[Federal Register Volume 63, Number 222 (Wednesday, November 18, 1998)]
[Notices]
[Pages 64106-64132]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-30691]


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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 Pub. L. 97-415, the U.S. Nuclear Regulatory Commission 
(the Commission or NRC staff) is publishing this regular biweekly 
notice. Pub. L. 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 October 24, 1998, through November 5, 1998. 
The last biweekly notice was published on November 4, 1998 (63 FR 
59584).

[[Page 64107]]

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By December 18, 1998, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's

[[Page 64108]]

Public Document Room, the Gelman Building, 2120 L Street, NW., 
Washington DC, by the above date. A copy of the petition should also be 
sent to the Office of the General Counsel, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, and to the attorney for the 
licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: October 16, 1998.
    Description of amendment request: The proposed amendments would 
lower the power level below which the turbine control valve (TCV) and 
turbine stop valve (TSV) closure scram signals and the end-of-cycle 
recirculation pump trip (EOC-RPT) signals are not in effect. The bypass 
setpoint (Pbypass) would be reduced from 30 percent rated 
power to 25 percent rated power. The licensee also proposes to delete 
the reference to turbine first stage pressure as a measure of core 
thermal power in the Technical Specifications. To ensure that the trip 
functions will not be inadvertently bypassed when they are required to 
be operable, a requirement would be added to periodically verify that 
TCV and TSV scram trip functions and the ECO-RPT trip functions are not 
bypassed at greater than or equal to 25 percent of rated thermal power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated:
    The probability of an accident previously evaluated will not 
increase as a result of this change because the setpoint change does 
not alter any of the initiators of an accident or cause them to 
occur more frequently.
    The consequences of an accident previously evaluated are not 
impacted. LaSalle Units 1 and 2 each have approximately 30 percent 
bypass capability. Therefore, a scram on TCV or TSV closure signals 
is not needed until 30 percent core thermal power is reached, as 
adequate steam bypass capacity is available. A lower 
Pbypass remains conservative with respect to this 
criterion.
    LaSalle utilizes power and flow dependent thermal limits. The 
power dependent portion of these thermal limits is dependent on the 
Pbypass setpoint. These limits provide assurance that 
adequate fuel thermal-mechanical margin is maintained through 
adherence to the thermal limits Technical Specification 
requirements.
    Revised thermal limits have been determined based on the results 
of GE transient analyses. Adhering to these thermal limits ensures 
that the consequences of an accident or transient would not be 
increased from the consequences under the approved 30 percent 
setpoint. Adjustments to the thermal limits were determined through 
use of the NRC-approved ODYN reactor dynamic model for the limiting 
Load Rejection Without Bypass and the Feedwater Controller Failure 
events.
    The deletion of the reference to turbine first stage pressure 
and rewording the Technical Specifications Notes does not affect 
either accident initiators or plant equipment, as they are 
administrative changes.
    Adding the periodic verification that the bypass channels are 
set correctly ensures that scrams or EOC-RPT will not be 
inadvertently bypassed when Thermal Power is greater than or equal 
to 25 percent of Rated Thermal Power. The statement that 
specification 4.0.2 applies to the 18 month interval is needed, 
since the notes are not standard surveillance requirements and the 
interval is consistent with other similar instrumentation to which 
4.0.2 currently applies.
    Therefore, the proposed changes do not involve a significant 
increase in the probability of occurrence or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated:
    The setpoint change and proposed bypass verification notes 
ensure that the scrams for TSV closure and TCV fast closure, and 
EOC-RPT, will be enabled above 25 percent of rated thermal power, 
rather than above 30 percent of rated thermal power. This change 
results in simplified reload transient analyses and does not impact 
any other equipment.
    No other physical modifications are being proposed by this 
submittal. The only plant operational impact is that between 25 
percent and 30 percent power, the plant will now scram upon a 
turbine trip, which is an analyzed transient.
    The remaining changes to Technical Specification wording are 
administrative in nature and consistent with other Technical 
Specifications.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    (3) Involve a significant reduction in the margin of safety:
    LaSalle Units 1 and 2 each have approximately 30 percent bypass 
capability. Therefore, a scram on TCV or TSV closure signals in not 
needed until 30 percent core thermal power is reached, as adequate 
steam bypass capacity is available. However, reduction of this 
setpoint to 25 percent power actually aids the plant transient 
response between 25 percent and 30 percent power.
    The new thermal limits reflect the revised setpoint and have 
been determined based on revised limiting transient analyses that 
have included the new Pbypass value. If a transient were 
to occur, the revised operating limits ensure that adequate margin 
would be available to preclude violation of the Minimum Critical 
Power Ratio (MCPR) safety limit and the fuel thermal-mechanical 
limits.
    All other UFSAR [Updated Final Safety Analysis Report] events 
are either bounded by the analyses performed or are not impacted by 
the Pbypass change.
    The wording changes to the Technical Specifications do not 
change the requirement for the bypass function and for maintaining 
the bypass function and thus do not affect the analyses discussed 
above.
    The addition of the Notes periodically verifying the TCV and TSV 
Closure Trip Functions are not bypassed at greater than or equal to 
25 percent Rated Thermal Power ensures the trip functions will not 
be inadvertently bypassed when required to be Operable.
    Therefore, the proposed changes do 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 815 
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby, 
Illinois 61348-9692.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: July 22 and October 22, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to reflect the licensee's 
planned use of fuel supplied by Westinghouse. The

[[Page 64109]]

Westinghouse fuel has different design characteristics from the fuel 
currently in use. Accordingly, the following changes would need to be 
made to the TS: Figure 2.1.1-1, ``Reactor Core Safety Limits--Four 
Loops in Operation''; various core operating parameters specified by 
Surveillance Requirements 3.2.1.2, 3.2.1.3, and 3.2.2.2; Section 4.2.1, 
``Fuel Assemblies''; and Section 5.6.5, ``Core Operating Limits Report 
(COLR).''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, addressing the three standards of 10 CFR 50.92(c):

First Standard

    Implementation of this LAR [license amendment request] would not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. The revised Reactor Core Safety 
Limits Figure further restricts acceptable operation. Moving an 
uncertainty factor from the Improved Technical Specifications to the 
Core Operating Limits Report (COLR) does not exempt this factor from 
regulatory restrictions. COLR parameters are generated by NRC 
approved methods with the intent of ensuring that previously 
evaluated accidents remain bounding. The COLR is submitted to the 
NRC upon implementation of each fuel cycle or when the document is 
otherwise revised. No accident probabilities or consequences will be 
impacted by this LAR.

Second Standard

    Implementation of this LAR would not create the possibility of a 
new or different kind of accident from any previously evaluated. The 
revised Reactor Core Safety Limits Figure further restricts 
acceptable operation. Moving an uncertainty factor from the Improved 
Technical Specifications to the COLR does not exempt this factor 
from regulatory restrictions. Since the parameter in question is not 
being deleted, the possibility of a new or different kind of 
accident from any previously evaluated does not exist.

Third Standard

    Implementation of this LAR would not involve a significant 
reduction in a margin of safety. Margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design functions during and following an accident situation. 
These barriers include the fuel cladding, the reactor coolant 
system, and the containment system. Use of the ZIRLOTM 
cladding material has been reviewed and approved in Reference 1 (as 
listed in Chapter 2.1 of Topical Report DPC-NE-2009/DPC-NE-2009P, 
Duke Power Company Westinghouse Fuel Transition Report). 
ZIRLOTM cladding has been extensively used in 
Westinghouse nuclear reactors. The changes proposed in this LAR are 
necessary to ensure that the performance of the fission product 
barriers (cladding) will not be impacted following the replacement 
of one fuel design for another. No safety margin will be 
significantly impacted.

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

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. This notice supersedes 
the notice published on July 31, 1996 (61 FR 40015) in its entirety.
    Description of amendment request: For Beaver Valley Power Station, 
Unit No. 1 (BVPS-1) only, the proposed amendment would revise Technical 
Specification (TS) 4.4.5 and associated Bases; the Bases for TS 3/
4.4.6.2 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 4.4.5 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

[[Page 64110]]

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 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 &

[[Page 64111]]

Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: October 15, 1998.
    Description of amendment request: The proposed amendment would make 
several changes that are administrative in nature. The changes would 
(1) make editorial changes to delete obsolete material or material 
adequately described elsewhere, change action statement numbers, update 
the technical specification (TS) index pages, and make changes to be 
consistent with the guidance of the improved standard technical 
specifications (ISTS); (2) delete reporting requirements that duplicate 
reporting requirements contained in 10 CFR; and (3) relocate the 
requirement for meteorological monitoring instrumentation from the TS 
to the Licensing Requirements Manual.
    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?
    a. This change deletes an expired Unit 1 license condition and a 
Unit 2 license requirement that is not required since it is 
redundant to the reporting requirements addressed in 10 CFR 50.73. 
Deleting these requirements does not involve any increase in the 
probability or consequences of an accident previously evaluated.
    b. The reference to Specification 3.0.6 was omitted from 
Specification 3.0.1 in Unit 1 Amendment 213 and Unit 2 Amendment 90 
and is being added to 3.0.1 to be consistent with the Improved 
Standard Technical Specifications of NUREG 1431. This does not 
involve any increase in the probability or consequences of an 
accident previously evaluated.
    c. The Core Alteration definition has been updated to be 
consistent with the regulations and ISTS. The Offsite Dose 
Calculation Manual (ODCM) definition has been updated to be 
consistent with the change to Administrative Control 6.9.3. The 
Members of the Public definition has been changed to be consistent 
with 10 CFR 20.1003. This does not involve any increase in the 
probability or consequences of an accident previously evaluated.
    d. Changing Table 3.3-6 Action Statement 36 to Action Statement 
35 is an editorial change to eliminate redundant use of action 
statement numbers. This does not involve any increase in the 
probability or consequences of an accident previously evaluated.
    e. The technical specification index is being revised to address 
removal of the Meteorological Monitoring specification and title and 
page number changes to the administrative control reporting 
requirements section. The Meteorological Monitoring specification is 
being relocated to the Licensing Requirements Manual (LRM). 
Relocating the Meteorological Monitoring requirements is in 
accordance with the guidance in the Commission's Final Policy 
Statement and revisions to 10 CFR 50.36 on the content of the 
technical specifications and the ISTS. The Meteorological Monitoring 
requirements do not meet any of the criteria, 1 thru 4 of 10 CFR 
50.36 and can, therefore, be relocated from the Technical 
Specifications to the LRM. These changes do not involve any increase 
in the probability or consequences of an accident previously 
evaluated.
    f. The exclusion area boundary is adequately described in each 
unit's UFSAR [Updated Final Safety Analysis Report], therefore, 
design feature 5.1 Site Location is also being modified by deleting 
the description of the exclusion area boundary. This does not 
involve any increase in the probability or consequences of an 
accident previously evaluated.
    g. The change to refer to the Unit 1 Overpressure Protection 
System (OPPS) enable temperature in Specification 3.4.9.3 in lieu of 
specifying 275  deg.F was evaluated and found acceptable in the 
request for approval of Amendment 160. The deletion of the asterisk 
in Unit 2 Specification 3.9.8.1 was justified as part of the request 
for approval of Amendment 25. The inadvertent omission of the ACTION 
to take in the case that the temperature of the steam generator is 
precisely 50  deg.F above the cold leg temperature is being 
corrected. The cases of greater than and less than 50  deg.F are 
already included. These are editorial changes that do not involve 
any increase in the probability or consequences of an accident 
previously evaluated.
    h. The administrative control reporting requirements have been 
modified to incorporate various ISTS requirements. This requires 
changing titles and eliminating requirements addressed elsewhere, 
removing reference to deleted sections, and replacing reference to 
the administrative control section reporting requirements in various 
specifications with reference to 10 CFR 50.4. The 1993 NRC final 
policy statement set forth the criteria for determination of those 
requirements to be included in TS. The reporting requirements being 
removed from the TS do not meet the criteria for inclusion in the 
TS; therefore, the reporting requirements have been modified to 
reflect those requirements provided in the ISTS. These are editorial 
changes that do not involve any increase in the probability or 
consequences of an accident previously evaluated.
    i. The Technical Specification index has been modified to 
address the revised pages.
    These changes have been determined to be editorial and 
administrative in nature, and as such, would not affect any accident 
assumptions or radiological consequences of an accident. Therefore, 
the proposed changes would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The editorial changes, the elimination of reporting requirements 
which duplicate 10 CFR requirements and administrative improvements 
to incorporate the ISTS requirements are all changes that are 
administrative in nature. The proposed changes will not affect any 
plant system or structure, nor will they affect any system 
functional or operability requirements. Consequently, no new failure 
modes are introduced as a result of the proposed changes. Therefore, 
the proposed change will not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed amendment modifies reporting requirements and 
incorporates associated editorial changes that do not impact the 
UFSAR design basis or accident analyses assumptions. This change 
does not introduce any new operational modes or physical 
modifications to the plant; therefore, no action will occur that 
will involve a significant reduction in a margin of safety. In 
addition, the proposed change does not affect radiological release 
limits, monitoring equipment or operating practices. Therefore, the 
proposed amendment does not involve a significant reduction in a 
margin of safety.

