware modification nor the creation of a unique operating condition.
    3. Involve a significant reduction in a margin of safety. 
Removing the RHR spring relief valve as an additional requirement 
does not change the results of any of the FSAR Chapter 14 events. 
The PORVs remain operable to maintain the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director: 
John Zwolinski

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

    Date of amendment request: November 29, 1995
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3.14 to decrease the maximum steam 
generator (SG) primary-to-secondary leakage rate from 0.15 gpm to 0.10 
gpm and would modify TS 4.10 by revising the requirements for 
unscheduled SG tube inspections that are performed on each SG following 
a primary-to-secondary tube leak.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
A steam generator leakage assumption greater than the proposed 0.10 
gpm/SG limit has been used in the FSAR [Final Safety Analysis 
Report] Chapter 14 safety analyses. Thus, the FSAR Chapter 14 safety 
analyses remain bounding. Assuring that an adequate leakage limit 
exists that initiates corrective actions in a timely manner is 
important to ensuring a steam generator tube rupture event does not 
take place. This change modifies the steam generator post-leakage 
testing requirements to focus inspections on leaking tubes and areas 
likely to produce similar leakage, in lieu of an expanded test 
campaign of all three steam generators. Without this change, 
Technical Specifications require inspection of 3% of the tubes in 
each steam generator. By inspecting the critical areas of the 
affected steam generator and possibly expanding inspections to the 
critical areas of the remaining steam generators, the probability 
and/or consequences of previously evaluated accidents (e.g., steam 
generator tube rupture) are not increased.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed changes will not involve a modification to 
existing hardware at the plant. The decrease in the maximum 
allowable steam generator primary leakage rate tends to provide 
additional time for operator action to take place which, if timely 
enough, would avoid the consequences of a tube rupture event. The 
proposed inspection campaign requires inspection of the critical 
area and may be expanded to the other steam generators to ensure 
that additional tubes will not fail due to similar causes. This 
modified inspection campaign does not introduce the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. The FSAR Chapter 14 safety analyses assume a 
higher steam generator leakage rate and therefore remain 
conservative. The proposed reduction in the allowable leakage 
provides a greater margin of safety since it is more conservative 
than the present value. This change modifies inspection requirements 
of Technical Specifications and does not impact the plant design or 
equipment. The modified inspection requirements following a plant 
shutdown due to tube leakage concentrate steam generator tube 
inspections in those areas believed to be most susceptible to flaws. 
For these reasons, we believe the proposed changes increase the 
margin of safety by inspecting the critical areas of the steam 
generator(s) in lieu of additional random inspections.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, ME 04011NRC Deputy Director: 
John Zwolinski

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone 
Nuclear Power Station, Unit 1, New London County, Connecticut

    Date of amendment request: November 8, 1995
    Description of amendment request: The amendment request would 
revise the Technical Specifications (TS) for the jet pumps to be 
consistent with the limiting conditions for operation and surveillance 
requirements in the Standard Technical Specifications for General 
Electric Plants (NUREG-1433).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ...The proposed change does not involve an [significant hazards 
consideration] SHC because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The new LCO [Limiting Condition for Operation] does not diminish 
the existing requirement that all jet pumps must be operable, nor 
does it affect the time available to achieve cold shutdown should a 
pump become inoperable. The new LCO does eliminate the ability to 
continue to operate with the indication (but not the function) of a 
single jet pump inoperable. This does not increase the possibility 
of an unnecessary plant shutdown due to inoperable instrumentation 
since sufficient flexibility exists in the surveillance requirement 
so that operability of the jet pumps can be verified. This change 
eliminates the LCO that allowed continued operation with conditions 
that could potentially mask an inoperable pump. The new LCO is more 
limiting in ensuring that the plant is operated in a condition for 
which accidents were analyzed.
    The new surveillance requirement provides a more accurate method 
of ensuring

