X96-10731. Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

[Federal Register Volume 61, Number 148 (Wednesday, July 31, 1996)]
[Notices]
[Pages 40013-40035]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-10731]


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NUCLEAR REGULATORY COMMISSION
Biweekly Notice


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 6, 1996, through July 19, 1996. The 
last biweekly notice was published on July 17, 1996 (61 FR 37295).

Notice Of Consideration Of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, And Opportunity For A Hearing

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

[[Page 40014]]

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

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

    Date of amendment request: March 25, 1996 (NRC-96-0003)
    Description of amendment request: The proposed amendment would 
modify the charcoal testing standards for the Control Room Emergency 
Filtration System (CREFS) and the Standby Gas Treatment System (SGTS) 
to the current industry standard. The changes affect Surveillance 
Requirements (SRs) 4.6.5.3.b.2, 4.6.5.3.c, 4.7.2.1.c.2, and 4.7.2.1.d 
in Technical Specifications (TS) 3/4.6.5.3 ``Standby Gas Treatment 
System'' and TS 3/4.7.2 ``Control Room Emergency Filtration System.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. By providing an improved protocol for charcoal testing 
the proposal provides greater assurance that the installed charcoal 
can perform its design function and, thus, the consequences of 
evaluated accidents remain valid. The method of laboratory analysis 
has no effect upon how the plant is operated, including the method 
of sample removal. Therefore, the probability [or consequences] of 
any evaluated accident is unchanged.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposal has no effect on the manner of plant 
operation. The proposal does not involve any change to the plant 
design. Therefore, the change creates no new accident modes.

[[Page 40015]]

    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety. By providing an improved protocol 
for charcoal testing the proposal acts to maintain existing safety 
margins. The change to the SGTS charcoal acceptance criteria also 
acts to ensure that the existing margins, as discussed in Regulatory 
Guide 1.52, Revision 2 [Design, Testing and Maintenance Criteria for 
Post-Accident Engineered Safety-Feature Atmosphere Cleanup System 
Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear 
Power Plants], are maintained.
    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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: Mark Reinhart, Acting

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

    Date of amendment request: June 18, 1996
    Description of amendment request: For Beaver Valley Power Station, 
Unit No. 1 (BVPS-1) only, the proposed amendment would revise Technical 
Specification (TS) 3.4.5 and associated Bases; the Bases for TS 3.4.6.2 
would also be revised. The proposed changes are editorial in nature and 
are intended to provide consistency between the TSs and associated 
Bases. Index page XIX would be revised to reflect the revision of page 
numbers for TS Tables 4.4-1 and 4.4-2 due to shifting of text.
    For Beaver Valley Power Station, Unit No. 2 (BVPS-2) only, the 
proposed amendment would implement a voltage-based repair criteria for 
steam generator tubes similar to the changes approved for BVPS-1 by 
License Amendment No. 198. The proposed changes are intended to reflect 
the guidance provided in NRC Generic Letter 95-05, ``Voltage-Based 
Repair Criteria for Westinghouse Steam Generator Tubes Affected by 
Outside Diameter Stress Corrosion Cracking.'' The proposed changes 
would revise TSs 3.4.5 and 3.4.6.2 and associated Bases. TS Table 4.4-2 
would be revised to reference TS 6.6 for reporting requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Tube burst criteria are inherently satisfied during normal 
operating conditions due to the proximity of the tube support plate 
(TSP). Test data indicates that tube burst cannot occur within the 
TSP, even for tubes which have 100% throughwall electric discharge 
machining notches, 0.75 inch long, provided that the TSP is adjacent 
to the notched area. Since tube-to-TSP proximity precludes tube 
burst during normal operating conditions, use of the criteria must 
retain tube integrity characteristics which maintain a margin of 
safety of 1.43 times the bounding faulted condition, main steamline 
break (MSLB) pressure differential. The Regulatory Guide (RG) 1.121 
criterion requiring maintenance of a safety factor of 1.43 times the 
MSLB pressure differential on tube burst is satisfied by 7/8'' 
diameter tubing with bobbin coil indications with signal amplitudes 
less than 8.6 volts, regardless of the indicated depth measurement.
    The upper voltage repair limit (VURL) will be determined 
prior to each outage using the most recently approved NRC database 
to determine the tube structural limit (VSL). The structural 
limit is reduced by allowances for nondestructive examination (NDE) 
uncertainty (VNDE) and growth (VGR) to establish 
VURL. Using the Generic Letter (GL) 95-05 NDE and growth 
allowances for an example, the NDE uncertainty component of 20% and 
a voltage growth allowance of 30% per full power year can be 
utilized to establish a VURL of 5.7 volts. The 20% NDE 
uncertainty represents a square-root-sum-of-the-squares (SRSS) 
combination of probe wear uncertainty and analyst variability. The 
degradation growth allowance should be an average growth rate or 30% 
per effective full power year, whichever is larger.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated MSLB 
outside of containment but upstream of the main steam isolation 
valve (MSIV) represents the most limiting radiological condition 
relative to the plugging criteria. In support of implementation of 
the revised plugging limit, analyses will be performed to determine 
whether the distribution of cracking indications at the tube support 
plate intersections during future cycles are projected to be such 
that primary-to-secondary leakage would result in postulated site 
boundary and control room doses exceeding 10 CFR 100, 10 CFR 50 
Appendix A, and GDC-19 [General Design Criterion-19] requirements, 
respectively. A separate calculation has determined the maximum 
allowable MSLB leakage limit in a faulted loop. This limit was 
calculated using the technical specification reactor coolant system 
(RCS) Iodine-131 activity level of 1.0 microcuries per gram dose 
equivalent Iodine-131 and the recommended Iodine-131 transient 
spiking values consistent with NUREG-0800. The projected MSLB 
leakage rate calculation methodology prescribed in Section 2.b of GL 
95-05 will be used to calculate the end-of-cycle (EOC) leakage. 
Projected EOC voltage distribution will be developed using the most 
recent EOC eddy current results and considering an appropriate 
voltage measurement uncertainty. The log-logistic probability of 
leakage correlation will be used to establish the MSLB leakrate used 
for comparison with the faulted loop allowable limit. Therefore, as 
implementation of the voltage-based repair criteria does not 
adversely affect steam generator tube integrity and implementation 
will be shown to result in acceptable dose consequences, the 
proposed amendment does not result in any increase in the 
probability or consequences of an accident previously evaluated in 
the Updated Final Safety Analysis Report (UFSAR).
    The proposed changes to the BVPS-1 Index, Specifications and 
associated Bases and the proposed change to BVPS-2 Table 4.4-2 are 
editorial in nature. Therefore, these changes do not involve an 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Implementation of the proposed steam generator tube voltage-
based repair criteria does not introduce any significant changes to 
the plant design basis. Use of the voltage-based repair criteria 
does not provide a mechanism which could result in an accident 
outside of the region of the tube support plate elevations as no 
outside diameter stress corrosion cracking (ODSCC) is occurring 
outside the thickness of the tube support plates. Neither a single 
or multiple tube rupture event would be expected in a steam 
generator in which the plugging limit has been applied (during all 
plant conditions).
    Duquesne Light Company will implement a maximum primary-to-
secondary leakage rate limit of 150 gpd [gallons per day] per steam 
generator to help preclude the potential for excessive leakage 
during all plant conditions. The RG 1.121 criterion for establishing 
operational leakage rate limits that require plant shutdown are 
based upon leak-before-break considerations to detect a free span 
crack before potential tube rupture during faulted plant conditions. 
The 150 gpd limit provides for leakage detection and plant shutdown 
in the event of the occurrence of an unexpected single crack 
resulting in leakage that is associated with the longest permissible 
crack length. RG 1.121 acceptance criteria for establishing 
operating leakage limits are based on leak-before-break 
considerations such that plant shutdown is initiated if the leakage 
associated with the longest permissible crack is exceeded.
    The single through-wall crack lengths that result in tube burst 
at 1.43 times the MSLB pressure differential and the MSLB pressure 
differential alone are approximately 0.57 inch and approximately 
0.84 inch, respectively. A leak rate of 150 gpd will provide for 
detection of approximately 0.41 inch long cracks at nominal leak 
rates and approximately 0.62 inch long cracks at the lower 95% 
confidence level leak rates. Since tube burst is precluded during 
normal

[[Page 40016]]

operation due to the proximity of the TSP to the tube and the 
potential exists for the crevice to become uncovered during MSLB 
conditions, the leakage from the maximum permissible crack must 
preclude tube burst at MSLB conditions. Thus, the 150 gpd limit 
provides for plant shutdown prior to reaching critical crack lengths 
for MSLB conditions using the lower 95% leakrate data. Additionally, 
this leak-before-break evaluation assumes that the entire crevice 
area is uncovered during blowdown. Partial uncovery will provide 
benefit to the burst capacity of the intersection. Analyses have 
shown that only a small percentage of the TSPs are deflected greater 
than the TSP thickness during a postulated MSLB.
    As steam generator tube integrity upon implementation of the 
voltage-based repair criteria continues to be maintained through 
inservice inspection and primary-to-secondary leakage monitoring, 
the possibility of a new or different kind of accident from any 
accident previously evaluated is not created.
    The proposed change to BVPS-1 Index, Specifications and 
associated Bases and the proposed change to BVPS-2 Table 4.4-2 are 
editorial in nature. These changes do not change the performance of 
plant systems, plant configuration or method of operating the plant.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The use of the voltage-based repair criteria at BVPS-2 maintains 
steam generator tube integrity commensurate with the criteria of RG 
1.121. This guide describes a method acceptable to the Commission 
for meeting GDCs 14, 15, 30, 31, and 32 by reducing the probability 
or the consequences of steam generator tube rupture. This is 
accomplished by determining the limiting conditions of degradation 
of steam generator tubing, as established by inservice inspection, 
for which tubes with unacceptable cracking should be repaired or 
removed from service. Upon implementation of the proposed criteria, 
even under the worst case conditions, the occurrence of ODSCC at the 
tube support plate elevations is not expected to lead to a steam 
generator tube rupture event during normal or faulted plant 
conditions. The EOC distribution of crack indications at the tube 
support plate elevations will be confirmed to result in acceptable 
primary-to-secondary leakage during all plant conditions and that 
radiological consequences remain within the licensing basis.
    In addressing the combined effects of loss-of-coolant-accident 
(LOCA) + safe shutdown earthquake (SSE) on the steam generator 
component (as required by GDC 2), it has been determined that tube 
collapse may occur in the steam generators at some plants. This is 
the case as the tube support plates may become deformed as a result 
of lateral loads at the wedge supports at the periphery of the plate 
due to the combined effects of the LOCA rarefaction wave and SSE 
loadings. Then, the resulting pressure differential on the deformed 
tubes may cause some of the tubes to collapse. There are two issues 
associated with steam generator tube collapse. First, the collapse 
of steam generator tubing reduces the RCS flow area through the 
tubes. The reduction in flow area increases the resistance to flow 
of steam from the core during a LOCA which, in turn, may potentially 
increase peak clad temperature. Second, there is a potential that 
partial through-wall cracks in tubes could progress to complete 
through-wall cracks during tube deformation or collapse.
    The results of an analysis using the larger break inputs show 
that the LOCA loads were found to be of insufficient magnitude to 
result in steam generator tube collapse or significant deformation. 
Since the leak-before-break methodology is applicable to the reactor 
coolant loop piping, the probability of breaks in the primary loop 
piping is sufficiently low that they need not be considered in the 
structural design of the plant. The limiting LOCA event becomes the 
pressurizer spray line break. Analysis results have demonstrated 
that no tubes were subject to deformation or collapse. No tubes have 
been excluded from application of the subject voltage-based steam 
generator tube repair criteria.
    Addressing RG 1.83 considerations, implementation of the 
voltage-based repair criteria is supplemented by: enhanced eddy 
current inspection guidelines to provide consistency in voltage 
normalization, the bobbin coil inspection will include 100% of the 
hot-leg TSP intersections and cold-leg intersections down to the 
lowest cold-leg TSP with known ODSCC, the determination of the TSPs 
having ODSCC will be based on the performance of at least 20% random 
sampling of tubes inspected over their full length, and rotating 
pancake coil inspection requirements for the larger indications left 
inservice to characterize the principal degradation as ODSCC.
    As noted previously, implementation of the tube support plate 
intersection voltage-based repair criteria will decrease the number 
of tubes which must be repaired. The installation of steam generator 
tube plugs reduces the RCS flow margin. Thus, implementation of the 
voltage-based repair criteria will maintain the margin of flow that 
would otherwise be reduced in the event of increased tube plugging.
    The proposed change to the BVPS-1 Index, Specifications and 
associated Bases and the proposed change to BVPS-2 Table 4.4-2 are 
editorial in nature. These changes will not reduce the margin of 
safety because they have no impact on any safety analysis 
assumptions.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the UFSAR or any 
BASES of the plant technical specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: May 9, 1996
    Description of amendment request: The proposed amendment changes 
both technical and administrative requirements associated with station 
batteries. The proposed changes are modeled after ``Standard Technical 
Specifications - Babcock and Wilcox Plants,'' NUREG-1430 and Nuclear 
Energy Institute guidance, ``IEEE Recommended Practice for Maintenance, 
Testing, and Replacement of Vented Lead-Acid Batteries for Stationary 
Applications,'' IEEE Std 450-1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The switchyard 125V DC control power source requirements do not 
meet the criteria for inclusion in Technical Specifications (TSs) as 
evaluated with respect to the selection criteria of 10 CFR 50-36. 
These control power sources are not assumed to mitigate accident or 
transient events. The effects of a loss of these control power 
sources are enveloped by the Loss of Offsite Power (LOOP) event and 
relocation is considered to have a non-significant impact on the 
probability or severity of a LOOP event. These requirements will be 
relocated from the TSs to an appropriate administratively controlled 
document and maintained pursuant to 10 CFR 50.59.
    Proposed changes incorporating the requirements of TS 3.7.1.D, 
3.7.2.E, 3.7.2.F, and 3.7.2.A, as related to the DC electrical power 
subsystems in the new TS 3.7.3 results in a more stringent 
requirement for the ANO-1 TSs in that reductions to lower conditions 
of operation in shorter periods of time are now required. These more 
stringent requirements are not assumed to be initiators of any 
analyzed events and will not alter assumptions relative to 
mitigation of accident or transient events.
    Proposed changes incorporating TS 3.7.4. requirements for the 
station batteries allowing the battery parameters to be outside