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

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: September 23, 1998.
    Description of amendment request: The proposed amendment would 
change Division III battery specific gravity acceptance criteria 
outlined in River Bend Station (RBS) Technical Specifications (TS). The 
change is required as a result of battery system design modifications 
which are scheduled to be implemented in April 1999 during refueling 
outage (RF) RF-8. During this time, the current Division III

[[Page 64112]]

battery will be replaced. The new battery, which also will have a 
greater capacity rating, will be supplied with a nominal specific 
gravity of 1.215 at 77 deg.F in contrast to the existing Division III 
battery supplied with a nominal specific gravity of 1.210 at 77 deg.F. 
Since TS Section 3.8.6, Table 3.8.6-1 values for specific gravity are 
based on the manufacturer's nominal specific gravity, these values will 
need to be updated.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. This request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The system loads, voltage requirements, and inrush currents have 
been calculated in accordance with IEEE Std. 485, ``IEEE Recommended 
Practice for Sizing Large Lead Storage Batteries for Generating 
Stations and Substations.'' To support these design requirements at 
a capacity of 80%, a new battery must be installed. The nominal 
specific gravity of the new battery, as provided by the manufacturer 
of the battery, is 1.215 at 77 deg.F.
    A review of USAR Chapter 15, including Appendix 15A, was 
conducted to determine what accidents, if any, may be impacted by 
the proposed change to the Division III battery specific gravity.
    USAR Sections 15.2, ``Increase in Reactor Pressure;'' 15.3, 
``Decrease in Reactor Coolant System Flow Rate;'' and Section 15.6, 
``Decrease in Reactor Coolant Inventory'' discuss accidents that 
involve the initiation of HPCS when reactor vessel level drops to 
the initiation point. The function of the HPCS System is to mitigate 
the consequences of an accident (i.e., to maintain reactor vessel 
coolant inventory after small breaks which do not depressurize the 
reactor vessel, or provide spray cooling heat transfer following 
larger breaks, Ref. USAR Section 6.3.1.2.1). The function of the 
Division III 125 Vdc Power System is to provide a highly reliable, 
continuous, and independent source of control and motive power for 
the HPCS System logic, HPCS diesel generator set control and 
protection, and all Division III related control (Ref. USAR Section 
8.3.2.2.1). This is a support function for the HPCS System.
    USAR Section 15.5, ``Increase In Reactor Coolant Inventory,'' 
postulates an inadvertent HPCS actuation resulting from operator 
error. The proposed changes to the Division III battery specific 
gravity cannot result in an inadvertent HPCS actuation/injection. 
The proposed changes to the allowable specific gravity values 
provided in Technical Specification 3.8.6 are in agreement with the 
manufacturer's nominal specific gravity. The revision simply ensures 
that the battery has sufficient capacity to meet the energy 
requirements of its critical loads. The proposed change does not 
create any new internally generated missiles, nor does it affect the 
High Energy Line Break Analysis or any other accident described in 
Chapter 15 of the USAR. Neither the function nor the operation of 
the Division III battery is impacted by the proposed change.
    The replacement Division III battery will be supplied by the 
manufacturer with a nominal specific gravity of 1.215 at 77 deg.F. 
The battery manufacturer's rated performance is based on the 
specific gravity of the battery being maintained near the nominal 
specific gravity. Since the Division III design basis calculation 
depends on the battery manufacturer's rated performance, battery 
parameters upon which that performance is based must be monitored. 
The current Technical Specification values for specific gravity are 
based upon a nominal specific gravity of 1.210 at 77 deg.F. The 
proposed values accurately reflect the manufacturer's nominal 
specific gravity. Testing the Division III battery to the proposed 
values provides assurance that the HPCS functions supported by the 
125 Vdc System will not be adversely affected by the Division III 
battery.
    The proposed changes will not affect failure modes of existing 
equipment. The proposed changes do not affect the ability of any 
structures, systems or components to perform their safety functions. 
Therefore, no undue risk to the health and safety of the public has 
been created by the proposed changes, nor is there any change in the 
radiological consequences at the site boundary.
    By incorporating the correct value for battery specific gravity 
verification in Table 3.8.6-1, the Technical Specifications will 
accurately reflect the new design basis value for the Division III 
battery specific gravity. This change allows the performance of the 
Division III battery to be verified against the correct design basis 
value, thus providing assurance that the Division III 125 Vdc power 
system function will remain as assumed in the accident analysis. 
Therefore, the proposed change cannot affect any accidents 
previously evaluated (probability or consequences). Consequently, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. This request does not create the possibility of occurrence of 
a new or different kind of accident from any previously evaluated.
    Since a battery's capacity decreases as specific gravity 
decreases below the manufacturer's nominal value, monitoring the 
battery's specific gravity is one means of ensuring that the battery 
will adequately supply the minimum energy required to support the 
system function assumed in the accident analysis.
    All safety systems will continue to function as originally 
designed. The subject equipment will not function in a manner 
different than described in USAR Section 8.3.2.2. The functional and 
performance requirements of the Division III 125 Vdc System and its 
associated interfaces have not been altered. The proposed change 
simply ensures that the HPCS battery performance is verified against 
the correct design basis value. This value provides assurance that 
the HPCS System functions will not be adversely affected by the 
capacity of the battery. Therefore, the proposed changes do not 
create the possibility of occurrence of a new or different kind of 
accident from any previously evaluated.
    3. This request does not involve a significant reduction in a 
margin of safety.
    This proposed change updates the acceptance criteria of the 
current specific gravity for the Division III battery. This 
acceptance criteria is in accordance with manufactures 
recommendations. The design and license basis for the Division III 
systems and functions remain unchanged and the battery will continue 
to supply the 125 Vdc loads necessary to support these functions. 
This value will reflect the manufacturer's nominal specific gravity 
for the Division III battery. With the system functions supported as 
assumed in the accident analyses, the margin to safety remains 
unchanged.
    As a result, the proposed change does not involve a significant 
reduction in a margin to 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 are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW, Washington, DC 20005.
    NRC Project Director: John N. Hannon.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: October 8, 1998.
    Description of amendment request: The proposed amendment would 
implement Boiling Water Reactor Owners Group (BWROG) Enhanced Option I-
A (EIA) Reactor Stability Long Term Solution as documented in NEDO-
32339, Revision 1, ``Reactor Stability Long-Term Solution, Enhanced 
Option I-A.'' The EIA long term solution has been accepted by the NRC 
in Safety Evaluation Report, ``Reactor Stability Long-Term Solution, 
Enhanced Option I-A Generic Technical Specifications (TS), NEDO-32339, 
Supplement 4.''
    The proposed changes to the RBS TS will enable the full 
implementation of the Enhanced Option I-A (EIA) long term solution to 
the neutronic/thermal hydraulic instability issue. Specifically, the 
proposed change deletes the limits

[[Page 64113]]

on power and flow conditions associated with the implementation of the 
guidance in General Electric Service Information Letter #380, Revision 
1, ``BWR Core Thermal Hydraulic Stability'' (current TS 3.4.1, Figure 
3.4.1-1 and RBS plant procedures), adds new specifications, to 
establish limits for Fraction of Core Boiling Boundary (FCBB) and the 
Period Based Detection System (PBDS), modifies the RPS instrumentation 
specification and the description of the contents of the Core Operating 
Limits Report (COLR) in current TS 5.6.5. The two new specifications 
require maintaining stability control and the availability of a 
stability detection system during operation in defined regions of the 
power and flow operating domain.
    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 amendments do no involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendments allow the implementation of the Enhanced 
Option I-A (EIA) long term solution to the neutronic/thermal 
hydraulic instability issue. Current TS restrictions on power and 
flow conditions, number of operating recirculation loops, and 
operator actions implemented to reduce the probability of neutronic/
thermal hydraulic instability are eliminated and new stability 
requirements consistent with NEDO-32339-A, Supplement 4, Revision 1, 
are imposed.
    While the proposed amendments permit operation in regions of the 
power and flow operating domain postulated to be susceptible to 
neutronic/thermal hydraulic instability, the implementation of the 
EIA solution ensures there is not a significant increase in the 
probability or consequences of an accident previously evaluated. 
Operation in these regions does not increase the probability of 
occurrence of initiators and precursors of other previously analyzed 
accidents. The proposed amendments permit the implementation of the 
features of the EIA solution which prevent neutronic/thermal 
hydraulic instability. The features include pre-emptive reactor 
scram upon entry into the regions of the power and flow operating 
domain most susceptible to neutronic/thermal hydraulic instability--
the Exclusion Region. The EIA solution prevents neutronic/thermal 
hydraulic instability during operation in regions of the power and 
flow operating domain previously excluded from operation and 
therefore does not significantly increase the probability of a 
previously analyzed accident.
    The EIA solution also requires implementation of stability 
control prior to entry into a region of the power and flow operating 
domain which is potentially susceptible, in the absence of stability 
control, to neutronic/thermal hydraulic instability. The modified 
rod block functions providing the restricted region entry alarm 
(RREA), boiling boundary limits, and PBDS functions are required on 
entry into the Restricted Region of the power to flow map. The 
boiling boundary limits, and Period Based Detection System (PBDS) 
functions are required on entry into the Monitored Region of the 
power to flow map. The EIA solution prevents or allows for detection 
and suppression of neutronic/thermal hydraulic instability during 
operation in these regions of the power and flow operating domain.
    The EIA solution includes restrictions on power and flow 
conditions and actions associated with the modified APRM flow biased 
scram and RREA functions. Required actions include adherence to the 
boiling boundary limit stability control prior to entry and during 
operation in the region of the power and flow operating domain which 
is potentially susceptible to neutronic/thermal hydraulic 
instability--in the absence of stability control. In addition, the 
proposed amendments require operator actions based upon control room 
indications generated by a new PBDS. The PBDS is designed to provide 
alarm indication that conditions consistent with a significant 
degradation in the stability performance of the reactor have 
occurred and the potential for imminent onset of neutronic/thermal 
hydraulic instability may exist. The PBDS also provides analog 
indication of the highest and second highest successive period 
confirmation count for all of the LPRMs monitored. This provides the 
plant operators with continuous indication of reactor stability 
operating conditions. The PBDS system provides indication only and 
does not affect plant structures, systems, or components in any way 
that could increase the probability or consequences of an accident. 
Rather, the improved control room indications provide the operator 
with more accurate and timely information.
    The EIA solution allows for the ``Setup'' of APRM flow biased 
scram and control rod block function. The EIA solution requires 
adherence to certain boiling boundary limit stability controls prior 
to selection by the operator of APRM flow biased scram and control 
rod block function ``Setup'' setpoints. This ``Setup'' function 
allows operation in a region of the power and flow operating domain 
potentially susceptible to neutronic/thermal hydraulic instability 
provided the additional limits of the flow control boiling boundary 
(FCBB) and PBDS are met. After exiting the region requiring the 
stability control to be met, the setpoints can be manually reset to 
their normal values. Stability controls are required to be in place 
when setpoints are ``Setup''. As a backup EIA feature, the APRM flow 
biased setpoints automatically reset to their normal values above a 
pre-determined flow condition. This automatic reset to the more 
conservative setpoints ensures that the pre-emptive reactor scram 
will prevent operation as a result of an anticipated operational 
occurrence in the region most susceptible to neutronic/thermal 
hydraulic instability should the operator not select the more 
conservative setpoints appropriate for operation following exit from 
the region requiring stability control. The FCBB, PBDS, and 
automatic reset of the APRM flow biased scram and control rod block 
function ``setup'' setpoints allow for the use of the ``setup'' 
feature and help ensure that there is not an increase in the 
probability or consequences of an accident.
    Operation in the regions of the power and flow operating domain 
excluded by current TS 3.4.1 and Figure 3.4.1-1 can occur as a 
result of anticipated operational occurrences. In the absence of 
operator actions the severity of these anticipated operational 
occurrences may increase due to the potential occurrence of 
neutronic/thermal hydraulic instability as a result of operation in 
these regions. Upon entry, as a result of an anticipated operational 
occurrence, into the region most susceptible to neutronic/thermal 
hydraulic instability the pre-emptive reactor scram prevents 
neutronic/thermal hydraulic instability. Therefore, the consequences 
of an accident do not significantly increase while operating with 
stability control in place.
    The required EIA features is designed to limit possible 
neutronic/thermal hydraulic instabilities and to detect and suppress 
further neutronic/thermal hydraulic instabilities. These features 
include: a pre-emptive automatic scram, the control rod block and 
alarms associated with entry into the region susceptible to 
neutronic/thermal hydraulic instabilities, automatic reset of APRM 
flow biased setpoints, PBDS, FCBB, and the required operator 
actions, including manual reactor scram. Therefore, the proposed 
amendments prevent the occurrence of neutronic/thermal hydraulic 
instability during operation or as a consequence of an anticipated 
operational occurrence and do not significantly increase the 
consequences of any previously analyzed accident.
    2. The proposed amendments do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The proposed amendments eliminate existing restrictions on power 
and flow conditions and impose alternative restrictions which permit 
the implementation of the EIA long term stability solution. The 
current restrictions on the power and flow conditions do not prevent 
entry into regions of the power and flow operating domain most 
susceptible to neutronic/thermal hydraulic instability and therefore 
the possibility of neutronic/thermal hydraulic instability exists in 
the absence of operator action. The required features of the EIA 
solution implement a pre-emptive scram upon entry into the region 
most susceptible to neutronic/thermal hydraulic instability, without 
operator action. The accessible operating domain allowed by the 
proposed amendments is essentially a subset of the power and flow 
operating domain currently allowed. Initial conditions are bounded 
by the current initiators and precursors of accidents and 
anticipated operational occurrences. Accordingly, no new accident of 
initiator is present. Therefore, the proposed amendments do not 
create the possibility of a new or different kind of accident from 
that previously evaluated.
    Concurrent with the implementation of the proposed amendments, a 
modified Flow