[[Page 13529]]
the jet pumps remain operable. The new surveillance criteria are 
more sensitive to jet pump failures and the degradation of the jet 
pumps prior to failure.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The new LCO and surveillance does not change the manner in which 
the plant is operated, nor does it reduce the operability 
requirements of any jet pump, Therefore, no new or different kind of 
accident can be created by the new specification. The surveillances 
that will be performed do not require any new hardware or plant 
evolutions. Therefore, the proposed change to the LCO and 
surveillance cannot create the possibility of a new or different 
kind of accident.
    3. Involve a significant reduction in the margin of safety.
    The margin of safety that currently exists is not diminished by 
this change. The requirement to place the reactor in cold shutdown 
within 24 hours should a jet pump become inoperable is maintained. 
The LCO which allowed continued operation with indication for one 
pump inoperable has been eliminated.
    The new surveillance requirement continues to demonstrate the 
operability of the jet pumps and during operation, continues to be 
performed at the same interval as in the current technical 
specifications. The note (which allows the surveillance to be 
deferred until four hours after the associated recirculation loop is 
in operation and 24 hours after exceeding 25% of rated thermal 
power) does not significantly affect the margin of safety. The time 
that the unit would be operating in these conditions would be small, 
and the stress placed on the pump at less than 25% power is lower.
    Based on the above, this change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: November 3, 1995
    Description of amendment request: The proposed amendment will 
extend the allowed outage time from 48 hours to 7 days for an emergency 
core cooling system train that is declared inoperable as a result of an 
inoperable low pressure safety injection subsystem.
    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:
    Pursuant to 10CFR50.92, Northeast Nuclear Energy Company (NNECO) 
has reviewed the proposed change to extend the allowed outage time 
(AOT) for an inoperable low pressure safety injection (LPSI) 
subsystem from the existing limit of 48 hours to 7 days. In 
addition, the change to modify the completion time for the Action 
Statement and the criteria for the Surveillance Requirements were 
also reviewed. NNECO concludes that these changes do not involve a 
significant hazards consideration (SHC) since the proposed change 
satisfies the criteria in 10CFR50.92(c). That is, the proposed 
change does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed amendments for Millstone Unit No. 2 will extend the 
action completion AOT for a single inoperable LPSI train from 48 
hours to 7 days. A LPSI subsystem is designed as a part of each 
emergency core cooling system (ECCS) train to supplement safety 
injection tank inventory during the early stages of mitigating a 
design basis accident (DBA). As such, components of the LPSI 
subsystem are not accident initiators, and an extended AOT to 
restore operability of an inoperable LPSI subsystem would not 
increase the probability of occurrence of accidents previously 
analyzed.
    The safety analyses for Millstone Unit No. 2 demonstrates that 
ECCS performance acceptance criteria are satisfied with only one of 
the two redundant ECCS trains operating during the postulated DBA. 
The proposed technical specification revisions involve the AOT for a 
single inoperable LPSI subsystem, and do not change the conditions 
assumed for the minimum amount of operating equipment needed for 
accident mitigation. Therefore, the consequences of an accident 
previously evaluated will not be significantly increased.
    In addition, CE NPSD-995 recognizes that when an ECCS train is 
inoperable due to a LPSI subsystem being unavailable, due either to 
being declared inoperable (by failing a surveillance requirement) or 
is intentionally taken out-of-service (for corrective or preventive 
maintenance), the core damage frequency (CDF) during power operation 
increases. The results of the PRA presented in CE NPSD-995 show that 
the proposed increase in the ECCS AOT (due to LPSI unavailability) 
from 48 hours to 7 days does not cause a significant increase in the 
overall CDF of Millstone Unit No. 2.
    The analyses indicate that continued plant operation with a 
single LPSI subsystem out-of-service may result in a small increase 
in ``at power risk;'' however, that risk increase will be negligibly 
small and controlled effectively via the Maintenance Rule and the 
risk monitor program that minimizes the outage time and prevents 
entering into an unacceptable risk configuration. In addition, the 
proposed AOT extension for the LPSI subsystem is evaluated as having 
negligible impact on the large early radiological release 
probability for Combustion Engineering pressurized water reactors in 
the event of a design basis accident.
    Therefore, operation in accordance with the proposed amendment 
would not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed amendment will not change the physical plant or the 
modes of plant operation defined in the technical specifications. 
The changes do not involve the addition or modification of equipment 
nor do they alter the design of plant systems. Therefore, operation 
of Millstone Unit No. 2 in accordance with its proposed amendment 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The margin of safety associated with the ECCS train is 
established by acceptance criteria for system performance defined in 
10CFR50.46. The proposed amendment will not change this acceptance 
criteria nor the operability requirements for equipment that is used 
to achieve such performance as demonstrated in the Millstone Unit 
No. 2 safety analyses. Moreover, an integrated assessment of the 
risk impact of extending the AOT for a single inoperable LPSI train 
has concluded that the risk contribution is small. Therefore, 
operation of Millstone Unit No. 2 in accordance with its proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.

[[Page 13530]]

    NRC Project Director: Phillip F. McKee

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: September 12, 1995
    Description of amendment request: The amendment would revise and 
reformat Technical Specification (TS) 6.3.1 to add the requirement that 
the Assistant Operations Manager shall hold a senior reactor operator 
(SRO) license if the Operations Manager does not hold an SRO license 
for Millstone Unit 3. Also the footnote would be deleted from TS 6.3.1 
that previously granted a one-time three year exception to the 
qualification requirements for the Operations Manager and an exception 
for the Assistant Operations Manager to hold a license instead of the 
Operations Manager.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ...The proposed change does not involve an [significant hazards 
consideration] SHC because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed change affects an administrative control, which was 
based on the guidance of ANSI N18.1-1971. ANSI N18.1-1971 
recommended that the Operations Manager hold an SRO license. The 
current guidance in Section 4.2.2 of ANSI/ANS 3.1-1987 recommends, 
as one option, that the Operations Manager have held a license for a 
similar unit and the Operations Middle Manager hold an SRO license. 
While the Operations Middle Manager position does not exist at 
Millstone Unit No. 3, [Northeast Nuclear Energy Company] NNECO has 
created the position of Assistant Operations Manager. The individual 
in this position would meet the requirements for, and would have 
responsibilities as recommended in, ANSI/ANS 3.1-1987 for the 
Operations Middle Manager position.
    Therefore, the proposed change requests an exception to ANSI 
N18.1-1971 to allow use of ANSI/ANS 3.1-1987 in a limited 
circumstance. Specifically, the proposed revision to Technical 
Specification 6.3.1 would require the Operations Manager to either 
hold an SRO license at Millstone Unit No. 3 or have held an SRO at a 
[pressurized water reactor] PWR.
    If the Operations Manager does not hold an SRO license at 
Millstone Unit No. 3, the specification will require the Assistant 
Operations Manager to hold, and continue to hold, an SRO license. 
The proposed change includes the requirement for the Operations 
Manager to have held a license for a similar unit (a PWR) in 
accordance with Section 4.2.2 of ANSI/ANS 3.1-1987. For those areas 
of knowledge that require an SRO license, the Assistant Operations 
Manager will provide the technical guidance normally provided by the 
Operations Manager.
    The proposed change does not alter the design of any system, 
structure, or component, nor does it change the way plant systems 
are operated. It does not reduce the knowledge, qualifications, or 
skills of licensed operators, and does not affect the way the 
Operations Department is managed by the Operations Manager. The 
Operations Manager will continue to maintain the effective 
performance of his personnel and ensure the plant is operated safely 
and in accordance with the requirements of the operating license. 
Additionally, the Control Room Operators will continue to be 
supervised by the licensed Shift Supervisors.
    The proposed change does not detract from the Operations 
Manager's ability to perform his primary responsibilities. In this 
case, by having previously held an SRO license, the Operations 
Manager has achieved the necessary training, skills, and experience 
to fully understand the operation of plant equipment and the watch 
requirements for operators. In summary, the proposed change does not 
affect the ability of the Operations Manager to provide the plant 
oversight required of his position. Thus, it does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change to Technical Specification 6.3.1 does not 
affect the design or function of any plant system, structure, or 
component, nor does it change the way plant systems are operated. It 
does not affect the performance of NRC licensed operators. Operation 
of the plant in conformance with technical specifications and other 
license requirements will continue to be supervised by personnel who 
hold an NRC SRO license. The proposed change to Technical 
Specification 6.3.1 ensures that the Operations Manager will be a 
knowledgeable and qualified individual to have held an SRO license 
at a PWR. Based on the above, the proposed change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The proposed change involves an administrative control that is 
not related to the margin of safety. The proposed change does not 
reduce the level of knowledge or experience required of an 
individual who fills the Operations Manager position, nor does it 
affect the conservative manner in which the plant is operated. The 
Control Room Operators will continue to be supervised by personnel 
who hold an SRO license. Thus, 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: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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: November 21, 1995
    Description of amendment request: The licensee proposes to change 
Technical Specification Section 1.33 and Bases Sections 3/4.3.3.9 and 
3/4.3.3.10, and 3/4.11.2.1. The changes clarify the definition of 
source check to include a source check from a light emitting diode 
(LED), as well as from ionizing radiation.
    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:
    ... NNECO concludes that these changes do not involve a 
significant hazards consideration since the proposed changes satisfy 
the criteria in 10CFR50.92(c). That is, the proposed changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed changes to the definition of source check clarifies 
the source check for the liquid and gaseous effluent radiation 
monitors. These monitors do not provide a safety function and only 
serve to provide radiological information to plant operators, 
therefore, the changes will not increase the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed changes to the definition of source check have no 
effect on the ability of the monitors to perform their designed 
function. The clarification to the surveillance do not involve any 
physical modifications to any equipment, structures, or components. 
The monitors already have the internal LEDs which were originally 
used to perform the source check. The proposed changes have no 
impact on design basis accidents, and the changes will not modify 
plant response or create a new or unanalyzed event.
    3. Involve a significant reduction in the margin of safety.
    