[[Page 40017]]

the limits of the Battery Inspection Program for 31 days do not 
result in an increase in the frequency of consequences of any 
analyzed accident, as the actions require more frequent checks of 
other parameters to ensure battery capability during this 31 day 
period. The Battery Inspection Program also requires evaluations to 
determine battery operability in the event these limits are 
exceeded. If an evaluation shows the battery is incapable of 
performing its design basis function, that DC electrical subsystem 
will be declared inoperable, and the appropriate actions taken.
    Proposed changes to allow the use of float current in lieu of 
specific gravity incorporate current industry guidance on 
operability measures for station batteries, as stated in IEEE-450, 
``IEEE Recommended Practice for Maintenance, Testing, and 
Replacement of Vented Lead-Acid Batteries for Stationary 
Applications.'' This Surveillance Requirement is not considered to 
initiate or mitigate any analyzed accident.
    The proposed incorporation of a Battery Inspection Program 
relocates maintenance requirements from the TSs to a program under 
10 CFR 50.59 control and allows the TSs to concentrate on those 
items required to ensure battery operability. These relocated 
requirements are not considered to be initiators of any analyzed 
accident. Battery operability is assured by the combination of TS 
Surveillance Requirements and Battery Inspection Program maintenance 
requirements based on IEEE-450 guidance.
    Proposed changes in Surveillance Requirements and Frequencies 
reflect current industry guidance on maintenance and testing of the 
station batteries. These requirements, in themselves, are not 
considered to be initiators of any analyzed accident condition. 
Although some frequencies have been extended, continued performance 
of maintenance activities in accordance with IEEE-450, in addition 
to the required Surveillance Requirements, ensures that corrective 
maintenance can be performed prior to a condition challenging an 
operability limit.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed changes do not change the design, configuration, or 
method of operation of the plant.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    Relocation of the switchyard 125V DC control power source 
requirements has no impact on any safety analysis assumptions. In 
addition, the requirements associated with these control power 
sources are relocated to an owner controlled document for which 
future changes will be evaluated pursuant to the requirements of 10 
CFR 50.59.
    Proposed changes incorporating the requirements of TS 3.7.1.D, 
3.7.2.E, 3.7.2.F, and 3.7.2.A, as related to the DC electrical power 
subsystems, in the new TS 3.7.3 impose more stringent requirements 
than previously specified for ANO-1.
    Proposed changes incorporating TS 3.7.4 requirements for the 
station batteries allowing the battery parameters to be outside the 
limits of the Battery Inspection Program for 31 days may involve an 
incremental reduction in the margin of safety since the battery may 
be in a slightly degraded state. However, this reduction is not 
considered significant in that the associated actions require more 
frequent checks of other parameters to ensure battery capability 
during this 31 day period. The attery Inspection Program also 
requires evaluations to determine battery operability in the event 
these limits are exceeded.
    If an evaluation shows the battery is incapable of performing 
its design basis function, that DC electrical subsystem will be 
declared inoperable, and the appropriate actions taken.
    The proposed change to allow the use of float current in lieu of 
specific gravity as a measure of battery operability is expected to 
result in a more representative measure of operability. IEEE-450 
states that specific gravity may not be an appropriate measure of 
battery capability following addition of electrolyte or when the 
battery is on recharge following a discharge.
    Proposed incorporation of a Battery Inspection Program relocates 
maintenance requirements from the TSs to a program under 10 CFR 
50.59 controls and allows the TSs to concentrate on those items 
required to ensure battery operability. The relocation of these 
requirements is not considered to be a reduction in the margin of 
safety. Battery operability is assured by the combination of TS 
Surveillance Requirements and Battery Inspection Program maintenance 
requirements based on IEEE-450 guidance.
    Proposed changes in Surveillance Requirements and Frequencies 
reflect current industry guidance on maintenance and testing of the 
station batteries. Although some frequencies have been extended, 
continued performance of maintenance activities in accordance with 
IEEE-450, in addition to the required Surveillance Requirements, 
ensures that corrective maintenance can be performed prior to a 
condition challenging an operability limit.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: June 27, 1996
    Description of amendment request: The proposed amendment will 
modify Technical Specification 3/4.3.3.6, ``Accident Monitoring 
Instrumentation,'' based on the Combustion Engineering improved 
Standard Technical Specifications (STS) issued by the NRC as NUREG 
1432. The amendment will also revise the Technical Specification (TS) 
to include Accident Monitoring Instrumentation as recommended by 
Regulatory Guide (RG) 1.97, Revision 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change deletes all non-Type A and non-Category 1 
instruments from the requirements of TS 3/4.3.3.6, ``Accident 
Monitoring Instrumentation.'' Type A variables provide the primary 
information required to permit the control room operators to take 
specific manually controlled actions, for which no automatic control 
is provided, that are required for safety systems to accomplish 
their safety functions during a DBA [Design Basis Accident]. 
Category 1, non-Type A variables are important in reducing public 
risk and are retained in TS because they are intended to assist 
operators in minimizing the consequences of accidents. Category 2 
instruments are generally designated for indicating system operating 
status and are not designated as essential key variables necessary 
for the safe shutdown of the plant. The proposed change preserves 
the safety requirements of RG 1.97, Revision 3, and will not 
adversely affect any material condition of the plant that could 
directly contribute to causing or mitigating the affects of an 
accident.
    The proposed change also adds two parameters to TS 3/4.3.3.6 
which were previously controlled administratively or per another TS. 
Containment Pressure (Wide Wide Range) is being added because it is 
a Category 1 parameter required in addition to Containment Pressure 
(Wide Range), which is currently in the TS. Neutron Flux is being 
added to distinguish the RG 1.97 channels from the non-RG 1.97 
channels and to provide action and surveillance requirements 
consistent with the other accident monitoring instrumentation. These 
additions to TS 3/4.3.3.6 contribute to the overall safety of the 
plant and therefore in no way increase the probability or 
consequences of an accident previously evaluated.
    Additionally, the proposed change also extends the AOTs [Allowed 
Outage Times] for TS 3/4.3.3.6 and replaces the HOT SHUTDOWN 
requirement for the number of OPERABLE channels being less than the 
Required Number of channels with a Special Report requirement. These 
changes are based

[[Page 40018]]

on the relatively low probability of an accident occurring which 
would require these instruments, the passive nature of these 
instruments, and alternate means of monitoring available. This is 
consistent with the CE improved STS and associated safety analyses 
which have been approved and issued by the NRC as NUREG 1432.
    The remainder of the proposed change provides enhancements and 
clarifications to TS 3/4.3.3.6 which have no potential to impact 
plant operations. No previous accident scenario is changed, and 
initiating conditions and assumptions remain as previously analyzed. 
Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed change will not alter the operation of the plant or 
the manner in which the plant is operated. No new or different 
failure modes have been introduced. TS 3/4.3.3.6 ensures the 
OPERABILITY of essential Post Accident Monitoring Instrumentation. 
This instrumen-tation provides information to the control room 
operators during an accident so that appropriate actions can be 
taken to mitigate the consequences of the accident. These 
instruments are passive in nature in that no critical automatic 
action is assumed to occur from these instruments. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change revises TS 3/4.3.3.6 based on the 
information provided in CE improved STS, NUREG 1432. The deletion 
and addition of specific components from the TS per this change is 
commensurate with the safety significance of their associated 
parameters. The proposed change ensures the operability of the post 
accident monitoring instrumentation which has been designated, by RG 
1.97 and Waterford 3's associated analysis, as essential for 
availability during and following a DBA. The proposed change 
preserves the single failure criteria required for this 
instrumentation and maintains the level of safety currently 
established in the Technical Specifications. The proposed change 
will not affect any physical protective boundary. Therefore, the 
proposed change will not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: 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: July 17, 1996 (TSCR 242, Rev. 2)
    Description of amendment request: The proposed change to the 
Technical Specifications would allow the implementation of 10 CFR Part 
50, Appendix J, Option B. This application supersedes the previously 
submitted application dated February 23, 1996, which was noticed in the 
Federal Register on March 27, 1996 (61 FR 13526).
    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 involves no 
significant hazards considerations as defined by NRC in 10 CFR 
50.92.
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or occurrence or the consequences of an accident of 
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 not 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 [Performance-Based Containment Leak Test Program] 
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 [primary containment integrated leak rate tests] 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 any 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 greater 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 surveillance interval cannot introduce 
any new type of accident or malfunction.
    The PCILRT is presently performed at 20 psig. Performance of the 
PCILRT at PGG5Ga(35 psig) will provide a more direct leak rate for 
analysis. Pa is the design pressure of the torus (the drywell 
design pressure is 44 psig, but the torus is non isolable from the 
drywell). Therefore, Pa 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

[[Page 40019]]

or components are altered as a result of this change. Therefore, no 
new accident or malfunction different from 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 from 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 La, 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 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 
(Pa). 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 La acceptance criteria. 
The La 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

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

    Date of amendment request: June 28, 1996
    Description of amendment request: This amendment would allow 
implementation of Option B to 10 CFR Part 50, Appendix J, which permits 
performance based determination of the frequency of containment leak 
rate testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:
    The proposed change has been evaluated against the standards in 
10 CFR 50.92 and determined not to involve a significant hazards 
consideration, in that the editorial changes do not change the 
meaning or intent of the technical specifications, and operation of 
the facility in accordance with the proposed amendment.
    1. Would not involve a significant increase in the probability 
of occurrence or the consequences of an accident previously 
evaluated, because the proposed changes are either purely 
administrative changes (involving format, wording, or reporting 
requirements) or changes in containment leakage test requirements 
(minor scope changes or increased intervals between containment 
leakage tests). None of these changes are related to conditions 
which cause accidents. The proposed changes do not involve a change 
to the plant design or operation.
    NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' 
contributed to the technical bases for Option B of 10 CFR 50 
Appendix J. NUREG-1493 contains a detailed evaluation of the 
expected leakage from containment and the associated consequences. 
The increased risk due to lengthening of the intervals between 
leakage tests was also evaluated and found to be acceptable. Using a 
statistical approach, NUREG-1493 determined the increase in the 
expected dose to the public from extending the testing frequency to 
be extremely small.
    2. Would not create the possibility of a new or different kind 
of accident from any accident previously evaluated, because the 
testing or reporting requirements associated with this change do not 
involve a physical alteration of the plant design or changes in the 
methods governing normal plant operation. No safety related 
equipment or safety related functions are altered as a result of 
this change. As a result, the proposed change does not affect any of 
the parameters or conditions that could contribute to initiation of 
any accidents.
    3. Would not involve a significant reduction in a margin of 
safety because the proposed changes are either purely administrative 
(involving format, wording, or reporting requirements) or changes in 
containment leakage test requirements (minor scope changes or 
increased intervals between containment leakage tests) such that the 
allowable containment leakage rates presently specified in the 
Technical Specifications remain unchanged. The Technical 
Specifications and the Reactor Building Leakage Rate Testing Program 
will ensure that containment system testing is performed in full 
compliance with 10 CFR 50 Appendix J.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:
    Law/Government Publications Section, State Library of 
Pennsylvania, (REGIONAL DEPOSITORY) Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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, 8 (two letters), and 23, June 27, and July 9, 1996.
    Description of amendment request: The proposed amendment would 
allow extension of the standby diesel generator allowed outage time to 
14 days, and extension of the essential cooling water loop and the 
essential chilled water loop allowed outage times to 7 days. The 
proposed change would also add to Administrative Controls a description 
of the Configuration Risk Management Program (CRMP) used to assess 
changes in core damage probability resulting from applicable plant 
configurations. This application was previously published in the 
Federal

[[Page 40020]]