[[Page 64114]]

Control Trip Reference (FCTR) card, EIA FCTR card, and a new Period 
Based Detection System (PBDS) will be installed as required by the 
EIA solution. The function of the EIA FCTR card is to aid the 
operator in the identification of entry into regions of the power 
and flow operating domain potentially susceptible to neutronic/
thermal hydraulic instability in the absence of stability controls 
and to initiate a pre-emptive scram upon entry into the regions most 
susceptible to neutronic/thermal hydraulic instability. This is 
accomplished by altering the existing values of setpoints of the 
APRM flow biased scram and the control rod block functions generated 
by the EIA FCTR card.
    The design of the EIA digital FCTR card is a functional 
equivalent of the original analog FCTR card. The Failure Modes and 
Effects Analysis (FMEA) for the card detailed in NEDC-32339P-A 
Supplement 2 found no single failure that would increase the 
consequences of an accident. The EIA FCTR card maintains the 
original basis for the NMS interface functions of the analog FCTR 
card it replaces. The plant specific environmental conditions 
(temperature, humidity, pressure, seismic, and electromagnetic 
compatibility) have been confirmed to be enveloped by the 
environmental qualification values for the EIA FCTR cards. 
Therefore, the potential for spurious scrams or common mode failures 
induced by environmental effects (e.g., electromagnetic 
interference) is considered negligible. The installation of the EIA 
FCTR card will therefore not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The function of the PBDS is to provide the operator with an 
indication that conditions consistent with a significant degradation 
in the stability performance of the reactor has occurred and the 
potential for imminent onset of neutronic/thermal hydraulic 
instability may exist. This is accomplished by the installation of a 
new PBDS card in the Neutron Monitoring System in accordance with 
NRC approved BWROG and GE design. The PBDS card takes inputs from 
individual local power range monitors and provides analog indication 
of the highest and second highest successive period confirmation 
count, provides a Hi DR and Hi-Hi DR alarm, and INOP status 
indication to the operator in the control room. These displays can 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated. The plant specific 
environmental conditions (temperature, humidity, pressure, seismic, 
and electromagnetic compatibility) have been confirmed to be 
enveloped by the PBDS environmental qualification values. Therefore, 
the installation of the PBDS card will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendments do not involve a significant 
reduction in the margin of safety.
    The proposed amendments permit the implementation of the EIA 
long term solution to the stability issue. Under certain conditions, 
existing BWR designs are susceptible to neutronic/thermal hydraulic 
instability. GDC 10 of 10 CFR 50, Appendix A, requires that 
specified acceptable fuel design limits not be exceeded during 
anticipated operational occurrences. General Design Criterion (GDC) 
12 of 10 CFR 50, Appendix A, requires thermal hydraulic instability 
to be prevented by design or be readily and reliably detected and 
suppressed. When the design of the reactor system does not prevent 
the occurrence of neutronic/thermal hydraulic instability, 
instability is considered an anticipated operational occurrence. The 
proposed amendments and the associated design modifications provide 
automatic features and operational information to the Control Room 
that replace the existing BWROG Interim Corrective Actions (ICAs). 
Thus the EIA solution assures compliance with GDC-10 and GDC 12 by 
providing for reliable detection and suppression and by the 
prevention of neutronic/thermal hydraulic instability. This 
therefore precludes neutronic/thermal hydraulic instability from 
becoming a credible consequence of an anticipated operational 
occurrence. As a result the margins of safety are maintained.
    Analyses performed by the BWROG indicate that neutronic/thermal 
hydraulic instability induced power oscillations could result in 
conditions exceeding the MCPR SL prior to detection and suppression 
by the current design of the Neutron Monitoring System and Reactor 
Protection System. To ensure compliance with GDC 12 the BWROG 
developed Interim Corrective Actions (ICAs) to enhance the 
capability of the operator to readily and reliably detect and 
suppress neutronic/thermal hydraulic instability. The BWROG ICAs 
also provided additional guidance for monitoring local power range 
monitors beyond the requirements of current TS 3.4.1 to ensure 
adequate margin to the onset of neutronic/thermal hydraulic 
instability. Reliance on operator actions to comply with GDC 12 was 
accepted on an interim basis by the NRC pending final implementation 
of a long term solution to the stability issue. The modified design 
of the Reactor Protection System (APRM flow biased scram) and 
stability control prior to entry into a region of the power and flow 
operating domain which is potentially susceptible, in the absence of 
stability control, to neutronic/thermal hydraulic instability 
implemented with the EIA solution prevents neutronic/thermal 
hydraulic instability. In addition, significant backup protection 
features, including the PBDS and specified operator actions, are 
required to be implemented. As a result, the margin to the onset of 
neutronic/thermal hydraulic instability provided by the existing TS 
requirements and BWROG ICAs recommendations is not reduced by the 
implementation of the EIA solution. The EIA solution assures 
compliance with GDC 12 by the prevention of neutronic/thermal 
hydraulic instability and therefore precludes neutronic/thermal 
hydraulic instability from becoming a credible consequence of an 
anticipated operational occurrence. The consequences of anticipated 
operational occurrences will not increase and the margin to the MCPR 
SL will not decrease upon implementation of the EIA solution. 
Therefore, the proposed amendment does not involve a reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW, Washington, DC 20005.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: October 1, 1998.
    Description of amendment request: The proposed change modifies 
Technical Specification (TS) 3.3.3.7.3 and Surveillance Requirement 
4.3.3.7.3 for the broad range gas detection system. A change to 
Technical Specification Basis 3/4.3.3.7 has been included to support 
this change. This change to the TS is necessary for the installation of 
a new, more reliable broad range gas detection 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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The broad range gas detection system has no effect on the 
accidents analyzed in Chapter 15 of the Final Safety Analysis 
Report. It's only effect is on habitability of the control room, 
which will be enhanced by installation of the new monitoring system 
and this change to the Technical Specifications. Qualitative 
analysis based on a quantitative risk assessment has shown that the 
impact on operator incapacitation and subsequent core damage risk of 
the periodic automatic background/reference spectrum check is 
negligible and that the probability of malfunction of the BRGMs due 
to a slowly increasing toxic chemical concentration is negligible.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of

[[Page 64115]]

accident from any accident previously evaluated?
    Response: No.
    The proposed Technical Specification change in itself does not 
change the design or configuration of the plant. The new broad range 
toxic gas monitoring system performs the same function as the old 
system, but it accomplishes this function with increased 
reliability.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The broad range gas detection system has no effect on a margin 
of safety as defined by Section 2 of the Technical Specifications. 
Its only effect is on habitability of the control room, which will 
be enhanced by installation of the new monitoring system and this 
change to the Technical Specifications.
    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 NW, Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: July 30, 1998 (LAR-222).
    Description of amendment request: The proposed amendment will 
change the Improved Technical Specifications (ITS) to add a new 
Required Action for the existence of breaches in the Control Complex 
Habitability Envelope (CCHE) that are in excess of allowances. A new 
surveillance requirement for the performance of a periodic integrated 
leak test of the CCHE boundary on a 24-month frequency would also be 
added. Changes to the current Ventilation Filter Test Program (VFTP) 
are proposed to adopt current standards for laboratory testing, change 
acceptable values of control room emergency ventilation flow rate and 
filter differential pressure, and add the Auxiliary Building 
Ventilation Exhaust Filters to the VFTP. Conforming changes to the ITS 
Bases 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. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated. The Control Room Emergency Ventilation System (CREVS) and 
the Control Complex Habitability Envelope (CCHE) are designed to 
limit the radiation dose to the control room operating staff 
following a design basis accident. Since these systems are only 
effective in limiting dose following an accident, the existence of 
limited breaches in the CCHE, the performance of periodic leak 
tests, and changes to the Ventilation Filter Test Program (VFTP) 
would not increase the probability of occurrence of any evaluated 
event. The features of the CREVS and the Control Complex emergency 
filters, or the CCHE have no direct function in mitigating the 
offsite consequences of any evaluated accident. The Auxiliary 
Building exhaust filters are not credited with reducing offsite 
doses, however, if available would filter releases from the 
Auxiliary Building. Adding them to the VFTP will not increase the 
consequences calculated for any evaluated accident.
    The proposed changes are consistent with the revised control 
room operator dose calculations as presented in the Control Room 
Habitability Report dated July 1998. Since all calculated doses are 
within 10 CFR Part 50, Appendix A GDC 19 limits there is no 
significant increase in consequences.
    It is conceivable that the existence of additional breaches in 
the CCHE could result in an increase in operator dose, however the 
low probability of a catastrophic reactor accident, the relatively 
short time allowed for breaches to be open in excess of approved 
dose calculation assumptions, and the ability to close breaches 
expeditiously makes the risk increase insignificant.
    The changes to the ITS Bases improve information on the 
operation and function of CREVS, and establish that CREVS 
operability is dependent on maintaining CCHE integrity. The 
inclusion of this information reinforces the importance of 
maintaining the CCHE boundary, and will help to ensure the CREVS is 
capable of performing its intended safety function.
    The Control Room Habitability Report, dated July 1998, provided 
with this LAR presents the methodology used and the results of the 
operator dose calculations for the Maximum Hypothetical Accident, 
toxic gas release, and other design basis accidents. The report 
provides the information needed for NRC review of LAR 222, Revision 
I and the associated unreviewed safety question. This evaluation 
concludes that the current level of CCHE integrity provides adequate 
protection for the control room operator.
    Based on the foregoing, the proposed amendment does not 
significantly increase the probability or consequence of an accident 
previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Neither performance of periodic CCHE leak tests nor changes to 
the existing VFTP can create the possibility of a new or different 
kind of accident. During the period of time when CCHE breaches are 
greater than the design calculation, there exists the possibility 
that control room dose from an analyzed accident may be greater than 
specified in General Design Criterion 19. This condition will not 
however create the possibility of a new or different kind of 
accident. Since CREVS and the emergency filtration units function to 
provide protection following a radiological accident the changes 
proposed to improve their performance cannot create a new or 
different kind of accident. Changes to the Bases to provide better 
information on determining CREVS and CCHE operability cannot create 
the possibility of a new or different kind of accident.
    3. Does not involve a significant reduction in a margin of 
safety.
    The proposed amendment does not involve a significant reduction 
in a margin of safety. Neither performance of periodic CCHE leak 
tests nor changes to the existing VFTP can create a reduction in the 
margin of safety. The changes to both of these programs will result 
in improved assurance that the CREVS and CCHE will perform as 
expected if required for operator protection. Changes to the Bases 
of the CREVS Technical Specification which clarify the conditions 
necessary for operability will improve understanding of the 
requirements for maintaining control room habitability, and will not 
create a reduction in the margin of safety. The existence of 
additional breaches in the CCHE for short periods of time does not 
significantly increase the risk of control room operator exposure to 
airborne radioactivity or toxic gas. There is no change in the risk 
to the public since the CCHE has no direct function in mitigating 
the offsite consequences of any evaluated accident. Any event that 
could create these exposures has an extremely low probability of 
occurrence, and while the potential for higher operator exposure 
exists if additional breaches are open, the short duration allowed 
would not significantly increase the risk of exposure. Therefore, 
for the reason stated above the existing margin of safety would not 
be reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied.
    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619