[[Page 13531]]

    The proposed changes to the definition of source check do not 
have any impact on the protective boundaries and, therefore, have no 
impact on the safety limits for these boundaries. The 
instrumentation associated with these changes do not provide a 
safety function and only serve to provide radiological information 
to plant operators. The instrumentation has no affect on the 
operation of any safety-related equipment. As such, these changes 
have no impact on the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of application for amendments: February 15, 1996
    Description of amendment request: The amendment changes the 
Technical Specifications to implement 10 CFR Part 50, Appendix J, 
Option B, by creating Technical Specification Section 5.5.12, ``Primary 
Containment Leakage Rate Testing Program,'' which refers to Regulatory 
Guide 1.163, ``Performance-Based Containment Leakage-Test Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1) The proposed changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated.
    The adoption of 10 CFR 50, Appendix J Option B will not involve 
a significant increase in the probability or consequences of any 
accident previously evaluated. The proposed changes to the TS 
[Technical Specifications] reflect the use of the performance-based 
containment leakage-testing program. The USNRC has approved the use 
of a performance-based option for containment leakage testing 
programs when it amended 10 CFR 50, Appendix J (60 FR 49495). For 
adoption of the revised regulation, licensees are required to 
incorporate into their TS, by general reference, the USNRC 
regulatory guide or other plant-specific implementing document used 
to develop their performance-based leakage testing program. A new 
Administrative Control subsection (5.5.12, ``Primary Containment 
Leakage Rate Testing Program'') has been added that requires the 
establishment and maintenance of a Primary Containment Leakage Rate 
Testing Program. The TS will still require the performance of a 
periodic general visual inspection of the containment to ensure 
early detection of any structural deterioration of the containment 
that may occur.
    As concluded in NUREG-1493, given the insensitivity of risk to 
containment leakage rate and the small fraction of leakage paths 
detected solely by ILRT [Integrated Leak Rate Test] testing, 
increasing the interval between ILRTs is possible with minimal 
impact on public risk. Additionally, performance-based alternatives 
to current LLRT [Local Leak Rate Test] requirements are feasible 
without significant risk impacts. Additionally, these changes will 
not alter any safety limits which ensure the integrity of fuel 
barriers, and will not result in a significant increase to onsite or 
offsite dose.
    No physical changes are being made to the plant, nor are there 
any changes being made in the operation of the plant as a result of 
these changes which could involve a significant increase in the 
probability or consequences of any accident previously evaluated. 
Additionally, these changes will not alter the operation of 
equipment assumed to be available for the mitigation of accidents or 
transients.
    2) The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The adoption of 10 CFR 50, Appendix J Option B will not create 
the possibility of a new or different type of accident from any 
previously evaluated. These changes to the PBAPS, Units 2 and 3 TS 
will not involve any changes to plant systems, structures or 
components (SCCs) which could act as new accident initiators. These 
changes will not impact the manner in which SSCs are tested such 
that a new or different type of accident from any previously 
evaluated could be created.
    3) The proposed changes do not result in a significant reduction 
in the margin of safety.
    No margins of safety are reduced as a result of the proposed 
adoption of 10 CFR 50, Appendix J Option B. As stated previously, 
the USNRC has approved the use of this performance-based option for 
containment leakage testing programs when it amended 10 CFR 50, 
Appendix J (60 FR 49495). These changes will not impact core limits 
or any other parameters that are used in the mitigation of a UFSAR 
[Updated Final Safety Analysis Report] design-basis accident or 
transient. Additionally, these changes do not introduce any hardware 
changes, and will not alter the intended operation of plant 
structures, systems or components utilized in the mitigation of 
UFSAR design-basis accidents or transients. These changes will not 
introduce any new failure modes of plant equipment not previously 
evaluated.
    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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101
    NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket No. 50-387, 
Susquehanna Steam Electric Station, Unit 1, Luzerne County, 
Pennsylvania
    Date of amendment request: January 26, 1996
    Description of amendment request: The proposed amendment removes 
three pressure relief valves from Technical Specification Table 3.6.3-
1, ``Primary Containment Isolation Valves,'' since these valves are no 
longer needed to support the steam condensing mode of the residual heat 
removal (RHR) system and are being removed from the plant.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    With the prior deletion of the steam condensing mode of RHR and 
the isolation of the high and low pressure interfaces, the three 
pressure relief valves that are being removed from the plant have no 
active function. Their passive function of maintaining system or 
containment integrity will be fulfilled by blind flanges. Also, the 
RHR and RCIC [reactor core isolation cooling] piping are provided 
with overpressure protection from other pressure relief valves. 
Therefore, the removal of these pressure relief valves does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The pressure relief valves that are being removed had two 
primary functions. First,