Register on February 8, 1996, (61 FR 4805).
    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 Standby Diesel Generators are not accident initiators, 
therefore the increase in Allowed Outage Times for this system does 
not increase the probability of an accident previously evaluated. 
The three train design of the South Texas Project ensures that even 
during the seven days the Essential Cooling Water loop or the 
Essential Chilled Water loop is inoperable there are still two 
complete trains available to mitigate the consequences of any 
accident. If the Essential Cooling Water and the Essential Chilled 
Water loops are operable during the 14 days the Standby Diesel 
Generator is inoperable, the Engineered Safety Features bus and 
equipment in the train associated with the inoperable Standby Diesel 
Generator will be operable. This ensures that all three redundant 
safety trains of the South Texas Project design are operable. In 
addition the Emergency Transformer will be available to supply the 
Engineered Safety Features bus normally supplied by the inoperable 
Standby Diesel Generator. These actions will ensure that the changes 
do not involve a significant increase in the consequences of 
previously evaluated accidents.
    The addition of the Configuration Risk Management Program to the 
Administrative Section of the Technical Specifications does not 
affect current accident analyses.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes affect only the magnitude of the Standby 
Diesel Generator, Essential Cooling Water and the Essential Chilled 
Water Allowed Outage Times as identified by the marked-up Technical 
Specification. As indicated above, the proposed change does not 
involve the alteration of any equipment nor does it allow modes of 
operation beyond those currently allowed. Therefore, implementation 
of these proposed changes does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes result in no significant increase in core 
damage or large early release frequencies. Three sets of PSA 
[probabilistic safety assessment] results have been presented to the 
NRC for the South Texas Project. One submitted in 1989 from the 
initial Level 1 PSA of internal and external events with a mean 
annual average CDF [core damage frequency] estimate of 1.7E-4, a 
second one submitted in 1992 to meet the IPE [individual plant 
examination] requirements from the Level 2 PSA/IPE with a CDF 
estimate of 4.4E-5, and an update of the PSA that was reported in 
the August 1993 Technical Specifications submittal with a variety of 
CDF estimates for different assumptions regarding the rolling 
maintenance profile and different combinations of modified Technical 
Specifications. The South Texas Project PSA was updated in March of 
1995 to include the NRC approved Risk-Based AOTs [allowed outage 
times] and STIs [surveillance test intervals], Plant Specific Data 
and incorporate the Emergency Transformer into the model. This 
update resulted in a CDF estimate of 2.07E-5 per reactor year. When 
the requested changes are modeled, the resulting CDF estimate is 
2.18E 10-5 (sic) [2.18E-5] per reactor year. This corresponds to 
5.2% decrease in the Core Damage Frequency calculated for the 
previously submitted 21 Day AOT. The Large, Early Release Frequency 
is quantified as 4.69E-07 per reactor year which represents a 
decrease of 7.5% from the value calculated for the previously 
submitted 21 Day AOT. Therefore, it is concluded that there is no 
significant reduction in the margin of safety.
    Based on the above evaluation, the South Texas Project has 
concluded that these changes do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: William D. Beckner

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

    Date of amendment request: July 5, 1996
    Description of amendment request: The proposed Technical 
Specification (TS) amendment would support implementation of Noble 
Metal Chemical Addition (NMCA) at the Duane Arnold Energy Center (DAEC) 
as a method to enhance the effectiveness of Hydrogen Water Chemistry 
(HWC) in mitigating Intergranular Stress Corrosion Cracking (IGSCC) in 
Boiling Water Reactor (BWR) vessel internal components. The proposed 
amendment would raise the reactor water conductivity limit in STARTUP 
and HOT SHUTDOWN only during the application of NMCA. The reactor water 
conductivity will be restored after the NMCA.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS amendment will not significantly increase the 
probability or consequences of any previously evaluated accidents.
    It is expected that during the NMCA application period, the 
reactor water conductivity will increase and exceed the conductivity 
limit of 2.0 [micro]mhos/cm specified in our current TS. Our current 
TS requires that whenever the reactor is in STARTUP or HOT SHUTDOWN 
Mode, the conductivity shall not exceed 2.0 [micro]mhos/cm for more 
than 48 continuous hours or be in HOT SHUTDOWN within the next 12 
hours and in COLD SHUTDOWN within the following 24 hours.
    The expected increase in conductivity is due to the presence of 
noble metal chemistry in the reactor water and is appropriate during 
the [NMCA] application period. The deposited layer of noble metals 
is beneficial for mitigating IGSCC in reactor vessel internal 
components. Other reactor water chemistry parameters such as 
chloride and sulfate are not expected to change; pH is expected to 
change but not out of the acceptable range. The reactor water 
chemistry parameters will be analyzed to ensure they are within the 
normal range, on a frequency consistent with the existing TS, 
Sections 4.6.B.2.c and 4.6.B.2.d when conductivity is elevated 
during the NMCA application.
    During and after the application, the Reactor Water Cleanup 
(RWCU) system will continue to operate to remove the excess ions 
from the reactor water and restore the reactor water conductivity to 
the limit specified in Section 3.6.B. Therefore, this proposed TS 
amendment will not significantly increase the probability or 
consequences of any previously evaluated accidents.
    2. The proposed TS amendment will not create the possibility of 
a new or different kind of accident. The proposed TS amendment will 
only permit a higher value of the reactor water conductivity limit 
during the application period of NMCA. The application is 
anticipated to increase the reactor water conductivity.
    During and after the application, the RWCU system will continue 
to operate to remove the excess ions and restore the reactor water 
conductivity to the limit specified in Section 3.6.B. As is 
discussed above, the deposited layer of noble metals is beneficial 
for mitigating IGSCC in reactor vessel internal components. 
Therefore, this proposed TS amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS amendment will only permit a higher value of 
the reactor water conductivity limit during the application period 
of NMCA. The increase in

[[Page 40021]]

conductivity is anticipated during the application and is 
appropriate. The deposited layer of noble metals is beneficial for 
mitigating IGSCC in reactor vessel internal components. During and 
after the application, the RWCU system will continue to operate to 
remove the excess ions and restore the reactor water conductivity to 
the limit specified in Section 3.6.B. Therefore, no margin of safety 
is reduced as a result of the anticipated increase in conductivity 
due to the addition of the known noble metals.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, 
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: June 21, 1996
    Description of amendment request: The proposed amendment would 
modify Section 5.7, ``High Radiation Areas,'' of the ``Administrative 
Controls'' section of the Clinton Power Station technical 
specifications (TS). The proposed changes include: (1) allowing 
utilization of a Radiation Work Permit (RWP) ``or equivalent'' to 
control entry into a high radiation area; (2) clarifying the example 
given in the TS of individuals who are qualified in radiation 
protection procedures; (3) clarifying the requirements for when 
specified access controls and barriers for high radiation areas within 
large areas like the containment must be established; (4) clarifying 
that it is acceptable for an RWP to specify a maximum dose, i.e., a 
specified setpoint on an alarming dosimeter in lieu of a stay time for 
entry into a high radiation area (where an individual could receive a 
deep dose equivalent of 3000 mrem in one hour); (5) eliminating the 
upper dose limit for specifying the applicability of the requirements 
of Specification 5.7.1; (6) providing additional flexibility regarding 
who may control the keys to locked doors for preventing unauthorized 
entry into high radiation areas; (7) reorganizing TS Sections 5.7.1, 
5.7.2, and 5.7.3 into four sections (5.7.1, 5.7.2, 5.7.3 and 5.7.4); 
and (8) making minor edits to enhance readability.
    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) None of the proposed changes involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    The proposed changes do not change the design or the operation 
of the plant. The proposed changes are only related to the control 
of access to high radiation areas for the purpose of controlling 
dose to plant personnel. Because no change to plant design is 
proposed, there is no impact to any accident mitigating system. 
Likewise, because there is no proposed change to plant operating 
procedures, plant operation is not impacted. This proposed change 
does not impact any accident scenario or the previously calculated 
post-accident doses. Therefore, the limits of 10 CFR 100 will 
continue to be met. No probability or consequence of any accident 
previously evaluated is impacted by the proposed changes to TS.
    (2) None of the proposed changes create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment is administrative in nature and does not 
impact directly or indirectly the design or the operation of the 
Clinton Power Station, thus no new accident can be created.
    (3) None of the proposed changes involve a significant reduction 
in a margin of safety.
    There is no reduction to the margin of safety because the 
operating limits and functional capabilities of plant safety systems 
are unaffected by the proposed changes to administrative 
requirements. As noted previously, the proposed changes do not 
impact any accident analyses, including the associated dose 
calculations. With respect to controls for controlling operational 
dose to plant personnel, the proposed changes are intended to 
provide clarity and/or flexibility with respect to the 
administration and programmatic controls for controlling such dose, 
and yet maintain an adequate margin of safety for minimizing dose to 
site personnel consistent with the requirements of 10 CFR 20 and 
guidance of Regulatory Guide 8.38.
    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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: June 28, 1996
    Description of amendment request: The proposed amendment would 
allow removal of the Inclined Fuel Transfer System (IFTS) primary 
containment blind flange while primary containment is required to be 
operable. This will provide flexibility to operate the IFTS for the 
purpose of testing and exercising the system during such conditions. 
Primary containment integrity will be provided by an alternate means 
while the blind flange is removed. The change would be incorporated via 
a provisional note into Technical Specification (TS) Surveillance 
Requirement 3.6.1.3.3, associated with TS 3.6.1.3, ``Primary 
Containment Isolation Valves (PCIVs).''
    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 allows operation of the IFTS while 
primary containment operability is required. The proposed change 
does not involve any modifications to plant systems or design 
parameters or conditions that contribute to the initiation of any 
accidents previously evaluated. Therefore, the proposed change 
cannot increase the probability of any accident previously 
evaluated.
    The proposed change potentially affects the leak-tight integrity 
of the containment structure which is designed to mitigate the 
consequences of a loss-of-coolant accident (LOCA). The function of 
the primary containment is to maintain functional integrity during 
and following the peak transient pressures and temperatures that 
result from any LOCA. The primary containment is designed to limit 
fission product leakage following the design basis LOCA. Because the 
proposed change does not alter the plant design, only the extent of 
the boundaries that provide primary containment isolation for the 
IFTS penetration, the proposed change does not result in an increase 
in primary containment leakage. However, temporarily using the IFTS 
transfer tube and its attached appurtenances as part of the primary 
containment boundary (which have not been fabricated or installed to 
exactly the same requirements as a fully certified primary 
containment penetration) can increase the probability that a LOCA 
would challenge the pressure retaining integrity of these 
components. Since the subject components have been built to 
withstand pressure, temperature, and seismic conditions similar to 
those of the existing penetration, they are judged to be an

[[Page 40022]]

acceptable barrier to prevent the uncontrolled release of post-
accident fission products for the purposes of this amendment 
request.
    Further, it has been shown that the largest potential leakage 
pathway, the IFTS transfer tube itself, would remain sealed by the 
depth of water required to be maintained in the fuel building fuel 
transfer pool. The transfer tube drain line constitutes the other 
possible leakage pathway, and will be required to be capable of 
being isolated via administrative control of the manual isolation 
valve in the drain line. Additionally, due to the physical 
relationships of the buildings and components involved, any leakage 
from either of these pathways is fully contained within the 
boundaries of the secondary containment and would be filtered by the 
Standby Gas Treatment System prior to release to the environment.
    Based on the above, Illinois Power has concluded that the 
proposed change will not result in a significant increase in the 
probability or consequences of any accident previously evaluated.
    (2) The proposed change does not involve a change to the plant 
design or operation (except when the IFTS is operated). As a result, 
the proposed change does not affect any of the parameters or 
conditions that could contribute to the initiation of any accidents. 
No new accident modes are created by this change. Extending the 
primary containment boundary to include portions of the IFTS has no 
influence on, nor does it contribute to the possibility of a new or 
different kind of accident or malfunction from those previously 
analyzed.
    Based on the above, Illinois Power has concluded that the 
proposed change will not create the possibility of a new or 
different kind of accident not previously evaluated.
    (3) The request does not involve a significant reduction in a 
margin of safety. The proposed change only affects the extent of a 
portion of the primary containment boundary. Precautions will be 
taken to administratively control the IFTS transfer tube drain path 
so that the proposed change will not increase the probability that 
an increase in leakage from the primary containment to the secondary 
containment could occur.
    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 La, which is defined by the Clinton Power Station 
Technical Specifications to be 0.65% of primary containment air 
weight per day at the calculated peak constant pressure (Pa). 
The limitation on containment leakage rate is designed to ensure 
that total leakage volume will not exceed the value assumed in the 
accident analyses at the peak accident pressure (Pa). 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 La acceptance criteria. The 
La value is not being modified by this proposed technical 
specification change. The IFTS will continue to provide an 
acceptable barrier to prevent containment leakage during a LOCA, and 
therefore this change will not create a situation causing the 
containment leakage rate acceptance criteria to be violated.
    As a result, Illinois Power has concluded that the proposed 
change will not result in a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525
    NRC Project Director: Gail H. Marcus