[[Page 64116]]

W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: September 30, 1998 (LAR-238).
    Description of amendment request: The proposed amendment will 
correct the reactor coolant system (RCS) leakage detection capability 
of the Reactor Building atmosphere gaseous radioactivity monitor 
described in the Improved Technical Specification Bases and the Final 
Safety Analysis Report (FSAR). These documents currently identify that 
the gaseous radioactivity monitor is capable of detecting a one gallon 
per minute (gpm) RCS leak within one hour. The licensee has determined 
that it would take approximately 14 hours for this instrument to detect 
a one gpm RCS leak using currently accepted assumptions. The capability 
of other monitors to detect a one gpm RCS leak within one hour is not 
affected by this change.
    The licensee cited several factors which contribute to the 
difficulty in reliably detecting RCS leakage increases of one gpm 
within one hour using a gaseous radioactivity monitor. These include 
the relatively long half-life of Xe-133 (primary nuclide of detection), 
fluctuations in background levels of radioactivity, the existence of 
minor RCS leaks, improved performance of nuclear fuel, and improved 
primary water chemistry control. Based on RCS radioactivity 
concentrations assumed in the Environmental Report, half-lives of the 
most abundant gaseous nuclides, and background radioactivity levels, 
the licensee indicated a one gpm leak can conservatively be detected in 
approximately 14 hours by the gaseous monitor. The licensee has 
determined that this change to the licensing basis is an unreviewed 
safety question.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No. The function of the RM-A6 gaseous radioactivity monitor is 
to detect leakage from the RCS that may develop as a result of a 
flaw in a pressure boundary component. The previously identified 
capability to detect a one gpm leak within one hour would have 
provided an earlier warning of a small RCS leak than the actual 
detection capability now identified. However, RCS loss of coolant 
accidents evaluated in the FSAR cover the full spectrum of break 
sizes up to and including a complete severance of the largest RCS 
piping. The results of these analyses demonstrate that the 
consequences of such leaks are acceptable.
    No other equipment relies on the capability of the RM-A6 gaseous 
monitor's ability to detect RCS leakage to perform its function. 
Likewise, no accident analyses rely on RCS leak detection for 
successful mitigation. Identifying the detector's actual capability 
to detect an RCS leak will not increase the probability of 
occurrence of an RCS leak. Detection time for an RCS leak was a 
consideration in granting a partial exemption to General Design 
Criterion 4. However, the capability of the RCS piping to resist 
propagation of a flaw from a leak into a break was based on material 
fracture analysis and material properties, not on the ability to 
detect low levels of leakage.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No. The function of the RM-A6 gaseous radioactivity monitor is 
to detect RCS leakage that may develop from a flaw in a pressure 
boundary component. The monitor is a passive component that provides 
an indication of possible leakage for further operator evaluation. 
Identifying that a longer response time is required for the monitor 
to detect a small leak will not create the possibility of a new or 
different kind of accident. Existing analyses for small and large 
break loss of coolant accidents provide an evaluation of the full 
spectrum of RCS break sizes.
    3. Involve a significant reduction in a margin of safety.
    No. The RM-A6 gaseous radioactivity monitor is included in plant 
technical specifications as one of two containment atmosphere RCS 
leak detection instruments required to be operable to satisfy a 
limiting condition for operation. If the RM-A6 particulate monitor 
is not operable, then the response time of the containment 
atmosphere monitor will be increased. RCS piping analyses have 
demonstrated that the propagation of a small primary loop leak into 
a pipe break would not occur rapidly. NRC acceptance of the 
applicable analyses included significant safety factors for the 
propagation of flaws into pipe breaks which were based on low 
probability stress combinations of normal plus safe shutdown 
earthquake loads. Considering the actual detection capability of the 
RM-A6 gaseous monitor and the existence of other diverse leak 
detection capabilities, detection of a leak in a relatively short 
period of time is anticipated. In the event an RCS leak developed 
into a pipe break, current accident analyses would bound the effects 
of the pipe break on and off site. Therefore, the possibility of 
increased time to detect an RCS leak does not represent a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: October 16, 1998 (LAR-229).
    Description of amendment request: The proposed amendment would 
change the Crystal River Unit 3 (CR-3) Final Safety Analysis Report 
(FSAR), Improved Technical Specifications (ITS) and ITS Bases to 
resolve an Unreviewed Safety Question (USQ). This USQ was created by 
changing the normal standby position of valves DHV-34 and DHV-35 (low 
pressure injection (LPI) pump suction valves from borated water storage 
tank) from normally open to normally closed. Maintaining these valves 
normally closed is necessary to ensure assumptions used in fire 
protection analyses remain valid. The proposed amendment would also add 
new ITS surveillance requirements for verifying on a periodic basis 
that the LPI system components and piping, and the building spray 
suction piping, are full of water.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Valves DHV-34 and DHV-35 are located in the suction lines 
between the borated water storage tank (BWST) and the low pressure 
injection (LPI) and building spray (BS) pumps. These valves are 
maintained normally closed, and are designed to automatically open 
upon receipt of a reactor coolant system (RCS) low-low pressure 
signal

[[Page 64117]]

of 500 psig or a reactor building (RB) high pressure signal of 4 
psig from the engineered safeguards actuation system (ESAS). The 
designed full stroke time of these valves is within the assumptions 
of the accident analyses performed for the specific design basis 
accidents that require the LPI and/or BS systems for accident 
mitigation. This is the original design basis for these valves. 
Therefore, the valves are fully capable of performing their intended 
safety functions while being maintained normally closed.
    The failure of one of these valves to open does not impact the 
mitigation of previously analyzed accidents that require the 
operation of the LPI and/or BS systems, and cannot increase the 
probability of these accidents occurring. No RCS or secondary system 
pressure boundaries are compromised, no release paths for 
radioactive materials are created, and no challenge to any safety 
limit or acceptance limit are created by maintaining these valves 
normally closed.
    A single, active failure causing one of these valves to fail to 
open upon demand would render one train of LPI and BS unavailable 
for accident mitigation. However, the accident analyses have already 
accounted for the possibility of only one train of LPI and BS being 
available, and the consequences of previously evaluated accidents 
would therefore remain unchanged.
    Undetected voiding in the LPI piping and components, and BS 
suction piping, is highly unlikely to occur. Based on the design and 
physical layout of the LPI system and BS system, and the monitoring 
of the systems performed on a periodic basis, any potential for LPI 
piping and components and BS suction piping voiding will be quickly 
and easily recognized and corrected. Therefore, since voiding is not 
likely to occur, the consequence of previously evaluated accidents 
would not be significantly increased by the proposed change.
    2. Create the possibility of a new or different kind of accident 
from previously evaluated accidents?
    Failure of either valves DHV-34 or DHV-35 to open upon demand on 
an ESAS signal will not create the possibility of a new or different 
kind of accident. The LPI system and BS system are maintained in a 
standby condition during normal plant operations, and automatically 
actuate only after an accident has occurred to mitigate the effects 
of the initiating accident. No RCS or secondary system pressure 
boundaries are compromised, no release paths for radioactive 
materials are created, and no challenges to any safety limit or 
acceptance limit are created by maintaining these valves normally 
closed. Additionally, the possibility of undetected voiding in the 
LPI piping and components, and BS suction piping, is not likely to 
occur by maintaining these valves normally closed. Therefore, 
maintaining valves DHV-34 and DHV-35 normally closed will not be an 
initiator of a new or different kind of accident from previously 
evaluated accidents.
    3. Involve a significant reduction in a margin of safety?
    Maintaining valves DHV-34 and DHV-35 normally closed will not 
create a reduction in the margin of safety. Maintaining valves DHV-
34 and DHV-35 normally closed will ensure the capability to safely 
shut down the reactor under certain postulated fire scenarios, but 
will result in an extremely small increase in the probability of 
failure of one train of LPI and BS to perform its safety functions. 
Based on use of the CR-3 Probabilistic Safety Analysis (PSA) model, 
and assuming the failure of either valve DHV-34 or DHV-35 to open, 
the impact on the core-damage frequency was estimated and determined 
to slightly increase from 7.38 E-6 to 7.41 E-6 per year. This 
increase (3 E-8 or 0.4%) is in the range considered acceptable in 
Regulatory Guide 1.174, ``An Approach for Using Probabilistic Risk 
Assessment in Risk-Informed Decisions on Plant-Specific Changes to 
the Current Licensing Basis,'' dated July 1998.
    Maintaining these valves normally closed will not result in 
undetected voiding in the LPI piping and components, and BS suction 
piping, as a result of performance of periodic pressure monitoring. 
If voiding occurs, the Improved Technical Specifications specify the 
actions required to restore the affected systems to operable status, 
including correcting the external leakage creating the observed 
pressure decay. Therefore, the proposed monitoring will ensure the 
margin of safety is not reduced.
    Based on these benefits and risks, there is no discernible 
change in the risk to the public in mitigating the offsite 
consequences of any evaluated accident since the failure of one 
train of LPI and/or BS for any reason is bounded by the assumptions 
of the accident analyses. Failure of valve DHV-34 or DHV-35 to open 
upon demand results in extremely low increases in the potential for 
reactor core damage. Therefore, the existing margin of safety will 
not be reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: October 15, 1998.
    Description of amendment request: The proposed amendment request 
would revise the TMI-1 Updated Final Safety Analysis Report (UFSAR) 
Chapter 14 postulated accident analysis radiological dose consequences 
resulting from application of revised atmospheric dispersion factors 
(X/Q) at the Technical Specification Section 5.1.1 defined exclusion 
area boundry (EAB) and low population zone (LPZ).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed amendment has no effect on 
structures, systems or components. More extensive and recent 
meteorological data have been utilized for atmospheric dispersion 
factor (X/Q) determination for both EAB and LPZ. An evaluation of 
the design basis accidents with revised EAB and LPZ X/Q values 
results in increases in UFSAR Chapter 14 EAB and LPZ dose 
consequences which remain well within the guidelines of 10 CFR Part 
100.
    Therefore, this activity does not involve a significant increase 
in the probability of occurrence or the consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. The proposed 
amendment has no impact on any plant structures, systems or 
components. The proposed change revises the atmospheric dispersion 
factors for EAB and LPZ used in the existing UFSAR Chapter 14 
accident analyses, based on more extensive meteorological data. 
These changes only effect the postulated dose consequences of 
currently analyzed accidents. Therefore, this activity does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment has no impact on structures, systems 
or components. The proposed revisions to the EAB and LPZ X/Q values 
are based on recent more extensive meteorological data and 
Regulatory Guide 1. 145 methods. The increased X/Q values provide a 
more accurate assessment of meteorological conditions which result 
in postulated dose consequences which remain well within the 
guidelines of 10 CFR Part 100. Therefore, this activity does not 
reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 64118]]

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

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

    Date of amendment request: October 19, 1998.
    Description of amendment request: The proposed Technical 
Specification change request would add operability and surveillance 
requirements for the remote shutdown system similar to those in NUREG-
1430, ``Standard Technical Specifications--Babcock and Wilcox Plants'' 
Section 3.3.18 entitled ``Remote Shutdown 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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed amendment adds operability and 
surveillance requirements for the existing TMI-1 remote shutdown 
system similar to those contained in NRC NUREG-1430, ``Standard 
Technical Specifications--Babcock & Wilcox Plants''. The addition of 
these requirements to Technical Specifications provides further 
assurance of remote shutdown system operability in the event that 
operators must place and maintain the unit in a safe shutdown 
condition from outside the control room. The function and operation 
of the remote shutdown system has not changed. Therefore, this 
activity has no affect on the probability of occurrence or 
consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. The proposed 
amendment has no impact on any plant structures, systems or 
components. The function and operation of the remote shutdown system 
has not changed. Therefore, this activity does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment provides additional assurance of 
remote shutdown system operability. The function and operation of 
the remote shutdown system has not changed. Therefore, this activity 
does not reduce the margin of safety.