[[Page 13532]]

they provided overpressure protection for the RHR and RCIC piping 
during the steam condensing mode of RHR. Since the steam condensing 
mode has been deleted from the plant, these valves no longer have 
that function. Also, overpressure protection of the RHR and RCIC 
piping is provided by other existing pressure relief valves. Second, 
these valves maintained system or containment integrity. When the 
pressure relief valves are removed from the plant, they will be 
replaced with blind flanges or equivalent that will maintain system 
or containment integrity. Therefore, the removal of the three 
pressure relief valves 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.
    Since the steam condensing mode of RHR has been eliminated, the 
three pressure relief valves have no active function. Their passive 
function of maintaining system or containment integrity will be 
fulfilled by blind flanges or equivalent. Also, overpressure 
protection of RHR and RCIC piping is provided by other existing 
pressure relief valves. Therefore, the removal of the three pressure 
relief valves 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:  Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 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

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

    Date of amendment request: January 25, 1996
    Description of amendment request: The amendment proposes to revise 
the allowed out-of-service times for single inoperable Emergency Diesel 
Generators (EDGs) to accommodate on-line maintenance of the EDGs. In 
addition, two line item changes are proposed: (1) to improve safety by 
reducing EDG testing at power; and (2) to revise the ac power 
requirements during cold shutdown or refueling modes to make the James 
A. FitzPatrick (JAF) Technical Specifications consistent with the 
Standard Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    a. EMERGENCY DIESEL GENERATOR LCO [Limiting Conditions for 
Operation] AT POWER
    The proposed changes to the Technical Specifications will allow 
longer Allowed Out of Service Times [AOTs] to perform necessary 
repair and maintenance on individual Emergency Diesel Generators 
while at power. This extended AOT will enhance scheduling of 
preventive maintenance of individual EDGs without significantly 
increasing the probability or consequences of an accident previously 
evaluated. The risk evaluations contained in the JAF quantitative 
analyses of the EDGs determined that the probability of an accident 
by increasing the AOT for an individual EDG from 7 days to 14 days 
is non-risk-significant. The primary reason for this low relative 
risk is due to the designed redundancy and capability to respond to 
an accident when a single diesel generator is out of service. LOCA 
[loss-of-coolant accident] Analyses that assume the worst case line 
break while an EDG is out of service indicate the plant can be 
safely shut down with the remaining EDGs. Even if another EDG should 
fail during the AOT, at least one Core Spray and one Residual Heat 
Removal (RHR) Low Pressure Coolant Injection pump can provide the 
required flow to bring the plant to safe shut down. Furthermore, 
long term suppression pool and reactor shutdown cooling is provided 
by any one of the three remaining RHR pumps for a single EDG out of 
service or by two remaining RHR pumps assuming an additional EDG 
failure during the AOT.
    Increasing the EDG AOT does not involve physical alteration of 
any plant equipment and does not affect analysis assumptions 
regarding functioning of required equipment designed to mitigate the 
consequences of accidents. Further, the severity of postulated 
accidents and resulting radiological effluent releases will not be 
affected by the increased AOT for a single EDG.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN
    Changing the number of EDGs required during plant shutdown does 
not involve physical alteration of any plant equipment and does not 
affect analysis assumptions regarding functioning of required 
equipment designed to mitigate the consequences of accidents. 
Further, the severity of postulated accidents and resulting 
radiological effluent releases will not be affected by the change in 
the LCO during shutdown.
    c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION
    The proposed change to the Technical Specification will reduce 
the required number of tests to be performed when an EDG or EDG 
System is inoperable. This proposed change to TS requirements 
addresses the concern of excessive testing that could result in EDG 
wear which is counter-productive to safety in terms of equipment 
degradation and availability. This change is consistent with Generic 
Letter 93-05 guidance for implementing such recommendations. The 
proposed Technical Specifications will not result in a change to the 
design or operation of the facility, therefore, this change will not 
result in a significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    a. EMERGENCY DIESEL GENERATOR LCO AT POWER
    Extending the AOT for an individual EDG does not necessitate 
physical alteration of the plant or changes in parameters governing 
normal plant operation. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated for JAF plant.
    b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN
    Changing the number of EDGs required during shutdown does not 
necessitate physical alteration of the plant or changes in 
parameters governing normal plant operation. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated for JAF plant.
    c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION
    The proposed change does not change design, operation or the 
testing process. The nature of this change precludes the possibility 
of a new or different kind of accident. The proposed change to 
complete the required action does not involve any hardware changes, 
nor changes to the operation of the equipment nor does it change the 
ability of the equipment to perform its intended function. 
Performing the testing on an extended time cannot initiate any type 
of accident.
    3. Involve a significant reduction in the margin of safety.
    a. EMERGENCY DIESEL GENERATOR LCO AT POWER
    As discussed above, the JAF quantitative evaluation determined 
that the change in risk associated with extending the AOT for a 
single EDG is non-risk-significant. In addition, the design provides 
adequate redundancy for safe shut down during the AOT for a single 
EDG out of service. This is supported by the LOCA analyses including 
analyses for long term suppression pool and reactor shutdown 
cooling.
    b. EMERGENCY DIESEL GENERATOR LCO DURING PLANT SHUTDOWN
    The margin of safety is not affected by changing the number of 
EDGs required during shutdown. One offsite power source or one EDG 
ensure the availability of the