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: June 11, 1996 (AEP:NRC:80027)
    Description of amendment requests: The proposed amendments would 
remove from the technical specifications (TS) certain requirements for 
administrative controls, related to quality assurance requirements, in 
accordance with the guidance of NRC Administrative Letter 95-
06,Relocation of Technical Specifications Administrative 
Controls Related to Quality Assurance.''
    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:
    We have evaluated the proposed T/S changes and have determined 
that the changes should involve no significant hazards consideration 
based on the criteria established in 10 CFR 50.92(c). Operation of 
Cook Nuclear Plant in accordance with the proposed amendment will 
not satisfy any of the following criteria:
    (a) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not involve any physical alteration of 
plant configurations, changes to setpoints, or operating parameters. 
This proposed amendment is to relocate the T/S requirements for 
administrative controls that are related to quality assurance to the 
QAPD [Quality Assurance Program Description]. This is in accordance 
with the guidance provided in AL 95-06. Also, the relocated 
requirements and future changes are controlled by 10 CFR 50.54(a) 
which requires prior NRC approval for changes that reduce the 
commitments in the program description previously accepted by the 
NRC. Therefore, there will be no significant increase in the 
probability or consequences of an accident previously evaluated.
    (b) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    The proposed change does not involve any physical alteration of 
plant configurations, changes to setponts, or operating parameters. 
This proposed amendment is to relocate the T/S requirements for 
administrative controls that are related to quality assurance to the 
QAPD. This is in accordance with the guidance provided in AL 95-06. 
Also, the relocated requirements and future changes are controlled 
by 10 CFR 50.54(a) which requires prior NRC approval for changes 
that reduce the commitments in the program description previously 
accepted by the NRC. Therefore, this proposed change does not create 
the possibility of a new of different kind of accident from any 
previously analyzed.
    (c) Involve a significant reduction in a margin of safety.
    The proposed change does not involve any physical alteration of 
plant configurations, changes to setpoints, or operating parameters. 
This proposed amendment is to relocate the T/S requirements for 
administrative controls that are related to quality assurance to the 
QAPD. This is in accordance with the guidance provided in AL 95-06. 
Also, the relocated requirements and future changes are controlled 
by 10 CFR 50.54(a), which requires prior NRC approval for changes 
that reduce the commitments in the program description previously 
accepted by the NRC. Therefore, this 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 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: Mark Reinhart, Acting

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit No. 1, Berrien County, Michigan

    Date of amendment request: June 19, 1996 [AEP:NRC:1166AA]
    Description of amendment request: The proposed amendment would 
modify the technical specifications (T/

[[Page 40023]]

S) to allow continued use of the 2-volt steam generator (SG) tube 
plugging criteria for future operating cycles as discussed in NRC 
Generic Letter 95-05, ``Voltage-Based Repair Criteria for the Repair of 
Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress 
Corrosion Cracking.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    In accordance with the three factor test of 10 CFR 50.92(c), 
implementation of the proposed license amendment is analyzed using 
the following standards and found not to: 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 margin of safety. 
Conformance of the proposed amendment to the standards for a 
determination of no significant hazards as defined in 10 CFR 50.92 
(three factor test) is shown in the following paragraphs:
    1) Operation of Cook Nuclear Plant Unit 1, in accordance with 
the proposed license amendment, does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Testing of model boiler specimens for free 
span tubing
    (no TSP [tube support plate] restraint) at room temperature 
conditions show burst pressures in excess of 5000 psi for indications 
of outer diameter stress corrosion cracking [ODSCC] with voltage 
measurements as high as 19 volts. Burst testing performed on pulled 
tubes from Cook Nuclear Plant Unit 1 with up to a 2.02 volt indication 
shows measured burst pressure in excess of 10,000 psi at room 
temperature. Burst testing performed on pulled tubes from other plants 
show burst pressures in excess of 5,300 psi at room temperatures. 
Correcting for the effects of temperature on material properties and 
minimum strength levels (as the burst testing was done at room 
temperature), tube burst resistance significantly exceeds the safety 
factor requirements of RG [Regulatory Guide] 1.121 [Bases for Plugging 
Degraded PWR Steam Generatory Tubes]. As stated earlier, tube burst 
criteria are inherently satisfied during normal operating conditions 
due to the proximity of the TSP. Test data indicates that tube burst 
cannot occur within the TSP, even for tubes which have 100% throughwall 
electric-discharge machined notches 0.75 inch long, provided the TSP is 
adjacent to the notched area. Since tube-to-tube support plate 
proximity precludes tube burst during normal operating conditions, it 
follows that use of the proposed plugging criteria must, therefore, 
retain tube integrity characteristics which maintain the RG 1.121 
margin of safety of 1.43 times the bounding faulted condition (steam 
line break) pressure differential.
    During a postulated main SLB [steamline break], the TSP has the 
potential to deflect during blowdown, thereby uncovering the 
intersection. Based on the existing data base, the RG 1.121 
criterion requiring maintenance of a safety factor of 1.43 times the 
SLB pressure differential on tube burst is satisfied by 7/8 inch 
diameter tubing with bobbin coil indications with signal amplitudes 
less than VSL, regardless of the indicated depth measurement. A 
2 volt plugging criteria compares favorably with the current 
VSL (8.8 volt) structural limit, considering the previously 
calculated growth rates for ODSCC within Cook Nuclear Plant Unit 1 
SGs. Considering a voltage growth component of 0.8 volts (40% 
voltage growth based on 2 volts BOC [beginning of cycle] and a 
nondestructive examination uncertainty of 0.40 volts (20% voltage 
uncertainty based on 2 volts BOC), when added to the BOC plugging 
criteria of 2 volts, results in a bounding EOC [end of cycle] 
voltage of approximately 3.2 volts for a cycle operation. A 5.6 volt 
safety margin exists (8.8 - 3.2 volt EOC = 5.6 volt margin).
    For the voltage/burst correlation, the EOC structural limit is 
supported by a voltage of 8.8 volts. Using this VSL of 8.8 
volts, a BOC maximum allowable repair limit can be established using 
the guidance of RG 1.121. The BOC maximum allowable repair limit 
should not permit a significant number of EOC indications to exceed 
the VSL and should assure that acceptable tube burst 
probabilities are attained. By adding NDE [nondestructive 
examination] uncertainty allowances and an allowance for crack 
growth to the repair limit, the structural limit can be validated. 
The previous plugging criteria submittal established the 
conservative NDE uncertainty limit (VNDE) of 20% of the BOC 
repair limit. For consistency, a 40% voltage growth allowance 
(VGR) to the BOC repair limit is also included. This allowance 
is extremely conservative for Cook Nuclear Plant Unit 1. Therefore, 
the maximum allowable upper voltage repair limit VURL for BOC, 
based on the VSL of 8.8 volts, can be represented by the 
expression:
    VURL + (VNDE x VURL) + (VGR x VURL) = 
8.8 volts, or,
    the maximum allowable BOC repair limit can be expressed 
as,VURL = 8.8 volt structural limit/1.6 = 5.5 volts.
    This structural repair limit supports this application for 
plugging criteria implementation to repair bobbin indications 
greater than 2 volts based on RPC [rotating pancake coil] 
confirmation of the indication. Conservatively, an upper limit of 
5.5 volts will be used to repair bobbin coil indications which are 
above 2 volts but do not have confirming RPC calls.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated main 
SLB outside of containment, but upstream of the main steam isolation 
valve, represents the most limiting radiological condition relative 
to the plugging criteria. In support of implementation of the 
plugging criteria, it will be determined whether the distribution of 
crack indications at the TSP intersections at the EOC are projected 
to be such that primary-to-secondary leakage would result in site 
boundary doses within a small fraction of the 10 CFR 100 guidelines. 
A separate calculation has determined this allowable SLB leakage 
limit to be 8.4 gpm. Although not required by the Cook Nuclear Plant 
design basis, this calculation uses the recommended Iodine-131 
transient spiking values consistent with NUREG-0800 [Standard Review 
Plan], and the T/S reactor coolant system activity limit of 1 micro 
curie per gram dose equivalent Iodine-131. Control room dose 
calculations were also performed and found to be less limiting than 
the offsite dose leakrate. Therefore, the more conservative offsite 
dose leakrate is used. The projected SLB leakage rate calculation 
methodology prescribed in GL 95-05 and WCAP 14277 [Steam Line Break 
Leak Rate and Tube Burst Probability Analysis Methods for Outside 
Diameter Stress Corrosion Cracking at Tube Support Plate 
Intersections] will be used to calculate EOC leakage, based on 
actual EOC distributions and EOC projected distributions. Due to the 
relatively low voltage growth rates at Cook Nuclear Plant Unit 1 and 
the relatively small number of indications affected by the plugging 
criteria, SLB leakage prediction per GL 95-05 is expected to be 
significantly less than the permissible level of 8.4 gpm in the 
faulted loop.
    The inclusion of all intersections in the leakage model, along 
with application of a probability of detection of 0.6, will result 
in extremely conservative leakage estimations. Close examination of 
the available data shows that indications of less than 2.8 volts 
will not be expected to leak during SLB conditions.
    The proposed amendment does not result in any increase in the 
probability or consequences of an accident previously evaluated 
within the cook Nuclear Plant Unit 1 Final Safety Analysis Report 
(FSAR).
    2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the proposed SG tube plugging criteria does 
not introduce any significant changes to the plant design basis. Use 
of the criteria does not provide a mechanism which could result in 
an accident outside of the region of the TSP elevations. Neither a 
single nor a multiple tube rupture event would, under any plant 
conditions, be expected in a SG in which the plugging criteria has 
been applied. Specifically, we will continue to implement a maximum 
leakage rate limit of 150 gpd (0.1 gpm) per SG to help preclude the 
potential for excessive leakage during all plant conditions. The T/S 
limits imposed on primary-to-secondary leakage at operating 
conditions are a maximum of 0.4 gpm (600 gpd) for all SGs with a 
maximum of 150 gpd allowed for any one SG.
    The RG 1.121 criteria for establishing operational leakage rate 
limits that require

[[Page 40024]]

plant shutdown are based upon leak-before-break (LBB) considerations 
to detect a free span crack before potential tube rupture during 
faulted plant conditions. The 150 gpd limit should provide for 
leakage detection and plant shutdown in the event of the occurrence 
of an unexpected single crack resulting in leakage that is 
associated with the longest permissible crack length. Regulatory 
Guide 1.121 acceptance criteria for establishing operating leakage 
limits are based on LBB considerations such that plant shutdown is 
initiated if the leakage associated with the longest permissible 
crack is exceeded. The longest permissible crack is the length that 
provides a factor of safety of 1.43 against bursting at faulted 
conditions maximum pressure differential. A voltage amplitude of 8.8 
volts for typical ODSCC corresponds to meeting this tube burst 
requirement at a lower 95% prediction limit on the burst correlation 
coupled with 95/95 lower tolerance limit material properties. 
Alternate crack morphologies can correspond to 8.8 volts so that a 
unique crack length is not defined by the burst pressure versus 
voltage correlation. Consequently, typical burst pressure versus 
through-wall crack length correlations were used to define the 
``longest permissible crack'' for evaluating operating leakage 
limits. Consistent with the cycle 13, 14 and 15 license amendment 
requests for plugging criteria, and Section 5 of Enclosure 1 of the 
GL, operational leakage limits will remain at 150 gpd per SG. Axial 
cracks leaking at this level are expected to provide LBB protection 
at both the SLB pressure differential of 2560 psi and, while not 
part of any established LBB methodology, LBB protection will also be 
provided at a value of 1.43 times the SLB pressure differential. 
Thus, the 150 gpd limit provides for plant shutdown prior to 
reaching critical crack lengths for SLB conditions. Additionally, 
this LBB evaluation assumes that the entire crevice area is 
uncovered during blowdown. Partial uncovery will provide benefit to 
the burst capacity of the intersection.
    3) The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The use of the voltage-based bobbin probe interim TSP elevation 
plugging criteria at Cook Nuclear Plant Unit 1 is demonstrated to 
maintain SG tube integrity commensurate with the criteria of RG 
1.121. Regulatory Guide 1.121 describes a method acceptable to the 
NRC staff for meeting GDC [General Design Criteria] 14, 15, 31, and 
32 by reducing the probability or the consequences of SG tube 
rupture. This is accomplished by determining the limiting conditions 
of degradation of SG tubing, as established by in-service 
inspection, for which tubes with unacceptable cracking should be 
removed from service. Upon implementation of the criteria, even 
under the worst case conditions, the occurrence of ODSCC at the TSP 
elevations is not expected to lead to a SG tube rupture event during 
normal or faulted plant conditions. It will be confirmed by analysis 
and calculation that EOC distribution of crack indications at the 
TSP elevations will result in acceptable primary-to-secondary 
leakage during all plant conditions and that radiological 
consequences are not adversely impacted.
    In addressing the combined effects of a LOCA [loss-of-coolant 
accident] and SSE [safe-shutdown earthquake] on the SG component (as 
required by GDC 2), it has been determined that tube collapse may 
occur in the SGs at some plants. The postulated tube collapse 
results from a deformation of TSPs as a result of lateral loads at 
the wedge supports at the periphery of the plate. The lateral loads 
result from the combined effects of the LOCA rarefaction wave and 
SSE loadings. The resulting pressure differential on the deformed 
tubes may then cause some of the tubes to collapse.
    There are two issues associated with a postulated SG tube 
collapse. First, the collapse of SG tubing reduces the RCS [reactor 
coolant system] flow area through the tubes. The reduction in flow 
area increases the resistance to flow of steam from the core during 
a LOCA which, in turn, may potentially increase peak clad 
temperature. Second, there is a potential that partial through-wall 
cracks in tubes could progress to through-wall cracks during tube 
deformation or collapse.
    Consequently, since the LBB methodology is applicable to the 
Cook Nuclear Plant Unit 1 reactor coolant loop piping, the 
probability of breaks in the primary loop piping is sufficiently low 
that they need not be considered in the structural design of the 
plant. The limiting LOCA event becomes either the accumulator line 
break or the pressurizer surge line break. Loss of coolant accident 
loads for the primary pipe breaks were used to bound the Cook 
Nuclear Plant Unit 1 smaller breaks. The results of the analysis 
using the larger break inputs show that the LOCA loads were found to 
be of insufficient magnitude to result in SG tube collapse or 
significant deformation.
    Addressing RG 1.83 [In-Service Inspection of PWR Steam Generator 
Tubes] considerations, implementation of the bobbin coil probe, 
voltage-based interim tube plugging criteria of 2 volts is 
supplemented by enhanced eddy current inspection guidelines to 
provide consistency in voltage normalization, a 100% eddy current 
inspection sample size at the TSP elevation per T/S, and MRPC 
[motorized RPC] inspection requirements for the larger indications 
left in-service to characterize the principal degradation as ODSCC.
    As noted previously, implementation of the TSP elevation 
plugging criteria will decrease the number of tubes which must be 
repaired. The installation of SG tube plugs reduces the RCS flow 
margin. Thus, implementation of the plugging criteria will maintain 
the margin of flow that would otherwise be reduced in the event of 
increased tube plugging.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the FSAR or any 
Bases of the plant T/Ss.
    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: Mark Reinhart, Acting