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

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

    Date of amendment request: October 19, 1998.
    Description of amendment request: The proposed change to the TMI-1 
Technical Specification would revise the limit on reactor coolant 
system activity to a maximum allowable of 1.0 microcurie/gram dose 
equivalent I-131. The proposed revision provides an allowable reactor 
coolant system specific activity limit base on once-through steam 
generator (OTSG) inspection results performed each refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed amendment has no effect on 
structures, systems or components. The existing steam line break 
criteria are maintained. This change only accounts for radiological 
consequences resulting from a revised maximum allowable reactor 
coolant system (RCS) specific activity limit of 1.0 iCi/gm. 
The new radiological consequences of the revised MSLB accident, 
which also incorporate more conservative values for atmospheric 
dispersion, are below 10 CFR 100 limits and 10 CFR 50, Appendix A, 
GDC-19 limits for the control room. The use of revised atmospheric 
dispersion factors for other TMI-1 accident analysis is addressed in 
a separate license amendment request submittal. Therefore, this 
activity does not involve a significant increase in the probability 
of occurrence or the consequences of an accident previously 
evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. The proposed 
amendment has no impact on any plant structures, systems or 
components. OTSG tube structural integrity is maintained. Therefore, 
this activity does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment has no impact on structures, systems 
or components. OTSG tube structural integrity is maintained. The 
existing TMI-1 Technical Specification Section 3.1.4.1 Bases state 
that the limitations on the specific activity of the primary coolant 
ensure that the resulting 2-hour doses at the site boundary will be 
well within the Part 100 limit following associated design basis 
accidents postulated in conjunction with an assumed steady state 
primary-to-secondary steam generator tube leakage of 1.0 gpm. This 
margin of safety is preserved since resulting does consequences 
incorporating more conservative values for atmospheric dispersion 
remain well within the Part 100 limit. Therefore, this activity does 
not reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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, Pitman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

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

    Date of amendment request: October 23, 1998.
    Description of amendment request: The proposed amendment would 
allow implementation of a feedwater leakage control system to address 
leakage through the primary containment feedwater penetration isolation 
valves.

[[Page 64119]]

    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 implements a method of providing a 
qualified sealing system for the primary containment feedwater 
penetration isolation valves. This water sealing function, i.e., the 
FWLCS, constitutes a new operating mode of the Residual Heat Removal 
(RHR) system. The FWLCS introduces new piping that constitutes an 
extension of the reactor coolant system (RCS); however, such piping 
is designed to the same requirements as other RCS piping and as such 
introduces no significant increase in the probability of any 
accident previously evaluated. Notwithstanding, a postulated line 
break in any of the new FWLCS piping would not, by itself, introduce 
any new effects or consequences not already bounded by postulated 
line-break or LOCA events previously evaluated in the USAR. Since 
the proposed change does not affect any parameters or conditions 
that contribute to the initiation of any accidents previously 
evaluated, the proposed change cannot increase the probability of 
any accident previously evaluated.
    The proposed change potentially affects the leak-tight integrity 
of the primary containment designed to mitigate the consequences of 
a loss-of-coolant accident (LOCA). Once the FWLCS mode has been 
initiated and a water seal for the seating surfaces of the primary 
containment feedwater penetration isolation valves has been 
established (within one hour after the accident), post-LOCA primary 
containment atmosphere will be prohibited from leaking through the 
feedwater penetrations and thus bypassing the secondary containment.
    Calculations of post-accident DBA LOCA doses affected by this 
change use accepted ICRP 30 dose conversion factors and take credit 
for suppression pool scrubbing. Suppression pool scrubbing is 
effective in reducing iodine release but has no assumed effect on 
the removal of noble gases. Since the methodology and assumptions 
for scrubbing are acceptable to the NRC per the guidance in SRP 
Section 6.5.5 and the values for decontamination factors are 
conservative, considerable margin is preserved within the analysis. 
However, these calculations show increases in some of the previously 
evaluated post-accident doses when compared with dose calculations 
performed as part of the initial plant licensing basis. Although 
some of the newly calculated post-accident doses are larger than 
those that were previously approved, the increases remain small 
enough to be within the acceptance limits given in 10 CFR 50, 
Appendix A, GDC 19 and in 10 CFR 100.11.
    Since all of the newly calculated post-accident doses resulting 
from the proposed addition of a water sealing system for the 
feedwater primary containment penetration isolation valves are below 
the 10 CFR 50, Appendix A, GDC 19 and 10 CFR 100.11 acceptance 
limits, IP has concluded that the proposed change does not result in 
a significant increase in the consequences of an accident previously 
evaluated.
    2. The proposed change institutes a new operating mode of the 
RHR system (the FWLCS mode). When this mode is established, it will 
reduce primary containment atmosphere leakage to the environment in 
the event of a LOCA. Flow diverted from the RHR system to the FWLCS 
has been evaluated, and has been determined to have no adverse 
impact on the capability of the RHR system to perform its intended 
safety functions. Further, the additional piping added for the FWLCS 
is designed to appropriate requirements for the RCS, thus ensuring 
that RCS integrity is maintained per design. Sufficient isolation 
between the RCS and the RHR low-pressure piping will also be 
maintained per the FWLCS design. Thus, no safety functions are 
altered or impacted as a result of this change. Installing, 
operating, or testing the components that support the FWLCS mode 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. Because the USAR analysis already assumes leakage through 
the feedwater primary containment penetrations following a design 
basis LOCA, and the subject change does not affect the type of 
accident(s) that are postulated to occur, the proposed change does 
not present the possibility of an accident of a different type. 
Additionally, the change in dose analysis methodology does not 
create an accident or malfunction of a different type since it only 
involves the analysis of the effects of accidents or malfunctions 
previously evaluated in the USAR.
    Based on the above, IP has concluded that the proposed change 
will not create the possibility of a new or different kind of 
accident not previously evaluated.
    3. The margin of safety impacted by the proposed change involves 
the dose consequences of postulated accidents which are directly 
related to the primary containment leakage rate, specifically those 
consequences associated with dose attributable to leakage through 
the feedwater lines which are secondary containment bypass leakage 
paths.
    Although considerable conservatisms were included in the 
reanalysis, this reanalysis identified some dose values that 
increased above the previously licensed values as well as some dose 
values that decreased below the previously licensed values. However, 
all of the radiation dose consequences resulting from the proposed 
change will continue to be below the 10 CFR 50, Appendix A, GDC 19 
and 10 CFR 100.11 acceptance criteria.
    Except for providing a method of sealing the feedwater primary 
containment penetration isolation valves (and therefore the method 
of performing periodic leakage testing of these components) no other 
change in the method of primary containment leakage testing or 
secondary containment bypass leakage path testing is being proposed. 
All other primary and secondary containment bypass leakage testing 
will continue to be performed in accordance with existing Technical 
Specification requirements. Adequate programs are in place to ensure 
that proper maintenance and repairs are performed during the service 
life of the primary containment, systems and components penetrating 
the primary containment, and for all secondary containment bypass 
leakage paths.
    As a result, IP has concluded that the proposed change will not 
result in 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, IL 62525.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: September 28, 1998.
    Description of amendment request: The proposed amendment request 
would resolve an unreviewed safety question (USQ) and amend the 
operating license to allow manual override capability for the 
containment isolation actuation signal to reactor coolant system 
letdown isolation valves.
    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 modification does not change the probability of any 
accident previously evaluated since it does not change any mode of 
normal operation. Neither the accident signal (CIAS) nor the 
override feature is an initiator of an analyzed event. The 
consequences of an accident are also not changed significantly due 
to the fact that design and administrative controls ensure that 
previous accident analyses are bounding. The associated isolation 
valves will operate as they have in the past in response to an 
accident signal. There is no single failure that would prevent the 
letdown isolation function from occurring. The CIAS override feature 
can only be used if operators have verified that an UHE is the event 
which has taken place and safety functions are being met.

[[Page 64120]]

This ensures that no significant fuel failures will occur due to the 
event and the consequences of overriding CIAS will not adversely 
impact radiological conditions in the auxiliary building.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed modification does not create any failure mode which 
could impact the operation of the RCS or associated systems in a 
manner that would create a new or different kind of accident. With 
respect to the letdown isolation function, the plant will operate as 
it previously has and will respond the same way, automatically, to 
an accident signal. No new accidents have been identified.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The procedural restrictions associated with the use of the CIAS 
override feature will ensure that existing analyses addressing the 
consequences of an UHE will be bounding and that safety functions 
will be maintained as defined in EOPs. The radiological consequences 
of letdown restoration in the auxiliary building will be similar to 
normal operating conditions and will be bounded by that assumed in 
the EEQ analysis. RCS inventory and pressure control will be 
maintained within the established procedural limits.
    Letdown restoration capability already exists after ESF reset. 
The modification permits letdown restoration to occur earlier than 
it would previously have been possible.

    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.