[[Page 13533]]

required power to recover from postulated accident events during 
shutdown. When the required number of operable systems is not met, 
all work that could potentially initiate a postulated accident event 
during shutdown is suspended.
    c. EMERGENCY DIESEL GENERATOR SURVEILLANCE AT POWER OPERATION
    The proposed change to Technical Specifications reduces testing 
at reactor power. The overall effect is a net gain in plant safety 
by avoiding the potential for unnecessary wear that could degrade 
the EDGs at power. Implementation of these changes is consistent 
with the guidance provided by the NRC in Generic Letter 93-05. The 
proposed change to the EDG testing requirements does not reduce the 
ability of the equipment to perform its intended safety function.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Ledyard B. Marsh
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
    Date of amendment request: January 30, 1996
    Description of amendment request: The proposed Technical 
Specifications change will delete the requirement that oxygen 
concentrations for both normal and transient conditions not exceed 
saturation when the reactor coolant is below 250 degrees F. The 
Technical Specifications change will also eliminate the surveillance 
requirement for reactor coolant chemistry sampling of chloride, 
fluoride, and oxygen concentration during maintenance activities when 
fuel is removed from the reactor vessel and the Reactor Coolant System 
(RCS) is drained below the reactor vessel flange regardless of whether 
the upper internal and/or vessel heat are in place or not. 
Administrative result of the changes being made, capitalize Technical 
Specifications defined terms to maintain consistency within the 
Technical Specifications, and the word ``degrees'' is spelled-out when 
referring to the Fahrenheit temperature, rather than using the symbol.
    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:
    Specifically, operation of Surry Power Station in accordance 
with the proposed changes will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Since the RCS and the RHR [Residual Heat Removal] System are 
drained when the RCS inventory is reduced below the reactor vessel 
flange for maintenance or refueling activities, the concentrations 
of chlorides and fluorides will not change. During these maintenance 
or refueling activities, only controlled makeup to the RCS is 
planned, and any planned or unplanned makeup to the RCS would be 
detected by available level indication. Sampling for chloride and 
fluoride concentrations in the RCS will be performed prior to 
draining the system. Sampling of the reactor coolant for chloride 
and fluoride concentrations will resume when the RCS is filled. The 
chloride and fluoride concentrations will be known and will be 
maintained consistent with the Technical Specification Limiting 
Condition for Operation and Action Statements. Also, when the RCS 
inventory is drained below the reactor vessel flange, the RCS is 
vented and open to the containment building atmosphere with the 
reactor coolant liquid considered oxygen saturated. Technical 
Specification 3.1.F.4 allows normal and off-normal ``saturated'' 
oxygen concentrations when reactor coolant temperature is below 250 
degrees F. Consequently, sampling the reactor coolant for oxygen 
concentration under these conditions is not required and the 
Technical Specification Table 4.1-2B specified sampling frequency of 
five (5) times per week is not necessary since the oxygen 
concentration continues to remain in compliance with the Technical 
Specification limit, measures are available and action can be taken 
to correct the condition prior to any deleterious effect.
    Surry Technical Specifications 3.1.F.1 prohibits reactor coolant 
temperature from exceeding 250 degrees F unless chloride, fluoride, 
and oxygen concentrations are within specified limits. Therefore a 
significant increase in the probability or consequences of an 
accident previously evaluated does not exist.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The materials that are exposed to reactor coolant are corrosion 
resistant. They were chosen for specific applications within the 
system and for their compatibility with the reactor coolant. The 
chemical composition of the reactor coolant will be maintained 
within the specifications given within Technical Specification 
3.1.F, Updated Final Safety Analysis Report Table 4.2-2, and 
Technical Specification Table 4.1-2B. Because of the time dependent 
nature of any adverse affects from chloride, fluoride, and oxygen 
concentrations in excess of the Technical Specifications limits, 
measures are available and can be taken to correct the condition 
while the reactor is in a safe shutdown condition, prior to any 
deleterious effect. No hardware modifications are involved. System 
configuration and plant operations are not being changed. Therefore, 
the possibility of a new or different kind of accident from any 
accident previously evaluated has not been created.
    (3) Involve a significant reduction in the margin of safety.
    This change does not involve a significant reduction in the 
margin of safety since the chloride and fluoride concentrations are 
maintained within their specified values prior to RCS drain down and 
following refill. The time period during which the RCS inventory is 
reduced below the reactor vessel flange and fuel is removed from the 
vessel, is short and insignificant in terms of the parameters 
necessary to initiate a corrosion concern. Existing Technical 
Specifications Action Statements and Allowed Technical Specification 
values for normal and off-normal concentrations of chlorides and 
fluorides are not being changed. No hardware modifications are 
involved. System configuration and plant operations are not being 
changed. Surry Technical Specification 3.1.F.1 remains unaffected by 
this change and continues to prohibit reactor coolant temperature 
from exceeding 250 degrees F unless chloride, fluoride, and oxygen 
concentrations are within specified limits.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: Eugene V. Imbro
Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing
    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued

[[Page 13534]]

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.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket No. 50-498, South Texas Project, Unit 1, Matagorda County, 
Texas

    Date of amendment request: February 29, 1996
    Description of amendment request: The proposed amendment would 
include the addition of Technical Specification 3.10.8 which would 
allow a one-time only extension of the standby diesel generator (SDG) 
allowed outage time for a cumulative 21 days on ``A'' train SDG. In 
addition, it would also allow a one-time only extension of the allowed 
outage time on ``A'' train essential cooling water loop for a 
cumulative 7 days. This one-time only change would become effective on 
April 10, 1996, and expire on May 15, 1996.Date of individual notice in 
the Federal Register: March 8, 1996 (61 FR 9502)
    Expiration date of individual notice: April 8, 1996
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
Matagorda County, Texas

    Date of amendment request: May 1, 1995, as supplemented by letters 
dated June 22, August 28, November 22, and December 19, 1995, and 
January 4, January 8 (two letters), and January 23, 1996
    Description of amendment request: The proposed amendment would 
provide a special test exception that would allow an extension of the 
standby diesel generator (SDG) allowed outage time for a cumulative 21 
days on each SDG once per fuel cycle, and it would also allow an 
extension of the essential cooling water (ECW) loop allowed outage time 
for a cumulative 7 days on each ECW loop once per fuel cycle. These 
extended allowed outage times will be used to perform required 
inspections and maintenance on the SDGs and the ECW system during power 
operation.
    Date of individual notice in the Federal Register: February 8, 1996 
(61 FR 4805)
    Expiration date of individual notice: March 11, 1996
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

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

    Date of amendment request: March 1, 1996 (supersedes December 11, 
1995, application)
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 4.7, ``Surveillance Requirements 
for Primary Containment Automatic Isolation Valves.'' Specifically, the 
proposed amendment would revise the replacement frequency of the seat 
seals for the drywell and suppression chamber purge and vent valves 
from every 5 years to every six operating cycles.
    Date of individual notice in the Federal Register: March 8, 1996 
(61 FR 9504)
    Expiration date of individual notice: April 8, 1996
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

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 1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: November 7, 1995, as 
supplemented by letter dated January 17, 1996.
    Brief description of amendments: These amendments adopt the 
improved Standard Technical Specifications (NUREG-1432) format and 
content of Section 5.0, ``Design Features,'' as modified by approved 
changes to the improved Standard Technical Specifications.
    Date of issuance: March 6, 1996
    Effective date: March 6, 1996, to be implemented within 45 days of 
the date of issuance.
    Amendment Nos.: Unit 1 - Amendment No. 104; Unit 2 - Amendment No. 
93; Unit 3 - Amendment No. 76
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65673) The January 17, 1996, supplemental letter provided clarifying 
information and did not change the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated March 6, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location:  Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

[[Page 13535]]


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

    Date of application for amendment: February 16, 1996
    Brief description of amendment: The amendment allows a one-time 
extension for the performance of the trip actuating device operational 
test for one of the safety injection manual initiation switches listed 
in Technical Specification Table 4.3-2, Item 1a.Date of issuance: March 
11, 1996
    Effective date: March 11, 1996
    Amendment No. 63
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: Yes (61 FR 7125). That notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by March 27, 1996, but indicated that if the Commission makes a 
final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment. The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, and final determination of no significant hazards 
consideration is contained in a Safety Evaluation dated March 11, 1996
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket 
Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 
2, Will County, Illinois

    Date of application for amendments: January 11, 1996
    Brief description of amendments: The amendments revise the action 
statements and allowed outage time for inoperability of one channel and 
both channels of source range neutron flux instrumentation in Shutdown 
Modes 3, 4, and 5.
    Date of issuance: March 15, 1996
    Effective date: March 15, 1996
    Amendment Nos.: 80, 80, 72, and 72
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 31, 1996 (61 FR 
3509) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 15, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

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

    Date of application for amendments: November 14, 1995, as 
supplemented January 4, 1996 and February 29, 1996.
    Brief description of amendments: The amendments revise the 
Technical Specifications to incorporate 10 CFR Part 50, Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors,'' Option B.
    Date of issuance: March 11, 1996 Effective date: Immediately, to be 
implemented no later than June 30, 1996.
    Amendment Nos.: 110 and 95
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1995 (60 FR 
62896) The January 4, 1996, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
March 11, 1996. No significant hazards consideration comments received: 
No
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

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

    Date of application for amendment: September 20, 1995, as 
supplemented December 18 and December 22, 1995.
    Brief description of amendment: The amendment allows a one-time 
surveillance interval extension for certain 18-month surveillances 
listed in new Technical Specification Tables 4.0.2-1 and 4.0.2-2. Date 
of issuance: March 1, 1996
    Effective date:
    March 1, 1996, with full implementation within 90 days.
    Amendment No.: 106
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58400). The December 18, 1995, letter corrected a typographical 
error on one of the proposed TS pages and provided a corrected Table of 
Contents page to reflect the addition of the new Tables. The December 
22, 1995, letter provided additional information on the licensee's 
review of historical plant drift data. This information was within the 
scope of the original application and did not change the staff's 
initial no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 1, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

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

    Date of application for amendments: November 10, 1995
    Brief description of amendments: The amendments revise the 
Technical Specifications for containment systems to reflect the 
adoption of the requirements of 10 CFR Part 50, Appendix J, Option B, 
and the implementation of a performance-based containment leak-rate 
testing program at the Edwin I. Hatch Nuclear Plant, Units 1 and 2.
    Date of issuance: March 6, 1996
    Effective date: As of the date of issuance to be implemented within 
90 days
    Amendment Nos.: Unit 1 - 200 - Unit 2 - 141
    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65679) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 6, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