Northern States Power Company, Docket No. 50-282, Prairie Island 
Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota

    Date of amendment request: July 15, 1996
    Description of amendment request: The proposed amendment would 
allow the use of the moveable incore detector system for measurement of 
the core peaking factors with less than 75% and greater than or equal 
to 50% of the detector thimbles available. The amendment request is a 
one-time only change for Prairie Island, Unit 1, Operating Cycle 18. It 
is being submitted to allow for continued operation if the number of 
detector thimbles drops below 75%.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes do not involve an increase in the 
probability of an accident previously evaluated. The moveable incore 
detector system is used only to provide confirmatory information on 
the neutron flux distribution and is not required for the daily safe 
operation of the core. The system is not a process variable that is 
an initial condition in the accident analyses. The only accident 
that the moveable incore detector system could be involved in is the 
breaching of the detector thimbles which would be enveloped by the 
small break loss of coolant accident (LOCA) analysis. As the 
proposed changes do not involve any changes to the system's 
equipment and no equipment is operated in a new or more harmful 
manner, there is no increase in the probability of such an accident.
    The proposed amendments would not involve an increase in the 
consequences of an accident previously evaluated. The moveable 
incore detector system provides a monitoring function that is not 
used for accident mitigation (the system is not used in the primary 
success path for mitigation of a design basis accident). The ability 
of the reactor protection system or engineered

[[Page 40025]]

safety features system instrumentation to mitigate the consequences 
of an accident will not be impaired by the proposed changes. The 
small break LOCA analysis (and thus its consequences) continues to 
bound potential breaching of the system's detector thimbles.
    With greater than or equal to 50% and less than 75% of the 
detector thimbles available, core peaking factor measurement 
uncertainties will be increased, which could impact the core peaking 
factors and as a result could affect the consequences of certain 
accidents. However, any changes in the core peaking factors 
resulting from increased measurement uncertainties will be 
compensated for by conservative measurement uncertainty adjustments 
in the Technical Specifications to ensure that pertinent core design 
parameters are maintained. Sufficient additional penalty is added to 
the power distribution measurements such that this change will not 
impact the consequences of any accident previously evaluated.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed amendments would not create the possibility of a 
new or different kind of accident previously evaluated as they only 
affect the minimum complement of equipment necessary for operability 
of the moveable incore detector system. There is no change in plant 
configuration, equipment or equipment design. No equipment is 
operated in a new manner. Thus the changes will not create any new 
or different accident causal mechanisms. The accident analysis in 
the Updated Safety Analysis Report remains bounding.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes will not create the possibility of a new or 
different kind of accident.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes will not involve a significant reduction in 
a margin of safety. The reduction in the minimum complement of 
equipment necessary for the operability of the moveable incore 
detector system could only impact the monitoring/calibration 
functions of the system. Reduction of the number of available 
moveable incore detector thimbles to the 50% level does not 
significantly degrade the ability of the system to measure core 
power distributions. With greater than or equal to 50% and less than 
75% of the detector thimbles available, core peaking factor 
measurement uncertainties will be increased, but will be compensated 
for by conservative measurement uncertainty adjustments in the 
Technical Specifications to ensure that pertinent core design 
parameters are maintained. Sufficient additional penalty is added to 
the power distribution measurements such that this change does not 
impact the safety margins which currently exist. Also, the reduction 
of available detector thimbles has negligible impact on the quadrant 
power tilt and core average axial power shape measurements. 
Sufficient detector thimbles will be available to ensure that no 
quadrant will be unmonitored.
    Based on these factors, the proposed changes in this license 
amendment will not result in a significant reduction in the plant's 
margin of safety, as the core will continue to be adequately 
monitored.
    Based on the evaluation above, and pursuant to 10 CFR 50, 
Section 50.91, Northern States Power Company has determined that 
operation of the Prairie Island Nuclear Generating Plant in 
accordance with the proposed license amendment request does not 
involve any significant hazards considerations as defined by NRC 
regulations in 10 CFR 50, Section 50.92.
    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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Mark Reinhart, Acting

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska

    Date of amendment request: May 31, 1996
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to add a Limiting Condition 
for Operation (LCO) for trisodium phosphate (TSP) and increase the 
minimum required amount of TSP contained in the containment sump mesh 
baskets.
    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.
    Trisodium Phosphate Dodecahydrate (TSP) is stored in the 
containment sump to raise the pH of the sump and spray water 
following a loss of coolant accident (LOCA). As the pH of the water 
increases, more radioactive iodine is kept in solution and the 
possibility of airborne radioactivity leakage is decreased. An 
additional advantage of a higher pH is the beneficial reduction in 
chloride stress corrosion cracking (SCC) of austenitic stainless 
steel components in the containment following a LOCA.
    This chemical is an accident mitigator, not an accident 
initiator in that it is not used until after an accident (i.e., a 
LOCA) has occurred. At the time it begins to go into solution, the 
accident has occurred, containment spray has been activated and 
water is collecting in the containment sump. Therefore, increasing 
the Technical Specification (TS) minimum amount of TSP verified to 
be in containment will not involve a significant increase of the 
probability of an accident previously evaluated.
    The Updated Safety Analysis Report (USAR), Section 14.15, ``Loss 
of Coolant Accident,'' does not take credit for a post-LOCA minimum 
containment sump pH adjustment to 7.0 for the iodine removal and 
retention calculation until ten hours after initiation of the event. 
Increasing the amount of TSP (based on recent re-analysis) in the 
containment sump ensures that a pH greater than or equal to 7.0 is 
achieved and therefore does not increase the consequences of any 
accident previously evaluated.
    The proposed change to TS 2.3(4) represents a new Limiting 
Condition for Operation (LCO) which is added to establish overall 
consistency with the CE STS [Combustion Engineering Standard 
Technical Specifications] for TSP requirements. The proposed change 
establishes a minimum TSP volume that must be maintained during 
operating Modes 1 and 2 to ensure that a pH greater than or equal to 
7.0 is achieved within four hours following a LOCA; as well as, 
establishing times for accomplishing corrective actions should the 
LCO not be met. Therefore, this change does not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    The proposed change to TS 3.6(2)d(i) revises the required 
surveillance inventory of the TSP baskets consistent with the 
aforementioned calculation to ensure that a pH greater than or equal 
to 7.0 is achieved. Therefore, this change does not increase the 
consequences of any accident previously evaluated.
    The proposed change to TS 3.6(2)d(ii) moves the surveillance 
test amounts of chemical and water used from the Specification to 
the Basis section. This relocation will not alter the test method or 
acceptance criteria.
    In the Basis, the amount of TSP used in the test is changed to 
reflect the ratio of TSP to water that would be found in the 
containment sump following a LOCA. The specified concentration of 
boron in the test reflects the highest concentration that could be 
found in the containment sump following a LOCA. The test temperature 
is changed to 115 - 125 deg.F, which is well below the temperature 
expected to be found in the containment sump following a LOCA. The 
decanting of the solution does not change the intent of the test 
method since the dissolving period will still be conducted without 
agitation. Therefore, these changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

[[Page 40026]]

    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    TSP is currently present in the containment sump. The addition 
of TSP ensures that a pH greater than or equal to 7.0 is achieved 
following a LOCA. The increase in TSP inventory will be accomplished 
via a modification to be installed during the 1996 Refueling Outage.
    The proposed change to TS 2.3(4) represents a new LCO which is 
added to establish overall consistency with the CE STS for TSP 
requirements. The proposed change establishes a minimum TSP volume 
that must be maintained during operating Modes 1 and 2 to ensure 
that a pH greater than or equal to 7.0 is achieved following a LOCA, 
as well as, establishing corrective action term limits should the 
LCO not be met. This proposed change does not create a possibility 
of a new or different kind of accident from any previously analyzed.
    The proposed change to TS 3.6(2)d(ii) moves the surveillance 
test amounts of chemical and water used from the Specification to 
the Basis section to be consistent with the CE STS. This relocation 
will not alter the test method or acceptance criteria. In the Basis 
section, the amount of TSP used in the test is changed to reflect 
the ratio of TSP to water that would be found in the containment 
following a LOCA. The specified concentration of boron in the test 
reflects the highest concentration that could be found in the 
containment sump following a LOCA. The test temperature is changed 
to a range of 115 - 125 deg.F which is well below the temperature 
expected to be found in the containment sump following a LOCA. The 
decanting of the solution does not change the intent of the test 
method since the dissolving period will still be conducted without 
agitation. Therefore, these changes will not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    TSP is stored in the containment lower level to raise the pH of 
the containment sump and recirculated spray water following a LOCA. 
As the pH of the water increases, more radioactive iodine is kept in 
solution and the possibility of airborne radioactivity leakage is 
decreased. Additionally, a higher pH has the beneficial effect of 
reducing the possibility of chloride stress corrosion cracking of 
austenitic stainless steel components in the containment.
    The proposed change to TS 2.3(4) represents addition of a new 
LCO for TSP requirements during power operations and hot standby 
consistent with CE STS. This change does not involve a significant 
reduction in a margin of safety.
    TS 3.6(2)d(i) requires verification that a minimum volume of TSP 
is contained in the storage baskets in containment. This change 
proposes to increase that volume consistent with the latest ABB/CE 
calculation. The increased volume will ensure that the containment 
sump, when filled with water from the Reactor Coolant System, Safety 
Injection Refueling Water Tank, Safety Injection Tanks and Boric 
Acid Storage Tanks, will have a pH greater than or equal to 7.0 
within four hours following a LOCA. Therefore, this change does not 
involve a reduction in a margin of safety.
    The proposed change to TS 3.6(2)d(ii) would move the 
surveillance test amounts of chemical and water used from the 
Specification to the Basis section. This relocation is consistent 
with the CE STS and will not alter the test method or acceptance 
criteria. In the Basis, the amount of TSP used in the test is 
changed to reflect the ratio of TSP to water that would be found in 
the containment following a LOCA. The specified concentration of 
boron in the test reflects the highest post-LOCA concentration that 
could be found in the containment. The test temperature is changed 
to a range of 115 - 125 deg.F which is well below the temperature 
expected to be found in the containment sump following a LOCA. The 
decanting of the solution does not change the intent of the test 
method since the dissolving period will still be conducted without 
agitation. Therefore, these 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, NW., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station, Unit No. 1, Washington County, Nebraska