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

    Date of amendment request: October 15, 1998.
    Description of amendment request: The proposed Technical 
Specification (TS) changes involve revising TS Section 3/4.10 to 
include a new Special Test Exception allowing the reactor to be 
considered in operational condition (OPCON) 4 (cold shutdown) during 
inservice leak or hydrostatic testing with a reactor coolant water 
temperature greater than 200 deg.F and less than or equal to 212 deg.F. 
This is an exception to certain OPCON 3 (hot shutdown) requirements, 
including primary containment. The proposed TS changes will permit 
unrestricted access to the primary containment for the performance of 
required inspections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed TS changes do not make any physical alterations or 
modifications to plant systems or equipment. The proposed TS changes 
will permit the performance of inservice leak or hydrostatic 
testing, with the reactor in OPERATIONAL CONDITION (OPCON) 4 (COLD 
SHUTDOWN) and the average reactor coolant temperature greater than 
200 deg.F and less than or equal to 212 deg.F. The probability of a 
leak in the reactor coolant pressure boundary during inservice leak 
or hydrostatic testing is not increased by considering the reactor 
in OPCON 4 with reactor coolant temperatures greater than 200 deg.F 
and less than or equal to 212 deg.F. The inservice leak and 
hydrostatic testing is performed water solid or near water solid. 
The stored energy in the reactor core will be very low and the 
potential for failed fuel and a subsequent increase in reactor 
coolant activity above TS limits is minimal. In addition, Secondary 
Containment will be operable and capable of handling airborne 
radioactivity from leaks that could occur during the performance of 
inservice leak or hydrostatic testing. Requiring the Secondary 
Containment to be operable will ensure that potential airborne 
radioactivity from leaks will be filtered through the Standby Gas 
Treatment System (SGTS), thereby limiting any radioactivity releases 
to the environment.
    In the event of a large primary system leak, the reactor vessel 
would rapidly depressurize allowing the low pressure Emergency Core 
Cooling System (ECCS) subsystems to operate. The capability of the 
systems that are required for OPCON 4 would be adequate to keep the 
core flooded under this condition. Small system leaks would be 
detected by leakage inspections before significant inventory loss 
has occurred. This is an integral part of the hydrostatic testing 
program.
    Therefore, the proposed TS changes will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes do not make any physical alterations or 
modifications to plant systems or equipment. The proposed TS changes 
do not adversely impact the operation of any plant equipment. 
Allowing the reactor to be considered in OPCON 4 during hydrostatic 
or inservice leak testing, with a reactor coolant temperature 
greater than 200 deg.F and less than or equal to 212 deg.F, is an 
exception to certain OPCON 3 (HOT SHUTDOWN) requirements, including 
primary containment integrity. The hydrostatic or inservice testing 
is performed water solid, or near water solid. The stored energy in 
the reactor core will be very low and the potential for failed fuel 
and a subsequent increase in coolant activity above TS limits is 
minimal. In addition, the Secondary Containment will be operable and 
capable of handling airborne radioactivity from leaks that could 
occur during the performance of hydrostatic or inservice leakage 
testing.
    The inservice leak or hydrostatic test conditions remain 
unchanged. The potential for a system leak remains unchanged since 
the reactor coolant system is designed for temperatures exceeding 
500 deg.F with similar pressures. There are no alterations of any 
plant systems or components that cope with the spectrum of 
accidents.
    Therefore, the proposed TS changes will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed TS changes do not make any physical alterations or 
modifications to plant systems or equipment. The proposed changes 
will permit the performance of inservice leak and hydrostatic 
testing with a reactor coolant temperature greater than 200 deg.F 
and less than or equal to 212 deg.F and the reactor in OPCON 4. 
Since the reactor vessel head will be in place, Secondary 
Containment integrity will be maintained, and all systems required 
in OPCON 4 will be operable in accordance with the applicable TS 
requirements. The proposed TS changes will not have any significant 
impact on any design basis accident or safety limit. The hydrostatic 
or inservice leak testing is performed water solid, or near water 
solid. The stored energy in the reactor core is very low and the 
potential for failed fuel and a subsequent increase in coolant 
activity would be minimal. In the event of a large primary system 
leak, the reactor pressure vessel would rapidly depressurize and the 
low pressure ECCS subsystems would function as designed to maintain 
adequate reactor core coverage. This would ensure that the fuel 
would not exceed peak clad temperature limits.
    Also, requiring Secondary Containment integrity will assure that 
potential airborne radioactive material can be filtered through the 
SGTS. This will assure that any offsite doses remain well within the 
limits of 10 CFR 100 guidelines. Small system leaks would be 
detected by inspections before significant inventory loss could 
occur.
    Therefore, this proposed TS change will not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 64121]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: October 19, 1998.
    Description of amendment request: The proposed amendment would 
eliminate restrictions imposed by Technical Specification (TS) 3.0.4 
for the Filtration, Recirculation and Ventilation System (FRVS) during 
fuel movement and core alteration activities. Specifically, TS Limiting 
Conditions for Operation (LCOs) 3.6.5.3.1 and 3.6.5.3.2 would each be 
revised to add a note stating that the provisions of TS 3.0.4 are not 
applicable for initiation of handling of irradiated fuel in the 
secondary containment and core alterations provided that the plant is 
in Operational Condition 5, with reactor water level equal to or 
greater than 22 feet 2 inches.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS change does not involve any physical changes to 
plant structures, systems or components (SSC). FRVS will continue to 
function as designed. FRVS is an Engineered Safety Feature (ESF) 
designed to mitigate the consequences of an accident, and therefore, 
can not contribute to the initiation of any accident. For refueling 
accidents, the current design basis analysis of FRVS credits only 
the iodine removal capability of the FRVS ventilation unit and 
neglects the considerable iodine removal capability of the FRVS 
recirculation units. In addition, this proposed TS change will not 
increase the probability of occurrence of a malfunction of any plant 
equipment important to safety, since the time limits imposed by the 
current FRVS LCO Action Statements are not affected by these 
proposed changes. The proposed changes merely allow entry into the 
FRVS LCO Action Statement in order to support refueling activities.
    Therefore, the proposed TS changes, which would permit the 
initiation of core alterations and handling of irradiated fuel with 
only one operable FRVS ventilation unit and four operable FRVS 
recirculation units for a limited seven day period under specific 
refueling conditions, would not result in the increase of the 
consequences of an accident previously evaluated.
    Therefore, the proposed TS change does not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes do not involve any physical changes to 
plant SSC. The design and operation of the FRVS is not changed from 
that currently described in the [Updated Final Safety Analysis 
Report] UFSAR. FRVS will continue to function as designed to 
mitigate the consequences of an accident. No changes of any kind are 
being made to FRVS, or its support or supported systems. Deleting 
the restrictions imposed by TS 3.0.4 as proposed in this TS change 
request eliminates a compliance restriction imposed by the current 
TS. Since the current TS already provide a seven day period to 
perform refueling activities with inoperable FRVS ventilation and 
recirculation units, the proposed changes would not introduce plant 
operation in a configuration that is not already permitted in the 
TS. Therefore, there is no possibility that implementing this 
proposed TS change would create a different type of malfunction to 
the FRVS than any previously evaluated. In addition, the proposed TS 
changes do not alter the conclusions described in the UFSAR 
regarding operation of FRVS.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed TS change involves the elimination of TS 3.0.4 
restrictions imposed on the FRVS LCO. The TS 3.0.4 requirements 
impose an unnecessary challenge to performing refueling activities 
when the FRVS LCO Action Statements already sufficiently define the 
remedial measures to be taken. The time limits imposed by the 
current FRVS LCO Action Statements are not affected by these 
proposed changes. The FRVS LCO will retain sufficient configuration 
controls to appropriately maintain the capability of FRVS to 
mitigate design basis refueling accidents, no new FRVS 
configurations will be permitted by the proposed changes, and there 
will be no reduction in any margin of safety resulting from this 
proposed TS change. Therefore, the proposed TS change does not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: Robert A. Capra.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station 
(VCSNS), Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: September 18, 1998.
    Description of amendment request: The proposed amendment would 
revise the VCSNS Technical Specifications (TS) to address the Best 
Estimate Analyzer for Core Operations--Nuclear (BEACON) core power 
distribution monitoring and support system. The BEACON system provides 
continuous core monitoring capabilities to augment the flux mapping 
system when rated thermal power (RTP) is greater than 25%. The proposed 
amendment would also make editorial changes to TS 3.3.3.2 and 4.3.3.2.c 
to delete the reference to Fxy.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change allows the Power Distribution Monitoring 
System (PDMS) to be used for measuring power distribution limits 
when Thermal Power is greater than 25% RTP. This includes relocating 
manufacturing and measurement uncertainty values from the Technical 
Specification to the COLR [core operating limit report]. Also 
included in this change is the addition of a new specification and 
bases section for the Power Distribution Monitoring System (PDMS). 
The Technical Specification Power Distribution Limits are not being 
changed; only the method in which they are measured is being 
changed. The probability of an accident is not significantly 
increased. The measurement of power distribution limits and the 
location of manufacturing and measurement uncertainty values are not 
initiators of any analyzed event. The change will not affect the 
consequences of any analyzed event. The power distribution limits 
will still be measured and verified to be within limits as required 
by the current Technical Specification Surveillance. The cycle-
specific core operating limits, although not in Technical 
Specifications, will be followed in the operation of VCSNS. The 
actions as required by current Technical Specifications, when or if 
limits are exceeded are not being

[[Page 64122]]

changed. This change will not significantly affect the assumptions 
relative to the mitigation of accidents.
    Each accident analysis addressed in the VCSNS Final Safety 
Analysis Report will be examined with respect to changes in cycle-
dependent parameters, which are obtained from application of the 
NRC-approved reload design methodologies, to ensure that the 
transient evaluation of new reloads are bounded by previously 
accepted analyses. This examination, which will be performed per 
requirements of 10 CFR 50.59, ensures that future reloads will not 
involve an increase in the probability or consequences of an 
accident previously evaluated.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change allows the Power Distribution Monitoring 
System (PDMS) to be used for measuring power distribution limits 
when Thermal Power is greater than 25% RTP. This includes relocating 
manufacturing and measurement uncertainty values from the Technical 
Specification to the COLR. Also included is the addition of a new 
specification and bases section for the Power Distribution 
Monitoring System. No safety-related equipment, safety function, or 
plant operation will be altered as a result of this proposed change. 
No hardware is being added to the plant as part of the change. The 
cycle specific variables are calculated using the NRC-approved 
methods and submitted to the NRC to allow the Staff to continue to 
trend the values of these limits. The Technical Specifications will 
continue to require operation within the required core operating 
limits and appropriate actions will be taken when or if limits are 
exceeded. The change will not introduce any new accident initiators. 
Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed change allows the Power Distribution Monitoring 
System (PDMS) to be used for measuring power distribution limits 
when Thermal Power is greater than 25% RTP. The margin of safety 
presently provided by current Technical Specifications remains 
unchanged. Only the method in which the power distribution 
measurements are obtained is being changed. This method is verified 
by Westinghouse, and reviewed and approved by the NRC. Appropriate 
measures exist to control the values of the manufacturing and 
measurement uncertainties. The proposed amendment continues to 
require operation within the core limits, as obtained from NRC-
approved reload design methodologies. Appropriate actions required 
to be taken when or if limits are violated remain unchanged.
    Future changes to measurement and manufacturing uncertainties 
located in the current Technical Specification will be evaluated per 
the requirements of 10 CFR 50.59. Since the 10 CFR 50.59 process 
does not allow any reduction in the margin of safety, prior NRC 
approval is required prior to a reduction in the margin of safety. 
If the evaluation of the changes [does] not result in [an] 
unreviewed safety question, prior NRC approval will not be required. 
Additionally, the VCSNS Technical Specifications require that all 
revisions of the plant COLR be submitted to the NRC upon issuance.
    Therefore, the 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: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: October 12, 1998.
    Description of amendment request: The proposed amendments would 
revise Section 6, ``Administrative Controls,'' of the current Units 1 
and 2 Technical Specifications (TS) to recognize additional management 
positions associated with the steam generator replacement project and 
providing them the ability to approve procedures regarding this 
project.
    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 significantly increase the 
probability or consequences of an accident previously evaluated in 
the FSAR [Final Safety Analysis Report]. The proposed changes have 
no impact on the probability of an accident. The change being 
proposed is administrative in nature and involves no physical 
alteration of the plant or changes to setpoints or operating 
parameters. The change will provide an appropriate level of review 
and approval of procedures related to the FNP steam generator 
replacement without impacting the operational attention of the 
current on-site plant management. There is no change in the FNP 
design basis as a result of this change and, as a result, does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    (2) The proposed changes to the TS do not increase the 
possibility of a new or different kind of accident than any already 
evaluated in the FSAR. No new limiting single failure or accident 
scenario has been created or identified due to the proposed changes. 
Safety-related systems will continue to perform as designed. The 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    (3) The proposed changes do not involve a significant reduction 
in the margin of safety. Adding individuals with the appropriate 
knowledge base to the list of individuals who can approve 
procedures, which may affect plant nuclear safety, is administrative 
in nature. There is no impact on the accident analyses. The training 
and experience requirements for the newly designated management 
positions are similar to those requirements for other FNP management 
positions. Therefore the established level of procedure review and 
approval is not adversely impacted. In addition, these changes allow 
FNP management to remain focused on plant operations. Thus the 
proposed changes do 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: 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.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: September 28, 1998.
    Description of amendment request: The proposed amendment would 
modify the requirements applicable when one or more trains of fuel 
handling building exhaust air or control room makeup and cleanup 
filtration are inoperable, and eliminate the need to enter Technical 
Specification 3.0.3 when multiple trains of these systems are 
inoperable. In addition, the proposed changes would align the actuating 
instrumentation and logic system required actions with those that are 
applicable to the systems. Finally,

[[Page 64123]]

an administrative change is proposed to remove a footnote that is no 
longer applicable to the facility.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes consist of:
    (a) Assuring that the Specifications define consistent allowed 
outage times when the same safety function is addressed in multiple 
Specifications,
    (b) Allowing a system to remain inoperable when appropriately 
restrictive administrative controls are placed on operations that 
could result in a challenge to the safety function of the system,
    (c) Providing an appropriately short Allowed Outage Time for 
inoperability needed to permit required maintenance and testing that 
affects all trains of a system,
    (d) Redefining system operability and associated actions in a 
manner consistent with the system design and function,
    (e) Aligning a system to the actuated condition on the loss of 
an actuation channel,
    (f) Using consistent terminology throughout the Specifications.
    The proposed changes do not represent significant increases in 
the probability or consequences of an accident because:
    (a) The alignment of the action times between actuating system 
and actuated system operability requirements do not affect the 
probability or consequences since inoperability of the actuated 
system has the same effect as inoperability of the actuating system. 
Since the changes proposed to the actuating system action times will 
reflect those of the actuated system action times, no change to the 
allowed outage time applicable to the safety function addressed and 
fulfilled by both, will occur.
    (b) Administrative controls to prevent the conduct of operations 
that could lead to a challenge to the safety function of the system 
when the actuation system is inoperable, assures that the design 
bases functions of the system will not be challenged. Therefore, the 
probability or consequences of an event previously identified have 
not been significantly changed.
    (c) Allowing up to 12 hours to recover from the inoperability of 
all three trains of Control Room Ventilation or two or more trains 
of Fuel Handling Building HVAC does not represent a significant 
change to the probability of an accident because the inoperability 
of these ventilation systems are not identified as precursors to a 
design basis event. The low likelihood of a design basis accident 
during the limited period of allowed inoperability of these systems 
does not represent a significant increase in the consequences of an 
accident.
    (d) The redefinition of plant operability requirements into 
functional trains rather than individual components does not affect 
the required system functional operability. Therefore, this change 
does not represent an increase in the probability or consequences of 
an accident previously identified.
    (e) The alignment of the Control Room Ventilation System to the 
same configuration it would be placed in from an actuation of the 
inoperable radiation monitoring channel places the system in the 
design condition. This alignment would result in maintaining the 
control room envelope pressurized and increases the protection 
afforded to the operators.
    (f) The change in terminology does not change any requirements 
or actions in the Specification. Therefore this change does not 
represent an increase in the probability or consequences of any 
accident previously evaluated.
    Based on the above discussion, the individual changes do not 
represent an increase in the probability or consequences of any 
accident previously evaluated.
    In addition to the changes proposed to controls over Control 
Room Ventilation, Fuel Handling Building HVAC, and associated 
actuation logic, an administrative change is proposed to remove the 
footnote at the bottom of page 3/4 7-20. Since the footnote no 
longer has meaning or relevance to the operation of the facility, 
its removal does not increase the probability or consequences of any 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes make the existing Specifications internally 
consistent, manually align a system to the actuated position, 
provide an alternative measure that assures [that] a safety function 
which is unavailable is not required to [be] perform[ed], provide an 
extended period of allowance for all trains of a system to be 
inoperable, and redefines system operability to reflect its 
functional design. The proposed changes do not introduce any new 
equipment into the plant or significantly alter the manner in which 
existing equipment will be operated. The systems affected by the 
proposed changes are not identified as contributing causal factors 
in design basis accidents, their function is to assist in mitigation 
of accidents postulated to occur. Since the proposed changes do not 
allow activities that are significantly different from those 
presently allowed, no possibility exists for a new or different kind 
of accident from those previously evaluated.
    In addition to the changes proposed to controls over reactivity 
changes, an administrative change is proposed to remove the footnote 
at the bottom of page 3/4 7-20. Since the footnote does not perform 
any function and will never again apply to plant operations, its 
removal cannot create the possibility of a new or different kind of 
accident from those previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed changes do not involve a significant reduction in a 
margin of safety because the ability of the Fuel Handling Building 
HVAC and Control Room Ventilation Systems will be maintained. The 
margin of safety is defined by the ability of the systems to limit 
the release of radioactive materials and limit exposures to 
operators respectively following a postulated design basis accident. 
The only aspect of the proposed change that can be postulated to 
have any effect on a margin of safety is the proposed allowance for 
all trains of Control Room Ventilation or Fuel Handling Building 
HVAC to be inoperable for a limited period. The low probability of a 
design basis event that would require the system to perform its 
safety function during the limited period allowed by the proposed 
action assures that the change does not involve a significant change 
in a margin of safety. Therefore, the proposed changes do not 
significantly affect these operating restrictions and the margin of 
safety which support the ability to make and maintain the reactor in 
a safe shutdown and limit the release of radioactive material is not 
affected.
    In addition to the changes described above, an administrative 
change is proposed to remove the footnote at the bottom of page      
3/4 7-20. Since the footnote is no longer applicable to the 
facility, its removal cannot 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 standards of 10 CFR 50.92 
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, NW, Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: September 29, 1998.
    Description of amendment request: The licensee proposes to use a 
revised methodology to calculate mass and energy release following a 
postulated large-break loss-of-coolant accident. The amendment request 
also included proposed changes to the Updated Final Safety Analysis 
Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or