[[Page 13536]]


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

    Date of application for amendment: December 5, 1995
    Brief description of amendment: The amendment revises the submittal 
date for the Annual Exposure Data Report bringing Oyster Creek into 
conference with 10 CFR 20.2206 and relaxes an overly restrictive 
administrative requirement.
    Date of Issuance: March 4, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 183
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1629). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated March 4, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

    Date of application for amendment: December 14, 1995
    Brief description of amendment: The amendment modifies Technical 
Specification 3.4.2, ``Flow Control Valves (FCVs),'' by deleting 
Surveillance Requirement (SR) 3.4.2.2, which required periodic 
verification that the average rate of movement of each reactor 
recirculation system FCV was limited to less than or equal to 11% per 
second in the opening and closing directions. Due to a plant 
modification, the requirement is not applicable.
    Date of issuance: March 11, 1996
    Effective date: March 11, 1996
    Amendment No.: 103
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1630) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 11, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: November 10, 1995 
(AEP:NRC:0896X). This application superseded a request dated June 15, 
1995 (AEP:NRC:0896V).
    Brief description of amendments: The amendments change the 18-month 
emergency diesel generator surveillance test from a 24-hour run to an 
8-hour run and add voltage and frequency measurement and power factor 
monitoring.
    Date of issuance: March 11, 1996
    Effective date: March 11, 1996, with full implementation within 45 
days
    Amendment Nos.: Unit 1 - 207, Unit 2 - 191
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65682) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 11, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: June 20, 1995, as supplemented 
December 19, 1995.
    Brief description of amendments: The amendments relocate the fire 
protection program elements from the Technical Specifications and 
incorporate, by reference, the NRC-approved Fire Protection Program and 
major commitments, including the fire hazards analysis, into the 
Updated Final Safety Analysis Report. In addition, the amendments 
revise the operating licenses to include the NRC's standard fire 
protection license condition.
    Date of issuance: March 11, 1996
    Effective date: March 11, 1996, with full implementation within 180 
days
    Amendment Nos.: Unit 1 - 208, Unit 2 - 192
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications and the operating licenses.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47620). The December 19, 1995, supplement clarified the license 
conditions by providing specific approval dates for previous fire 
protection safety evaluations. This information was within the scope of 
the original application and did not change the staff's initial 
proposed no significant hazards consideration determination.The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated March 11, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: June 29, 1995
    Brief description of amendment: The amendment revises the Technical 
Specifications to extend the surveillance schedule from 18 months to 
each refueling interval (nominally 24 months) for specifications 
4.6.4.2, 4.7.1.2.1.c, 4.7.3.b, 4.7.4.b,and 4.7.10.e. It also deletes 
specification 4.6.4.2.a and the phrase ``during shutdown'' from these 
specifications.Date of issuance: March 4, 1996
    Effective date: As of the date of issuance, to be implemented 
within 90 days.
    Amendment No.: 127
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58402) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 4, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon 
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California

    Date of application for amendment: January 18, 1996
    Brief description of amendment: The amendment revises the combined 
Technical Specifications (TS) for the Diablo Canyon Nuclear Power 
Plant, Unit No. 1. TS 3.8.1.1, ``Electrical Power Systems - A.C. 
Sources - Operating,'' is revised to allow operation of Unit 1 in Mode 
3 (Hot Standby) during installation of a replacement non-vital 
auxiliary transformer 11, for a one time

[[Page 13537]]
extension of up to 48 hours beyond the 72 hours allowed by TS 3.8.1.1, 
Action Statement (a).
    Date of issuance: March 8, 1996
    Effective date: March 8, 1996
    Amendment No.: Unit 1 - Amendment No. 111
    Facility Operating License No. DPR-80: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1996 (61 FR 
3737) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 8, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendments: December 27, 1995
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) 3/4.6.1.1, Containment 
Integrity; 3/4.6.1.2, Containment Leakage; 3/4.6.1.3, Containment Air 
Locks; 3/4.6.1.6, Containment Structural Integrity; 3/4.6.3, 
Containment Isolation Valves; their associated Bases; and adds 
Specification 6.8.4 j., Containment Leakage Rate Testing Program to 
implement the performance based leakage rate testing program as 
permitted by 10 CFR Part 50, Appendix J, rather than paraphrasing the 
requirements of the regulation. These changes will support the 
implementation of the performance based testing of Option B to Appendix 
J, for Type A, B, and C containment leakage rate testing and the 
appropriate rescheduling of testing.
    Date of issuance: March 1, 1996 Effective date: March 1, 1996
    Amendment Nos.: Unit 1 - Amendment No. 110; Unit 2 - Amendment No. 
109
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 31, 1996 (61 FR 
3502) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 1, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendment: April 13, 1994, as supplemented 
December 6, 1995
    Brief description of amendment: The proposed changes revise the 
Quality Assurance audit frequencies in the Hope Creek Technical 
Specifications. These revisions will permit an audit frequency based on 
performance and transfer subsequent control over the audit program to 
the Updated Final Safety Analysis Report.
    Date of issuance: March 11, 1996
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 95
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29633) The December 6, 1995, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination nor the original Federal Register 
notice.The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 11, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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: April 13, 1994, as supplemented 
December 6, 1995.
    Brief description of amendments: The proposed changes revise the 
Quality Assurance audit frequencies in the Salem Unit Nos. 1 and 2 
Technical Specifications. These revisions will permit an audit 
frequency based on performance and transfer subsequent control over the 
audit program to the Updated Final Safety Analysis Report.
    Date of issuance: March 11, 1996
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment Nos. 181 and 162
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29633) The December 6, 1995, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination nor the original Federal Register 
notice.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 11, 1996No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendments: December 8, 1995 (TS 93-09)
    Brief description of amendments: The amendments revise the 
setpoints and time delays for the auxiliary feedwater loss-of-power and 
the 6.9-kilovolt shutdown board loss-of-voltage and degraded voltage 
instruments.
    Date of issuance: March 1, 1996
    Effective date: March 1, 1996
    Amendment Nos.: 219 and 209
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
181) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 1996.No significant hazards 
consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendments: January 4, 1996 (TS 95-22)
    Brief description of amendments: The amendments change the 
surveillance test frequency specified for the functional tests of the 
containment, fuel storage pool, and control room radiation monitors 
from monthly to quarterly.
    Date of issuance: March 4, 1996
    Effective date: March 4, 1996
    Amendment Nos.: 220 and 210
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: January 31, 1996 (61 FR 
3503)