    Date of amendment request: July 15, 1996
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to allow the use of either 
zircaloy or ZIRLO cladding and add a reference to Westinghouse Topical 
Report, WCAP-12610, June 1990.
    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 revision to TS 4.3.2 is based on improved STS 4.2 
of NUREG-1432. ZIRLO is similar in chemical composition, physical 
and mechanical properties to Zircaloy-4, but features improved 
corrosion performance and dimensional stability. These 
characteristics ensure that fuel rod cladding integrity and fuel 
assembly structural integrity are maintained. Fuel assemblies 
manufactured with ZIRLO clad fuel rods meet the same design bases 
requirements as fuel assemblies manufactured with Zircaloy-4 
cladding and the regulatory requirements of 10 CFR 50.46 are 
applicable to either material.
    No concerns have been identified pertaining to reactor operation 
with a core comprised of fuel assemblies manufactured with Zircaloy-
4 clad rods and fuel assemblies manufactured with ZIRLO clad rods. 
ZIRLO clad fuel rods do not require a change to the FCS [Fort 
Calhoun Station] reload design and safety analysis limits. 
Radiological consequences of previously evaluated accidents are not 
increased because the safety analysis dose predictions are not 
sensitive to the type of cladding material used. The proposed 
limited substitution of zirconium alloy or stainless steel filler 
rods in accordance with NRC-approved fuel rod configurations will 
allow leaking fuel rods (or potential leakers) to be removed. 
Therefore, the radiological consequences of accidents previously 
evaluated in the FCS Updated Safety Analysis Report (USAR) are not 
increased by this change.
    The revisions to TS 4.3.2 listed above will not result in a 
change to any of the process variables that might initiate an 
accident or affect the radiological release for an accident. The 
operating limits will not be changed and the analysis methods to 
demonstrate operation within the limits will remain in accordance 
with NRC-approved methodology. There are no physical changes to the 
plant associated with the change to TS 4.3.2 other than the changes 
to the fuel assemblies. Therefore, this revision does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated because the safety analysis to be 
performed for each cycle will continue to demonstrate compliance 
with all fuel safety design bases.
    The proposed revision of TS 4.3.2 is supported by Westinghouse 
Topical Report, WCAP-12610, ``VANTAGE + Fuel Assembly Report,'' 
dated June 1990 (Westinghouse Proprietary). This topical report 
describes the fuel rod design bases, criteria and models, which are 
affected by the use of ZIRLO cladding. Consequently, WCAP-12610 is 
proposed for addition to the list of analytical methods located in 
TS 5.9.5b that are used to determine the core operating limits.
    Based on the above discussion, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Fuel assemblies manufactured with ZIRLO clad fuel rods must meet 
original design criteria and thus they will not be an initiator for 
any new or different kind of accident. All design and performance 
criteria will continue to be met by fuel assemblies manufactured 
with ZIRLO clad fuel rods and

[[Page 40027]]

no new single failure mechanisms have been found.
    The use of fuel assemblies manufactured with ZIRLO cladding does 
not involve any alterations to plant equipment or procedures that 
would introduce any new or unique operational modes or accident 
precursors. The substitution of zirconium alloy, stainless steel 
filler rods, or lead test assemblies for fuel rods will be limited 
to NRC-approved fuel rod configurations. Therefore, the possibility 
of a new or different kind of accident from any accident previously 
evaluated is not created by this change.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of fuel assemblies manufactured with ZIRLO clad rods 
does not change the proposed FCS reload design and safety analysis 
limits. The normal operating conditions allowed for in the Technical 
Specifications will be taken into consideration for the use of these 
fuel assemblies. For each cycle reload core, the fuel assemblies 
will be evaluated using NRC-approved reload design methods to 
include consideration of the core physics analysis peaking factors 
and core average linear heat rate effects.
    NRC-approved methods will also be used to analyze each 
configuration of zirconium alloy or stainless steel filler rods in 
fuel assemblies to demonstrate continued safe operation within the 
limits that assure acceptable plant response to accidents and 
transients. Therefore, this 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

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

    Date of amendment request: June 21, 1996
    Description of amendment request: The proposed amendment would 
change the frequency of instrument channel calibrations in Table 4.1-1, 
``Minimum Frequencies for Checks, Calibrations and Test of Instrument 
Channels'' to accommodate operation with a 24-month operating cycle.
    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 operation with the proposed license amendment involve a 
significant increase in the probability or consequences of any 
accident previously evaluated?
    Response:
    The proposed changes do not involve a significant increase in 
the probability or consequences of any accident previously 
evaluated. The proposed changes are being made to extend the 
calibration frequency to 24-months for the:
    Pressurizer Pressure; Accumulator Level and Pressure; andVolume 
Control Tank Level.
    These changes are being made, using the guidance of Generic 
Letter 91-04, to accommodate a 24-month operating cycle. The 
proposed changes in the calibration frequencies do not involve any 
plant hardware changes (other than alarm adjustments) or the way the 
systems function. The results of the instrumentation drift analysis, 
loop accuracy/set point calculations and the evaluation of channel 
uncertainties indicate the calibrations can be safely extended to 
accommodate the 24-month operating cycle.
    The four pressurizer pressure channels are used for high and low 
pressure protection (i.e., reactor trip and safety injection) and 
for overpower-overtemperature protection. Three of the pressure 
channels are also used for pressure control and compensation signals 
for rod control. Pressurizer pressure indication is also provided in 
the control room for use during normal operation and while using the 
EOPs (emergency operating procedure). The loop accuracy/setpoint 
calculations confirm that sufficient margin exists between the 
pressurizer high and low pressure reactor trip, low pressurizer 
pressure SI [safety injection], and overtemperature delta-
temperature analytical limits and the existing field trip settings 
based on an extended calibration interval. A small increase in 
pressurizer pressure normal indication uncertainty due to increased 
sensor drift is within the readability of the indicator and has been 
incorporated into the pressurizer pressure initial conditions used 
in the evaluation of channel uncertainties (Reference 15) [see 
application dated June 21, 1996]. The post-accident indication 
uncertainties remain bounded by the existing uncertainties used in 
the EOPs. Assurance that the RPS [reactor protection system] and ESF 
[engineered safety feature] instrumentation and protection logic 
relays will function as required is also provided by on-line 
surveillance (channel checks performed each shift and quarterly 
channel functional tests) that are designed to detect potential 
instrument failures and verify operability of pressurizer pressure 
channels.
    Water level and pressure in each accumulator is monitored by two 
redundant channels designed to provide indication in the control 
room. High and low level alarm functions alert the operator to 
initiate operations to maintain the accumulator water volume or 
pressure within the Technical Specifications limits. The level and 
pressure instrumentation do not provide an active protective or 
control function and are not required to mitigate an accident 
condition. The level (or volume) and pressure limits are important 
since they are initial conditions assumed in the safety analysis. 
The loop accuracy/setpoint calculations for accumulator level and 
pressure were updated to include conservative values for 30-month 
calibration uncertainties using Westinghouse sensor drift values and 
extrapolated vendor specified uncertainties for rack and indicating 
components consistent with industry methods. The increased indicator 
uncertainty has been evaluated for both input parameters 
(accumulator level and pressure) assumed for the LOCA [loss-of-
coolant accident] and Containment Integrity events (Reference 15) 
and a non significant increase in both the peak clad temperature and 
containment pressure was identified.
    The volume control tank (VCT) level instrumentation is not 
required to mitigate the consequences of an accident. The 
instrumentation provides control room indication and initiates 
automatic actions of the chemical and volume control system (e.g., 
diverts letdown to the holdup tanks on high level, initiates makeup 
on low level, changes the charging pump suction on low low level). 
The loop accuracy/setpoint calculation for VCT level, updated based 
on the increased drift and uncertainty, determined that the existing 
setpoints remain valid to ensure the VCT instrumentation can perform 
the required design function.
    2. Does operation with the proposed license amendment create the 
possibility of a new or different kind of accident from any 
previously evaluated?
    Response:
    The proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated. The 
proposed changes extend the calibration frequency to 24 months for 
the Pressurizer Pressure, Accumulator Pressure and Level, and Volume 
Control Tank Level instrumentation to accommodate a 24-month 
operating cycle. The proposed changes in calibration frequencies do 
not involve any plant hardware changes, nor do they change the way 
that the systems function.
    The extension of the calibration and surveillance test intervals 
were evaluated and the results, documented in Reference 15, indicate 
that the calibrations can be safely extended to accommodate the 24-
month operating cycle.
    3. Does operation with the proposed license amendment involve a 
significant reduction in a margin of safety?
    Response:
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed changes extend the calibration 
frequency to 24 months for the Pressurizer Pressure, Accumulator 
Pressure and Level, and Volume Control Tank Level instrumentation to 
accommodate a 24-month operating cycle.
    The proposed changes result in an increased instrument channel 
uncertainty for the pressurizer pressure. An evaluation (Reference 
15) has determined that: all

[[Page 40028]]

current cycle 9 safety analysis limits based on pressurizer pressure 
uncertainties remain bounding for extended surveillance intervals 
(high and low pressure trips); the safety analysis limits for K1 (a 
constant used in the overtemperature [DELTA] T trip setpoint) remain 
applicable; and, Engineered Safety Feature Actuation System trip 
settings based on pressurizer pressure uncertainty remain bounding 
(low pressure safety injection).
    The proposed changes result in an increased instrument channel 
uncertainty for the accumulator level and pressure. An evaluation 
(Reference 15) has determined that increasing the uncertainty 
results in non-significant (defined by 10 CFR 50.46(a)(3)(i) as less 
than 50 deg.F) increases in the total peak clad temperature (less 
than 35 deg.F) for the large break and small break LOCA but the 
values remain well within regulatory acceptance criteria. The 
evaluation also determined that the peak calculated pressure in 
containment following a LOCA would increase due to the lower bound 
on pressure and the higher bound on volume in the accumulators. An 
assessment of the approximate effect on the peak containment 
pressure determined that the Technical Specification integrated leak 
rate testing value of 42.42 psig (the licensing basis peak pressure) 
remains bounding.
    The proposed changes result in an increased instrument channel 
uncertainty for the VCT level but there are no changes to any 
margins of safety because this instrumentation supports a control 
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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Jocelyn A. Mitchell, Acting Director

Southern California Edison Company, et al., Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit No. 1, San Diego County, 
California

    Date of amendment request: December 22, 1995
    Description of amendment request: The proposed change would revise 
the San Onofre Unit 1 License Condition to delete a reference to 
License Condition 2.C(4) from License Condition 2.D. This change is 
being requested to eliminate a reporting requirement for violations of 
the physical protection plans that is redundant to reporting 
requirements in 10 CFR 73.71 and 10 CFR 73 Appendix G.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility according to this proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. The proposed change is considered an administrative change. 
It has no impact on the probability or consequences of any of the 
accidents previously evaluated. This change revises License 
Condition 2.D to remove the burden of duplicate reporting 
requirements. This change does not affect the physical protection 
program as previously approved by the Nuclear Regulatory Commission 
(NRC).
    A reporting requirement in License Condition 2.D is being 
revised to remove the reference to License Condition 2.C(4) for the 
physical protection program. The reporting requirements for the 
physical protection program are located in the regulations, 10 CFR 
73.71 and 10 CFR 73 Appendix G.
    Therefore, the probability and consequences of an accidently 
previously evaluated are not affected by these proposed changes.
    2. Will operation of the facility according to this proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No. This proposed change is considered an administrative change. 
It has no impact on equipment, systems, or structures such that a 
new or different kind of accident is created. This change revises 
License Condition 2.D to remove duplicate and unnecessary reporting 
requirements for the physical protection program. There is no change 
associated with the implementation and maintenance of the physical 
protection program as previously approved by the NRC.
    Therefore, the possibility of a new or different kind of 
accident from an accident previously evaluated is not created.
    3. Will operation of the facility according to this proposed 
change involve a significant reduction in a margin of safety?
    No. This proposed change is considered an administrative change 
only. It has no impact on the margin of safety associated with the 
physical protection program. This change revises License Condition 
2.D to remove duplicative and unnecessary reporting requirements for 
the physical protection program. The maintenance and implementation 
of the physical protection program is not affected by this change.
    Therefore, there will not be a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the analysis of the licensee 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: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713
    Attorney for licensee: James A. Beoletto, Esquire, Southern 
California Edison Company, P.O. Box 800, Rosemead, California 91770
    NRC Project Director: Seymour H. Weiss

Southern California Edison Company, et al., Docket No. 50-206, San 
OnofreNuclear Generating Station, Unit No. 1, San Diego County, 
California

    Date of amendment request: March 13, 1996
    Description of amendment request: The proposed change would revise 
San Onofre Unit 1 License Condition 2.D in the Operating (Possession 
Only) License to remove a reporting requirement that is redundant to 
reporting requirements in 10 CFR 50.72 and 50.73. Additionally, the 
proposed change would make administrative and editorial changes in the 
Permanently Defueled Technical Specifications, which constitute 
Appendix A of the Operating (Possession Only) License.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility according to this proposed 
change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    No. San Onofre Nuclear Generating Station, Unit 1 (SONGS 1) has 
been permanently shut down with its reactor defueled and spent fuel 
from the reactor stored in the spent fuel pool. The proposed change 
will not modify any of the existing plant configurations, controls, 
procedures, or Permanently Defueled Technical Specifications (PDTS) 
requirements necessary to assure the integrity and safe operation of 
the spent fuel pool.
    The requested change to License Condition 2.D will result in not 
requiring violations of the PDTS to be reported based on License 
Condition 2.D. The basis for this change is that all types of 
reportable events applicable to a defueled plant are covered by 10 
CFR 50.72 and 50.73, which SONGS 1 is required to implement. Any 
other reporting requirements imposed through a license condition are 
redundant to reporting requirements contained in 10 CFR 50.72 and 
50.73. Therefore, this change is administrative.
    The requested changes to the PDTS are also administrative in 
nature. They consist of changes to reflect the current nuclear 
organization and responsibilities, modify administrative 
requirements relating to the Onsite Review Committee, modify a 
requirement relating to Final Safety Analysis Report documentation 
using NRC guidance, and make editorial corrections and improvements 
in the text. Since these changes are administrative, they have no 
effect on the accidents previously evaluated.