[[Page 64124]]

consequences of an accident previously evaluated.
    This proposal updates the design large break loss of coolant 
accident (LBLOCA) analysis and methodology described in the UFSAR to 
support replacement of Westinghouse Model E Original Steam 
Generators (OSG) with Westinghouse Delta-94 Replacement Steam 
Generators (RSG).
    A safety analysis has been performed, including evaluation of 
existing analyses and performance of bounding or confirming 
calculations, to determine effects of the proposed changes.
    Analysis of mass and energy releases and resultant containment 
pressure and temperature response for the RSG concluded a small 
reduction in peak pressure and temperature for the RSG compared to 
the OSG. Thus, the proposed amendment does not involve a significant 
increase in the probability of an accident previously evaluated.
    Changes to the LBLOCA model caused by installation of the RSGs 
and associated changes in analysis methodology result in no change 
in radiological consequence as delineated in 10 CFR 100 and the 
Standard Review Plan (NUREG-0800). Consequences of this design basis 
accident have not increased.
    Thus, changes in the LBLOCA design basis event analysis 
associated with replacement of OSGs with RSGs 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.
    This proposal updates the design basis large break loss of 
coolant accident (LBLOCA) analysis and methodology described in the 
Updated Final Safety Analysis Report (UFSAR) to support replacement 
of OSGs with RSGs.
    Fit, form, and design function of RSG equipment is not 
significantly changed from OSG equipment. Analyses of LBLOCA mass 
and energy releases and resultant containment system response 
indicates that performance with RSGs remains within the existing 
design limits. Thus, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    A safety analysis has been performed, including evaluations of 
existing analyses and performance of bounding and/or confirming 
calculations, to determine the effect of the proposed changes. 
Results of these analyses demonstrate that the proposed license 
amendment and operation of STP Units with Delta-94 steam generators 
installed will not produce post-accident Containment pressures or 
temperatures exceeding existing Technical Specification limits. 
Consequently, there are no effects on dose analyses due to design 
basis LBLOCA performance of the RSGs. Radiological consequences of 
the postulated accident did not change, and all results remain 
within the acceptance criteria of 10 CFR 100 and the Standard Review 
Plan (NUREG-0800).
    Thus, the change in LBLOCA analysis results and methodology 
descriptions in the UFSAR associated with replacement of Model E 
steam generators with Delta-94 steam generators 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 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, NW, Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: September 30, 1998.
    Description of amendment request: The proposed amendment would 
change the Updated Final Safety Analysis Report and revise the offsite 
dose licensing basis to account for operation of the existing steam 
generators at reduced feedwater inlet temperatures, and to account for 
operation of the new replacement steam generators. The calculated 
offsite dose consequences would increase for the main steamline break, 
reactor coolant pump shaft seizure, and rod cluster control assembly 
ejection accidents. The proposed increases in offsite doses are minimal 
and all doses remain below the dose limits for their respective 
accidents, as specified by 10 CFR Part 100 and the Standard Review Plan 
(NUREG-0800).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This document updates the facilities' radiological design basis, 
as described in the Updated Final Safety Analysis Report, to address 
both a reduction in allowed nominal feedwater temperature for Model 
E steam generators from 440  deg.F to 420  deg.F and the replacement 
of Model E steam generators with Delta-94 steam generators. 
Therefore, these changes do not change the probability of an 
accident previously evaluated.
    A safety analysis has been performed, including evaluations of 
existing analyses and performance of bounding and/or confirming 
calculations, to determine the impact of the proposed changes. 
Effects on the dose analyses due to the accompanying physical 
changes to the plant are slight. However, some improvements were 
made to the analytical models used in the analyses. These 
improvements were responsible for the majority of the increase in 
offsite doses. While the radiological consequences of some 
postulated accidents increased, all results remain within the 
acceptance criteria, as defined in 10 CFR 100 and the Standard 
Review Plan (NUREG-0800).
    The radiological consequences of the postulated accidents remain 
within their respective acceptance criteria with the use of the 
revised analysis methodologies. Therefore, the change to allow 
operation of the Model E steam generators at a reduced feedwater 
temperature of 420 deg.F and the replacement of Model E steam 
generators with Delta-94 steam generators 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.
    This document updates the facilities' radiological design basis, 
as described in the Updated Final Safety Analysis Report, to address 
both a reduction in allowed nominal feedwater temperature for Model 
E steam generators from 440  deg.F to 420  deg.F and the replacement 
of Model E steam generators with Delta-94 steam generators. Since 
the proposed changes to the Updated Final Safety Analysis Report are 
analytical in nature, the changes do 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.
    A safety analysis has been performed, including evaluations of 
existing analyses and performance of bounding and/or confirming 
calculations, to determine the impact of the proposed changes. 
Effects on the dose analyses due to the accompanying physical 
changes to the plant are slight. However, some improvements were 
made to the analytical models used in the analyses. These 
improvements were responsible for the majority of the increase in 
offsite doses. While the radiological consequences of some 
postulated accidents increased, all results remain within the 
acceptance criteria, as delineated in 10 CFR 100 and the Standard 
Review Plan (NUREG-0800), for the respective accidents.
    The radiological consequences of the postulated accidents remain 
within their respective acceptance criteria with the use of the 
revised analysis methodologies. Therefore, the change to allow 
operation of the Model E steam generators at a reduced feedwater 
temperature of 420  deg.F and the replacement of Model E steam 
generators with Delta-94 steam generators do not

[[Page 64125]]

involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: October 27, 1998.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 3/4.8.2.3, ``Electrical 
Power Systems--DC Distribution--Operating,'' and the associated bases. 
The surveillance requirements for battery testing would be revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit Number 1, in accordance with these changes 
would:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions, or assumptions are adversely affected by the proposed 
changes to station battery testing methodology and frequency.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are adversely affected by the proposed changes in 
station battery testing methodology and frequency. The proposed 
changes do not alter the source term, containment isolation, or 
allowable radiological releases. The proposed changes are consistent 
with the most recent IEEE Standard 450-1995, ``IEEE Recommended 
Practice for Maintenance, Testing, and Replacement of Vented Lead-
Acid Batteries for Stationary Applications,'' and the ``Improved 
Standard Technical Specifications for Babcock and Wilcox Plants,'' 
NUREG-1430, Revision 1.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes. The batteries are not an initiator or contributor to the 
initiation of an accident. No new accident scenarios, transient 
precursors, failure mechanisms, or limiting faults are introduced as 
a result of the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because the proposed TS changes do not significantly reduce or 
adversely affect the capabilities of any plant structures, systems 
or components. These changes increase the effectiveness and 
frequency of the battery tests being performed. Therefore, there is 
not 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 Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: October 27, 1998.
    Description of amendment request: The proposed amendment would 
relocate a Technical Specification (TS) surveillance requirement from 
TS Section 3/4.6.5.1, ``Shield Building-Emergency Ventilation System'' 
to TS Section 3/4.6.5.2, ``Shield Building Integrity.'' Administrative 
and bases changes would also be made.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit Number 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiator is 
affected by the proposed changes to the Technical Specifications 
(TS) Index; TS Definition 1.6, ``Shield Building Integrity''; TS 3/
4.6.5.1, ``Emergency Ventilation System''; TS 3/4.6.5.2, ``Shield 
Building Integrity''; TS Bases 3/4.6.5.1, ``Emergency Ventilation 
System''; or TS Bases 3/4.6.5.2, ``Shield Building Integrity.''
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions and 
assumptions are significantly affected by the above proposed 
changes. The proposed change to relocate existing TS Surveillance 
Requirement (SR) 4.6.5.1.d.4 to TS 3/4.6.5.2, and the subsequent 
application of the Limiting Condition for Operation (LCO) of TS 3/
4.6.5.2 should the Emergency Ventilation System (EVS) be unable to 
produce the required negative pressure in the annulus space due to 
an opening in the ventilation boundary, would allow 24 hours to 
restore the capability of maintaining the required negative pressure 
in the annulus. The current SR 4.6.5.1.d.4 and associated TS LCO 
3.6.5.1 would require entry into TS 3.0.3, thereby allowing only one 
hour for restoration before commencing plant shutdown. The allowed 
outage time of 24 hours is reasonable considering the limited 
leakage design of containment and the low likelihood of a Design 
Basis Accident (DBA) occurring during this time period. The proposed 
changes are consistent with the guidance of the ``Improved Standard 
Technical Specifications for Combustion Engineering Plants,'' NUREG-
1432, Revision 1 and the ``Improved Standard Technical 
Specifications for Westinghouse Plants,'' NUREG-1431, Revision 1. 
The ``Improved Standard Technical Specifications for Babcock and 
Wilcox Plants,'' NUREG-1430, Revision 1 does not contain guidance 
for shield building integrity because the DBNPS is the only Babcock 
and Wilcox-type plant with the containment vessel/annulus space/
shield building design. The proposed changes do not alter the 
drawdown capability of the EVS. Since the likelihood of a DBA 
occurring during this 24 hour period is low and the containment is 
of a low leakage design, the radiological consequences of a 
previously evaluated accident are not significantly increased. The 
proposed changes do not alter the source term, containment isolation 
or allowable radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes. No new accident scenarios, transient precursors, failure 
mechanisms, or limiting failures are introduced as a result of the 
proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because the proposed TS changes do not significantly reduce or 
significantly adversely affect the capabilities of any plant 
structures, systems or

[[Page 64126]]

components. The capability of the shield building/EVS to respond 
when necessary and to maintain a negative pressure will not be 
significantly changed by these proposed TS changes. Accordingly, 
there is not 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 Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: October 28, 1998.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 6, ``Administrative 
Controls.'' Several requirements would be modified and/or relocated to 
the Updated Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit Number 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions or assumptions are affected by the proposed changes to 
Technical Specification (TS) 6.5.1.6 ``[Station Review Board] 
Responsibilities''; TS 6.8.4.d, ``Radioactive Effluents Control 
Program''; TS 6.10, ``Record Retention''; TS 6.11, ``Radiation 
Protection Program''; TS 6.12, ``High Radiation Area''; and TS 6.15, 
``Offsite Dose Calculation Manual (ODCM).''
    These changes proposed to TS 6.5.1.6, TS 6.8.4.d, TS 6.10, and 
TS 6.15 are administrative changes that improve or update the 
content of TS Section 6.0, ``Administrative Controls.''
    The change proposed to TS 6.11 would relocate its content to the 
DBNPS Updated Safety Analysis Report, thereby removing it from the 
TS consistent with the NRC's NUREG-1430, Revision 1, ``Improved 
Standard Technical Specifications for Babcock and Wilcox Plants.''
    The changes proposed to TS 6.12 are based upon the current 
revision to 10 CFR Part 20, ``Standards for Protection Against 
Radiation,'' as published in the Federal Register, dated August 15, 
1994, and TS approved by the NRC for the San Onofre Nuclear 
Generating Station Units 2 and 3 in Operating License Amendments 127 
and 116, respectively. The changes to TS 6.12 also provide for the 
use of alternative methods for controlling access to high radiation 
areas and state-of-the-art radiation protection monitoring methods, 
such as closed circuit television and telemetry.
    Under the proposed changes, the TS would continue to satisfy the 
applicable requirements of 10 CFR 50.36(c)(5).
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected by the proposed changes. As described 
above, these changes are administrative changes or are proposed 
pursuant to the current revision to 10 CFR Part 20, ``Standards for 
Protection Against Radiation.'' The proposed changes do not alter 
the source term, containment isolation, or allowable releases. The 
proposed changes, therefore, will not increase the radiological 
consequences of a previously evaluated accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes. As described above, these changes are administrative 
changes or are proposed pursuant to the current revision to 10 CFR 
Part 20, ``Standards for Protection Against Radiation.''
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes are administrative changes or are 
proposed pursuant to the current 10 CFR Part 20 requirements. These 
proposed changes do not reduce or adversely affect the capabilities 
of any plant structures, systems or components.