[[Page 13538]]
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated Macrh 4, 1996.No significant hazards 
consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendment: January 16, 1996, and supplement 
dated March 1, 1996
    Brief description of amendment: This amendment approves that part 
of the request that defers the drywell bypass leakage test during the 
current refueling outage. The remainder of the licensee's request is 
still under NRC staff review.
    Date of issuance: March 8, 1996
    Effective date: March 8, 1996
    Amendment No. 82
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1996 (61 FR 
3951) The March 1, 1996, supplemental letter was clarifying in nature 
and did not affect the initital no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 8, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment: December 9, 1994, as 
supplemented by letters dated September 13, 1995, and February 9, 1996.
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) 4.3.2.2, TS 4.7.1.2.1, and the Bases for TS 3/4 
7.1.2 to decrease the frequency of auxiliary feedwater pump testing, 
remove inconsistencies in testing requirements for the turbine-driven 
auxiliary feedwater pump, and clarify performance parameters in the TS 
Bases.
    Date of issuance: March 11, 1996
    Effective date: March 11, 1996, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 108
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6314). The September 13, 1995, and February 9, 1996, supplemental 
letters provided additional clarifying information and 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 11, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendments: September 19, 1995
    Brief description of amendments: The amendments revised the maximum 
allowable power range neutron flux high setpoints for operation with 
inoperable main steam safety valves.
    Date of issuance: March 6, 1996
    Effective date: March 6, 1996
    Amendment Nos.: 199 and 180
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54724) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 6, 1996No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of amendment request: November 22, 1995
    Brief description of amendment: The amendment replaces the 
Technical Specification (TS) requirements associated with the boron 
dilution mitigation system (BDMS) with alarms, indicators, procedures 
and controls to allow proper resolution of potential boron dilution 
events.
    Date of issuance: March 1, 1996
    Effective date: March 1, 1996, to be implemented prior to the 
startup from the eighth refueling outage.
    Amendment No.: 96
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 31, 1996 (61 FR 
3503) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 1996.No significant hazards 
consideration comments received: No. Local Public Document Room 
locations: Emporia State University, William Allen White Library, 1200 
Commercial Street, Emporia, Kansas 66801 and Washburn University School 
of Law Library, Topeka, Kansas 66621
    Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas
    Date of amendment request: December 20, 1995, as supplemented by 
letter dated February 8, 1996.
    Brief description of amendment: The amendment revises the Technical 
Specifications to reflect the approval of the use of 10 CFR Part 50, 
Appendix J, Option B for the Wolf Creek Generating Station containment 
leakage rate test program.
    Date of issuance: March 1, 1996
    Effective date: March 1, 1996, to be implemented prior to startup 
from the eighth refueling outage.
    Amendment No.: 97
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 31, 1996 (61 FR 
3504) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

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

    Date of amendment request: December 13, 1995
    Brief description of amendment: The amendment revises the minimum 
and maximum flow requirements for the centrifugal charging pumps (CCPs) 
and safety injection pumps (SIPs) specified in Technical Specification 
(TS) Surveillance Requirement 4.5.2.h. Specifically, the amendment (1) 
decreases the minimum limits on the sum of the injection line flow 
rates,

[[Page 13539]]
excluding the highest flow rate, from 346 gallons per minute (gpm) to 
330 gpm for the CCPs and from 459 gpm to 450 gpm for the SIPs, and (2) 
revises the maximum pump flow rate for the SIPs from 665 to 670 gpm, 
but retains the CCPs maximum pump flow rate at its current value of 556 
gpm.Date of issuance: March 5, 1996
    Effective date: March 5, 1996, to be implemented prior to startup 
from the eighth refueling outage.
    Amendment No.: 98
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1639) The February 5, 1996, supplemental letter provided additional 
clarifying information and 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 5, 1996.No significant hazards consideration comments received: 
No. Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

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 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 26, 1996, 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

[[Page 13540]]
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. 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, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, 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).

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

    Date of application for amendment: March 6, 1996
    Brief description of amendment: This amendment revises TS 3/4 5.2, 
ECCS SUBSYSTEMS - T <INF>avg greater than or equal to 280 deg.F by 
modifying Surveillance Requirement 4.5.2.b to defer venting of the 
Emergency Core Cooling System flow path which does not have manual 
venting capability until the tenth refueling outage.
    Date of issuance: March 7, 1996
    Effective date: March 7, 1996
    Amendment No: 208
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendments, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated March 7, 1996.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus
    Dated at Rockville, Maryland, this 20th day of March 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga, Director,
Division of Reactor Projects - I/II,Office of Nuclear Reactor 
Regulation
[Doc. 96-7259 Filed 3-26-96; 8:45 am]
BILLING CODE 7590-01-F