[[Page 40029]]

    Therefore, operation of the facility in accordance with this 
proposed change will not involve a significant increase in the 
probability or consequences of an accidently previously evaluated.
    2. Will operation of the facility according to this proposed 
change create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No. The proposed changes do not alter the design, configuration, 
or method of operation of the plant. The changes to License 
Condition 2.D and the PDTS are administrative or editorial.
    Therefore, operation of the facility in accordance with this 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Will operation of the facility according to this proposed 
change involve a significant reduction in a margin of safety?
    No. The proposed changes do not alter the design, configuration, 
or method of operation of the plant. Since the proposed changes are 
administrative or editorial, the existing plant safety margins are 
not reduced.
    Therefore, operation of the facility in accordance with this 
proposed change will not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the analysis of the licensee 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: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713
    Attorney for licensee: James A. Beoletto, Esquire, Southern 
California Edison Company, P.O. Box 800, Rosemead, California 91770
    NRC Project Director: Seymour H. Weiss

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

    Date of amendment requests: May 29, 1996
    Description of amendment requests: The licensee proposes to revise 
improved Technical Specification (TS) 3.5.1, ``Safety Injection Tanks 
(SITs),'' to increase the minimum boron concentration in the safety 
injection tanks from 1850 parts per million (ppm) to 2200 ppm. This TS 
change is being requested to support the planned increase in the 
operating cycle length.
    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.
    Southern California Edison (Edison) is increasing the minimum 
boron concentration to maintain the ability of the Safety Injection 
Tanks (SITs) to perform their intended safety function consistent 
with the increase in fuel enrichment up to 4.8 weight percent (w/o) 
Uranium-235 and changing the burnable poison from B4C to Erbia 
(Erbium-Oxide Er2O3 and fuel mixture) to increase the 
length of the operating cycle. Increasing the minimum boron 
concentration in the SITs will maintain the ability of the Emergency 
Core Cooling System (ECCS) to control core reactivity during and 
following an accident.
    No change is being made to the design of the safety injection 
system. Consequently, there will be no impact on the probability of 
initiating an accident which has been previously evaluated.
    Increasing the boron concentration in the SITs will ensure the 
ability of this system to mitigate the accidents for which it is 
required. No other accident conditions, design conditions, Technical 
Specifications, or Technical Specification Bases are affected by 
this proposed change in boron concentration.
    Therefore, the operation of the facility in accordance with this 
proposed change does not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There is no change in plant design or operational methodology 
imposed by the increase in SIT boron concentration. This increase in 
boron concentration is required because Edison is increasing the 
fuel enrichment up to 4.8 w/o Uranium-235 and changing the burnable 
poison from B4C to Erbia to achieve a longer cycle length. 
Therefore, additional negative reactivity is required at the 
beginning of the fuel cycle for these alternate coolant sources.
    Edison believes this change in the SIT minimum boron 
concentration limit is, in essence, an administrative change. The 
SITs are filled from the refueling water storage tank (RWST), which 
has a technical specification minimum boron concentration 
requirement of 2350 ppm. Edison maintains the RWST boron 
concentration higher than the minimum limit. As a result, for the 
past several years the SIT boron concentration has been 
approximately 2500 ppm, even though the technical specification 
lower limit is 1850 ppm. The maximum boron concentration limit is 
not being changed. Increasing the SIT minimum boron concentration 
limit of the technical specification narrows the existing operating 
band, and maintaining the boron concentration between 2200 ppm and 
2800 ppm will keep the boron concentration between the current band 
of 1850 ppm to 2800 ppm. Therefore, changing the SIT minimum boron 
concentration from 1850 ppm to 2200 ppm does not involve a physical 
change to the plant.
    Therefore, the operation of the facility in accordance with this 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    With the increase in fuel enrichment up to 4.8 w/o Uranium-235 
and changing the burnable poison from B4C to Erbia to increase 
the length of the operating cycle, increasing the minimum boron 
concentration in the SITs is required to maintain the current 
margins of safety.
    The calculations were performed to ensure the core remains 
subcritical (i.e., conservatively 1% shutdown) with the proposed 
boron concentration. In addition to the conservative assumptions 
used in the calculation, 50 ppm was added to the results.
    Therefore, the operation of the facility in accordance with this 
proposed change does not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: William H. Bateman

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama

    Date of amendments request: June 12, 1996
    Description of amendments request: The proposed amendments would 
revise the reactor core safety limits, Overtemperature delta T (OTDT) 
and Overpressure delta T (OPDT) reactor trip setpoints and allowable 
values, and the power distribution limits associated with 
implementation of Relaxed Axial Offset Control (RAOC) and FQ 
surveillance. The proposed amendments also include changes to the Bases 
associated with these specifications and surveillances.
    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:

[[Page 40030]]

    1. The proposed safety limits, reactor trip setpoints, HNF [high 
neutron flux] setpoints for MSSVs [main steamline safety valves] out 
of service, F[delta]H for LOPAR [low parasitic], and RAOC strategy 
changes do not increase the probability or consequences of an 
accident previously evaluated in the FSAR [Final Safety Analysis 
Report]. The core safety limits and trip setpoints were determined 
using the NRC reviewed and approved DNB [departure from nucleate 
boiling] methodologies, namely RTDP, and approved DNB correlations. 
No new performance requirements are being imposed on any system or 
component in order to support the revised core limits. Overall plant 
integrity is not reduced. The DNB sensitive transients that are 
protected by [OPDT] and [OTDT] were reanalyzed or evaluated. The DNB 
design criterion continues to be met. None of these changes directly 
initiate an accident; therefore, the probability of an accident has 
not increased. No new performance requirements are imposed on any 
safety-related equipment. The acceptance criteria for the reanalyses 
continue to be met; therefore, the consequences of accidents 
previously evaluated in the FSAR are not significantly changed. All 
dose consequences have been evaluated for these changes and all 
acceptance limits continue to be met. All safety analyses that use 
the revised [OTDT] and [OPDT] setpoints continue to meet all 
acceptance criteria. [Loss-of-coolant accident] LOCA analyses are 
not affected by any of these proposed changes.
    2. The proposed Technical Specifications changes do not create 
the possibility of a new or different kind of accident than any 
accident already evaluated in the FSAR. No new accident scenarios, 
failure mechanisms or limiting single failures are introduced as a 
result of the proposed changes. The proposed Technical 
Specifications changes have no adverse effects on any safety-related 
system and do not challenge the performance or integrity of any 
safety-related system. The DNB design criterion continues to be met. 
The use of the revised core limits, reactor trip setpoints and RAOC 
have been shown to allow FNP [Farley Nuclear Plant] to operate in a 
safe configuration. Therefore, the possibility of a new or different 
kind of accident is not created.
    3. The proposed Technical Specifications changes do not involve 
a significant reduction in a margin of safety. All accident analysis 
acceptance criteria continue to be met. The DNB design criterion 
remains unchanged. The DNBR [departure from nucleate boiling ratio] 
design limit values have not changed. Therefore, the DNB design 
limit values associated with the DNB methodology and correlations, 
upon which the Technical Specifications changes are based, do not 
result in a significant reduction in the margin of safety because 
the DNB design criterion continues to be met. The proposed revisions 
to the Technical Specifications result in an operating configuration 
consistent with the analytic assumptions (including LOCA analyses) 
used to form the bases of the Technical Specifications.
    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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama

    Date of amendments request: June 20, 1996
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications (TS) to incorporate the 
requirements of 10 CFR Part 50, Appendix J, Option B. The 
Administrative Controls portion would be revised to establish and 
reference a ``Containment Leakage Rate Testing Program'' in accordance 
with the NRC's Regulatory Guide 1.163 dated September 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability of consequences of an accident previously evaluated. 
The proposed changes provide a mechanism within the TS for 
implementing a performance-based leakage rate test program which was 
promulgated by the revision to 10 CFR [Part] 50 to incorporate 
Option B to Appendix J. The proposed changes do not involve any 
physical or operational changes to structures, systems or 
components. The proposed TS Limiting Conditions for Operation (LCO) 
are consistent with 10 CFR [Part] 50, Appendix J requirements and 
are equivalent to the current LCO requirements. The current safety 
analyses and safety design basis for the accident mitigation 
functions of the containment, the airlocks, and the containment 
isolation valves are maintained. Since the allowable containment 
leakage is still maintained within the analyzed limit assumed in the 
accident analyses, there is no adverse effect on either onsite or 
offsite dose consequences. Furthermore, containment leakage is not 
an accident initiator. Therefore, these changes will not increase 
the probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
The proposed changes do not involve any physical or operational 
changes to structures, systems or components. No new failure 
mechanisms beyond those already considered in the current plant 
safety analyses are introduced. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety. Extending Type A, B, and C test intervals 
from those currently provided in the TS to those provided for in 10 
CFR [Part] 50 Appendix J, Option B slightly increases risk due to an 
increased likelihood of containment leakage corresponding to the 
increased testing intervals. However, this is somewhat compensated 
by the corresponding risk reduction benefits received from the 
reduction in component cycling, stress, and wear associated with the 
increased intervals. When considering the total integrated risk, 
which includes all analyzed accident sequences, the additional risk 
associated with increasing test intervals is negligible.
    The NRC letter to NEI [Nuclear Energy Institute] dated November 
2, 1995, recognizes that changes similar to the proposed changes at 
FNP [Farley Nuclear Plant] are required to implement Option B of 10 
CFR [Part] 50, Appendix J. In NUREG-1493, ``Performance-Based 
Containment Leak-Test Program,'' dated September 1995, which forms 
the basis for the Appendix J revision, the NRC concludes that 
adoption of performance-based test intervals for Appendix J testing 
will not significantly reduce the margin of safety. The containment 
leak rate data and component performance history at FNP are 
consistent with the conclusions reached in NUREG-1493 and NEI 94-01. 
Thus, the proposed license amendments do not involve a significant 
reduction in a margin of safety and will continue to support the 
regulatory goal of ensuring an essentially leak-tight containment 
boundary.
    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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

[[Page 40031]]

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 amendment request: May 28, 1996
    Description of amendment request: The proposed amendment would 
increase the test interval for Technical Specification (TS) 3/4.3.1.1, 
Reactor Protection System Instrumentation from monthly on a staggered 
test basis to semiannually on a staggered test basis for the control 
rod drive trip breakers and the reactor trip module logic. 
Additionally, the proposed amendment would increase the test interval 
from monthly to semiannually for the output logic of the anticipatory 
reactor trip system (ARTS) instrumentation as specified in TS 3/
4.3.2.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below: (1)
    Operation of the DBNPS in accordance with the proposed license 
amendment does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Increasing the surveillance interval will not affect the 
probability or consequences of an accident previously evaluated since 
performance of the surveillance test only ensures operability of the 
particular trip function at the time of the test. The licensee 
evaluated the maintenance history and surveillance test results of the 
control rod drive trip breakers, reactor trip module logic, and ARTS 
output logic to show these components have consistently met their 
design and operational requirements over the past 8 years.
    (2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not modify or affect system design, 
function, operation, or manner of testing.
    (3) Involve a significant reduction in a margin of safety.
    The licensee has performed a reliability evaluation that indicates 
insignificant change in reactor trip system unavailability and a 
reduction in the potential for spurious trips resulting from testing 
which support the conclusion that a significant reduction in a margin 
of safety will not occur.
    Based on the NRC staff review, it appears that the three standards 
of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: University of 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

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

    Date of amendment request: June 28, 1996
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications for shutdown margin to allow 
calculational determination of the highest worth control rod. Editorial 
changes are also included.
    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) During refueling, maintenance may be performed on either the 
control rods or the control rod drive mechanisms. Controls, such as 
refueling interlocks, are provided to assure inadvertent criticality 
does not occur during this maintenance. There are no proposed 
revisions to these controls except to lower the threshold for 
applicability, which constitutes a more restrictive change.
    These controls also continue to assure that the new, higher 
minimum shutdown margin is maintained to ensure the reactor can be 
returned to a subcritical condition should an inadvertent 
criticality occur. The proposed alternate calculational method for 
highest worth control rod has additional conservatism to account for 
any uncertainties in the calculation and provides equivalent margin. 
Therefore, this change will not significantly increase the 
probability or consequences of any previously analyzed accident.
    (2) The proposed change does not necessitate a physical 
alteration of the plant in that no new or different type of 
equipment will be installed. The proposed change does propose a 
higher minimum shutdown margin and a lower threshold of 
applicability for CRD [control rod drive] maintenance, both of which 
are more restrictive. The proposed change will provide effective 
methods to preserve the safety functions associated with the 
prevention or automatic mitigation of design basis accidents. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) The proposed changes to the controls provided to allow 
control rod withdrawal for the purposes of maintenance are more 
restrictive and thus preserve the safety functions associated with 
the prevention or automatic mitigation of design basis accidents. 
The addition of a higher minimum shutdown margin requirement and the 
proposed calculational alternative for highest worth rod, does not 
decrease any of the safety controls or functions to prevent 
inadvertent criticalities and provides equivalent or higher margins. 
Therefore, this change will not significantly reduce a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Attorney for licensee: R. K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624
    NRC Project Director: Jocelyn A. Mitchell, Acting Directorboro, VT 
05301

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

    Date of amendment request: July 3, 1996
    Description of amendment request: The proposed amendment would 
modify Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
Section 4.2.b, ``Steam Generator Tubes,'' to: revise the plugging 
criteria for tubes in the tubesheet crevice region; add new inspection 
criteria for tubes evaluated using the new plugging criteria; add 
definitions of terms used in the new plugging criteria; and add 
reporting requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. Operation of the KNPP in accordance with the proposed license 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The revised plugging criteria ensure that tubes in the tubesheet 
with indication(s) are sufficiently inspected and evaluated and, if 
necessary, rolled to meet the proposed

[[Page 40032]]

acceptance criteria based on the new definitions of acceptable 
distance between the indication and the rolled area. With sufficient 
distance between the indication(s) and the hard rolled region of the 
tube in the tubesheet, tube rupture probability and the consequences 
of tube rupture are the same as previously analyzed. Additionally, 
the potential for leakage is within previously analyzed limits.
    2. The proposed license amendment request does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the proposed tube plugging criteria and 
proposed inspection acceptance criteria based on the proposed 
definitions does not introduce any significant changes to the plant 
design basis. Use of these criteria will not introduce a mechanism 
that will result in an accident initiated outside of the tubesheet 
crevice region. Any hypothetical accident as a result of tube 
indications in the tubesheet crevice region of the tube will be 
bounded by the existing tube rupture analysis. Therefore, 
application of the revised acceptance criteria for indication(s) 
within the tubesheet crevice region will not create the possibility 
of a new or different kind of accident.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety.
    The use of the proposed inspection criteria and tube plugging 
acceptance criteria will maintain the integrity of the tube bundle 
commensurate with the requirements of Regulatory Guide 1.121 under 
normal and postulated accident conditions. The safety factors used 
in verification of the strength of tube(s) evaluated under the new 
plugging criteria are consistent with the safety factors in the ASME 
Boiler and Pressure Vessel Code used for steam generator design. The 
leak testing acceptance criteria are based on the primary-to-
secondary leakage limits in the TSs and the Updated Safety Analysis 
Report accident analyses will be maintained. Therefore, the proposed 
TS change will not result in a significant reduction in the margin 
of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed NoSignificant 
Hazards Consideration Determination,And Opportunity For A Hearing

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

Northeast Utilities Service Company, Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London, Connecticut

    Date of amendment request: July 3, 1996Brief
    Description of amendment request: The proposed amendments would 
provide a one-time change to Technical Specification 3.9.1, ``Refueling 
Operations, Boron Concentration.'' The proposed change would remove the 
requirement that the boron concentration in all filled portions of the 
Reactor Coolant System be ``uniform.''
    Date of publication of individual notice in Federal Register: July 
11, 1996 (61 FR 36583)
    Expiration date of individual notice: August 12, 1996
    Location Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut and the Wateford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, Connecticut

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

    Date of application for amendment: June 3, 1996, as superseded by 
application dated June 25, 1996
    Brief description of amendment request: The proposed amendment 
would revise Technical Specifications 3.3.11, ``Post Accident 
Monitoring Instrumentation,'' and 5.5.2.13, ``Diesel Fuel Oil Testing 
Program.'' The amendment would reinstate provisions of the current San 
Onofre Nuclear Generating Station, Unit Nos. 2 and 3 technical 
specifications that were revised as part of Amendment Nos. 127 and 116. 
These amendments adopted the recommendations of NUREG-1432, ``Standard 
Technical Specifications Combustion Engineering Plants.''
    Date of individual notice in Federal Register: July 2, 1996 (61 FR 
34452)
    Expiration date of individual notice: August 1, 1996
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: April 25, 1995
    Brief description of amendment request: The proposed amendment 
would add a reactor water cleanup system high blowdown containment 
isolation trip function and associated limiting condition for operation 
and surveillance requirements.
    Date of individual notice in Federal Register: June 28, 1996 (61 FR 
33777)
    Expiration date of individual notice: July 29, 1996
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: June 6, 1995, as supplemented by 
letter dated April 22, 1996.
    Brief description of amendment request: The proposed amendment 
would make administrative and editorial changes to Section 6.0 of the 
technical specifications for WNP-2.Date of individual notice in Federal 
Register: June 28, 1996 (61 FR 33779)
    Expiration date of individual notice: July 29, 1996
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

[[Page 40033]]

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.

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

    Date of application for amendment: March 29, 1996.
    Brief description of amendment: The amendment revises the technical 
specifications (TS) to add an allowance to complete a TS-required 
surveillance within 24 hours of discovery of a missed surveillance in 
accordance with the guidance of Generic Letter (GL) 87-09, ``Sections 
3.0 and 4.0 of the Standard Technical Specifications (STS) on the 
Applicability of Limiting Conditions for Operation and Surveillance 
Requirements.''
    Date of issuance: July 8, 1996
    Effective date: July 8, 1996
    Amendment No. 170
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25669) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 8, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550

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

    Date of application for amendments: November 2, 1994
    Brief description of amendments: The amendments delete the content 
of Appendix B, ``Environmental Protection Plan (EPP) 
(Nonradiological),'' and modify License Condition 2.C.(2) to delete 
that portion which refers to the EPP.
    Date of issuance: July 8, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 149 and 143
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Environmental Protection Plan and License Conditions.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25702) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 8, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: April 18, 1996
    Brief description of amendment: The amendment deleted a restriction 
on the 24-hour emergency diesel generator operation test in 
Surveillance Requirement 3.8.1.14 of the Technical Specifications for 
the Grand Gulf Nuclear Station, Unit 1. The deletion allows the test to 
also be conducted during power operation (i.e., during Modes 1 and 2), 
instead of the current requirement to only conduct the test when the 
plant is shut down.
    Date of issuance: July 15, 1996
    Effective date: July 15, 1996
    Amendment No: 124
    Facility Operating License No. NPF-29: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20847) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 15, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, 
Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, 
Claiborne County, Mississippi

    Date of application for amendment: May 6, 1996
    Brief description of amendment: The amendment reflects that the 
name of Mississippi Power & Light Company (MP&L) has been changed to 
Entergy Mississippi, Inc. The amendment revises Operating License No. 
NPF-29 and the Antitrust Conditions for the Grand Gulf Nuclear Station, 
Unit 1 (GGNS) to (1) add the phrase ``(now renamed Entergy Mississippi, 
Inc.)'', (2) replace the name of Mississippi Power & Light Company 
(MP&L) by the name Entergy Mississippi, Inc., and (3) replace a 
footnote by the statement: ``Amendment 125 resulted in a name change 
for Mississippi Power & Light Company (MP&L) to Entergy Mississippi, 
Inc.''.
    Date of issuance: July 16, 1996
    Effective date: July 16, 1996
    Amendment No: 125
    Facility Operating License No. NPF-29. Amendment revises the 
operating license.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28613) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 16, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

Florida Power and Light Company, et al., Docket No. 50-335 St. 
Lucie Plant, Unit No. 1, St. Lucie County, Florida

    Date of application for amendments: June 1, 1996
    Brief description of amendments: Revise Technical Specifications to 
reflect reduced reactor coolant system

[[Page 40034]]

flows resulting from increased percentage of plugged steam generator 
tubes.
    Date of Issuance: July 9, 1996
    Effective Date: July 9, 1996
    Amendment Nos.: 145
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 7, 1996 
(61FR29140). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 9, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of application for amendments: July 26, 1995
    Brief description of amendments: The amendments modify the 
Technical Specifications to allow operation with up to plus or minus 18 
steps of rod misalignment at or below 90 percent power.
    Date of issuance: July 12, 1996
    Effective date: July 12, 1996
    Amendment Nos. 186 and 180Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 
(60FR47616) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 12, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendment: May 7, 1996 (TSCR 247)
    Brief description of amendment: The amendment adopts the provisions 
of the Standard Technical Specifications, NUREG-1433, Rev. 1 which 
clarify surveillance requirement applicability and allow a maximum 
period of 24 hours to complete a surveillance requirement upon 
discovery that the surveillance has been missed.
    Date of Issuance: July 15, 1996
    Effective date: July 15, 1996
    Amendment No.: 185
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28615). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated July 15, 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

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

    Date of application for amendments: December 21, 1995
    Brief description of amendments: The amendments modify the Peach 
Bottom Atomic Power Station Units 2 and 3 Facility Operating Licenses 
to provide for elimination of outdated or superseded material 
regarding, among other things, environmental monitoring and 
modifications to the low pressure coolant injection system, and for 
making the FOLs for both units consistent.
    Date of issuance: July 15, 1996
    Effective date: Units 2 and 3, as of the date of issuance, to be 
implemented within 30 days.
    Amendments Nos.: 215 and 220
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: March 13, 1996 (61 FR 
10396) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 15, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

Public Service Electric & Gas Company, Docket No. 50-311, Salem 
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

    Date of application for amendment: May 7, 1996, as supplemented 
June 14, 1996
    Brief description of amendment: The amendment made a one-time 
change to Technical Specification 3/4.7.6, ``Control Room Emergency Air 
Conditioning System,'' which permits refueling of Unit 2 with the 
Control Room Emergency Air Conditioning System (CREACS) inoperable in 
Modes 5 and 6. This change will expire after the completion of the 
Control Room and CREACS upgrade, currently in progress, and the restart 
and entry into Mode 4 of Unit 2 from the current outage.
    Date of issuance: July 10, 1996
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment No. 165
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25710) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 10, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendments: April 22, 1996, as supplemented 
June 12, 1996
    Brief description of amendments: The amendments change the 
Technical Specifications to implement 10 CFR Part 50, Appendix J, 
Option B, for the Type A test by referring to Regulatory Guide 1.163, 
``Performance Based Containment Leakage-Test Program.''
    Date of issuance: July 11, 1996
    Effective date: Both units, As of date of issuance, to be 
implemented within 30 days.
    Amendment Nos. 184 and 166
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20856) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 11, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

Southern Nuclear Operating Company, Inc., Alabama Power Company, 
Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, 
Units 1 and 2, Houston County, Alabama

    Date of application for amendments: June 24, 1996
    Brief description of amendments: The amendments approve a unit 
cycle

[[Page 40035]]

specific (Unit 1, Cycle 14 and Unit 2, Cycle 11) Technical 
Specification change to Note 4 of Table 4.3-1 that permits continued 
operation of both Farley units without performing the required 
surveillance of the manual safety injection input to the reactor trip 
circuitry for the current operating cycle until the next unit shutdown, 
following which, this testing has to be performed prior to entering 
Mode 2.
    Date of issuance: July 19, 1996
    Effective date: July 19, 1996
    Amendment Nos.: 120 and 112
    Facility Operating License Nos. NPF-2 and NPF-8: The amendments 
revised the Technical Specifications.Public comments requested as to 
proposed no significant hazards consideration: Yes. (61 FR 34880 dated 
July 3, 1996). The 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 August 2, 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 a final no significant hazards consideration 
determination are contained in a Safety Evaluation dated July 19, 1996.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, P.O. Box 1369, Dothan, Alabama

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

    Date of application for amendment: January 2, 1996, as supplemented 
by letter dated April 12, 1996.
    Brief description of amendment: The amendment would revise TS 3.9.4 
and its associated Bases to allow the containment personnel airlock 
doors to be open during core alterations and movement of irradiated 
fuel in containment.
    Date of issuance: July 15, 1996
    Effective date: July 15, 1996, to be implemented within 30 days of 
the date of issuance.
    Amendment No.: 114
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5819). The April 12, 1996, supplemental letter provided 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 July 15, 1996.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendment: April 4, 1996
    Brief description of amendment: The amendment revises the Technical 
Specifications regarding secondary containment integrity including 
addition of required actions in the event secondary containment 
integrity is not maintained when required. It also requires 
surveillance of the secondary containment isolation valves under the 
licensee's in-service testing program.
    Date of issuance: July 10, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 147
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20859) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 10, 1996No significant 
hazards consideration comments received: No
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Dated at Rockville, Maryland, this 24th day of July 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga, Director,
Division of Reactor Projects - I/II Office of Nuclear Reactor 
Regulation
[Doc. 96-19317 Filed 7-30-95; 8:45 am]
BILLING CODE 7590-01-F