    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 Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: October 27, 1998.
    Description of amendment request: This proposed amendment request 
would modify Technical Specification (TS) 4.2.b, ``Steam Generator 
Tubes,'' to redefine the plugging limits for the Westinghouse Hybrid 
Expansion Joint sleeves (HEJs) and Westinghouse Laser Welded Sleeves 
(LWSs). Additional administrative changes are also proposed. The 
proposed changes are as follows:
    1. TS 4.2.b.3.c.1 would be changed to correct an oversight from a 
previous amendment. The current TS 4.2.b.2.c.1 makes reference to TS 
3.4.a.1.C. This reference is no longer valid because TS 3.4.a.1.C 
became TS 3.4.d as a result of TS Amendment 123. This change corrects 
an oversight from a previous amendment and is administrative.
    2. TS 4.2.b.4.a would be revised to specify the updated revision of 
WCAP-14685 and the addendum to WCAP-13088.
    3. TS 4.2.b.4.b would be revised to specify the corrected value for 
the plugging limit of the Westinghouse mechanical HEJ sleeves. The 
plugging limit would change from 24 percent to 23 percent or more 
sleeve wall degradation.
    4. TS 4.2.b.4.e would be revised to specify the corrected value for 
the plugging limit of Westinghouse laser welded sleeves. The plugging 
limit would change from 25 percent to 23 percent or more sleeve wall 
degradation.
    The associated bases pages for TS Section 4.2 would also be 
modified to reflect the above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The analysis of change in plugging limits was performed in 
accordance with RG 1.121 and ASME B&PV Code and, therefore, all 
required safety factors are met. The plugging limit or allowed 
degraded wall thickness value is not used in any accident analyses; 
therefore, this change has no significant

[[Page 64127]]

effect on any previously evaluated accidents. The change does not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    Because the maximum primary-to-secondary differential pressure 
parameter has changed, the conventional analysis techniques 
originally used to qualify the required weld width under predicted 
the shear stress in the LWS and LWR [laser weld repair] of HEJ 
welds. Consequently, a verification program using experimental 
analysis, as allowed by Section III of the ASME B&PV Code, was 
performed to show that the weld remains in compliance with the ASME 
B&PV Code. Using a different analysis technique to verify that the 
previously approved weld width for LWS and LWR of HEJs is still 
accurate does not increase the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    Recalculating the allowable sleeve wall degradation and plugging 
limits and verifying the acceptability of the 0.015 inch weld width 
ensures that currently approved conditions are maintained. Requiring 
tubes to be plugged at a smaller sleeve wall degradation value does 
not result in any new or different conditions which could create a 
new or different accident.
    Verification of the currently approved weld width using a 
different analysis technique does not have a physical effect on any 
plant equipment or operating parameters and, therefore, can not 
create a new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    These TS changes are being made to ensure that the current 
margins of safety are maintained. This is accomplished by reducing 
the allowable sleeve wall degradation and plugging limit. Verifying 
the required, minimum weld width by an allowed, alternate analysis 
technique, as described by ASME B&PV Code, ensures that an adequate 
margin of safety is maintained and there is not a significant 
reduction in the margin of safety.
    The minor administrative changes do not impact the technical 
content or implementation of the TS and therefore can not create a 
significant hazard.

    The changes to the steam generator tube and sleeve plugging limits 
are necessary because of an increase in the normal operating 
differential pressure between the primary and secondary coolant 
systems. The differential pressure was increased as a result of the 
effects of extensive tube plugging on primary to secondary heat 
transfer. Since, per Regulatory Guide 1.121, the safety factor for 
mimimum acceptable wall thickness for steam generator tubes is based on 
normal operating pressures, it was found necessary to recalculate the 
plugging limits.
    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 Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: October 23, 1998.
    Description of amendment request: The amendment would revise 
Technical Specification 3.5.1, ``Emergency Core Cooling Systems--
Accumulators,'' to increase the allowed outage time for the 
accumulators from 1 hour to 24 hours if an accumulator is inoperable 
for reasons other than not meeting its boron concentration 
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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The overall protection system performance will remain within the 
bounds of the accident analyses documented in Chapter 15 of the 
Updated Safety Analysis Report (USAR), WCAP-10961-P, and WCAP-11883, 
since no hardware changes are proposed. The impact of the increase 
in the accumulator AOT on core damage frequency for all the cases 
evaluated in WCAP-15049 is within the acceptance limit of 1.0E-06/yr 
for a total plant CDF less than 1.0E-03/yr. The incremental 
conditional core damage probabilities calculated in WCAP-15049 for 
the accumulator AOT increase meet the criterion of 5E-07 in 
Regulatory Guide DG-1065 for all cases except those that are based 
on design basis success criteria. As indicated in WCAP-15049, design 
basis accumulator success criteria are not considered necessary to 
mitigate large break LOCA events, and was only included in the WCAP-
15049 evaluation as a worst case data point. In addition, WCAP-15049 
states that the NRC has indicated that an ICCDP greater than 5E-07 
does not necessarily mean the change is unacceptable.
    The safety injection accumulators are credited in Section 15.6.5 
of the Updated Safety Analysis Report for large and small break 
LOCA. There will be no effect on these analyses, or any other 
accident analysis, since the analysis assumptions are unaffected and 
remain the same as discussed in Section 15.6.5. Design basis 
accidents are not assumed to occur during allowed outage times 
covered by the Technical Specifications. As such, the ECCS 
Evaluation Model equipment availability assumptions made in Section 
15.6.5 remain valid.
    The safety injection accumulators will continue to function in a 
manner consistent with the above analysis assumptions and the plant 
design basis. As such, there will be no degradation in the 
performance of, nor an increase in the number of challenges to, 
equipment assumed to function during an accident situation.
    The proposed technical specification change does not involve any 
hardware changes nor does it affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, engineered safety feature (ESF) actuation setpoints, 
accident mitigation capabilities, accident analysis assumptions or 
inputs. Therefore, this change will not increase the probability of 
an accident or malfunction.
    The corresponding increase in CDF due to the proposed change to 
increase the AOT of the accumulators from one hour to 24 hours is 
not significant. Pursuant to the guidance in Section 3.5 of NEI 96-
07, Revision 0, ``Guidelines for 10 CFR 50.59 Safety Evaluations,'' 
the proposed increase in AOT does not ``degrade below the design 
basis the performance of a safety system assumed to function in the 
accident analysis,'' nor does it ``increase challenges to safety 
systems assumed to function in the accident analysis such that 
safety system performance is degraded below the design basis without 
compensating effects.''
    Therefore, it is concluded that this change does not increase 
the probability of occurrence of a malfunction of equipment 
important to safety.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This change does not involve any change to the installed plant 
systems or the overall operating philosophy of WCGS.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. As described in Section 9.1 of the WCAP-
15049 evaluation, the plant design will not be changed with this 
proposed Technical Specification AOT increase. All safety systems 
still function in the same manner and there is no additional 
reliance on additional systems or procedures. The proposed 
accumulator AOT increase has a very small impact on core damage 
frequency. The WCAP-15049 evaluation demonstrates that the small 
increase in risk due to increasing the accumulator AOT is within the 
acceptance criteria provided in Draft Regulatory Guide DG-1065. No 
new accident or transients can be introduced with

[[Page 64128]]

the requested change and the likelihood of an accident or transient 
is not impacted.
    The malfunction of safety related equipment, assumed to be 
operable in the accident analyses, would not be caused as a result 
of the proposed technical specification change. No new failure mode 
has been created and no new equipment performance burdens are 
imposed. Therefore, the possibility of a new or different 
malfunction of safety related equipment is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety. There will be no change to the Departure from 
Nucleate Boiling Ratio (DNBR) Correlation Limit, the design DNBR 
limits, or the safety analysis DNBR limits discussed in Bases 
Section 2.1.1.
    The basis for the accumulator LCO, as discussed in Bases Section 
3/4.5.1, is to ensure that a sufficient volume of borated water will 
be immediately forced into the core through each of the cold legs in 
the event the RCS pressure falls below the pressure of the 
accumulators, thereby providing the initial cooling mechanism during 
large RCS pipe ruptures. As described in Section 9.2 of the WCAP-
15049 evaluation, the proposed change will allow plant operation in 
a configuration outside the design basis for up to 24 hours, instead 
of 1 hour, before being required to begin shutdown. The impact of 
this on plant risk was evaluated and found to be very small. That 
is, increasing the time the accumulators will be unavailable to 
respond to a large LOCA event, assuming design basis accumulator 
success criteria is necessary to mitigate the event, has a very 
small impact on plant risk. Since the frequency of a design basis 
large LOCA (a large LOCA with loss of offsite power) would be 
significantly lower than the large LOCA frequency of the WCAP-15049 
evaluation, the impact of increasing the accumulator AOT from 1 hour 
to 24 hours on plant risk due to a design basis large LOCA would be 
significantly less than the plant risk increase presented in the 
WCAP-15049 evaluation. It is therefore concluded that the proposed 
change does not involve a significant reduction in the margin of 
safety as described in Technical Specification Bases Section 3/
4.5.1.
    As discussed previously, the performance of the accumulators 
will remain within the assumptions used in the large and small break 
LOCA analyses, as presented in USAR Section 15.6.5. Also, there will 
be no effect on the manner in which safety limits or limiting safety 
system settings are determined nor will there be any effect on those 
plant systems necessary to assure the accomplishment of protection 
functions.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: William H. Bateman.

Yankee Atomic Electric Company, Docket No. 50-029, Yankee Nuclear Power 
Station, Franklin County, Massachusetts

    Date of amendment request: October 15, 1998.
    Description of amendment request: The licensee proposes to extend 
the interval of submission of Effluent and Waste Disposal Reports from 
semi-annual to annual pursuant to 10 CFR 50.36a(a)(2). This action 
would require a change to Technical Specification (TS) 6.8.2.b, a 
reporting requirement, and textual changes in other parts of the TS to 
make the change consistent throughout.
    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 changes to the Yankee Nuclear Power Station Defueled 
Technical Specifications proposed above are administrative in 
nature. The proposed changes are consistent with the revised 10 CFR 
50.36a, ``Technical specifications on effluents from nuclear power 
reactors,'' which require the submittal of one Radioactive Effluent 
Release Report per year. Furthermore, the NRC has already concluded 
in issuing the 10 CFR 50.36a rule change that implementation of the 
proposed technical specifications changes would not result in a 
reduction to the public health and safety or common defense and 
security.
    As such, the changes:
    (1) Will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The administrative nature of the changes do not affect the 
operation of YNPS in the permanently defueled condition. 
Furthermore, the changes do not result in a change to the plant 
design, configuration, or operating procedures. Because the physical 
plant is not affected, and the only change is the frequency with 
which reports are submitted to the NRC, the probability of an 
accident previously evaluated is not increased and the radiological 
consequences of an accident previously evaluated are not increased.
    (2) Will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The changes described do not modify the design, configuration, 
or operating procedures for any plant systems or components. The 
accident analyses for the facility are not affected by the proposed 
changes. The changes do not introduce any new failure mechanisms. 
Therefore, the changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    (3) Will not involve a significant reduction in the margin of 
safety.
    The changes described are administrative in nature. The changes 
do not modify the design, configuration, or operating procedures for 
any plant systems or components. The changes do not affect the 
facility's accident analyses. Radioactive effluent release limits 
remain unchanged. The submittal of reports to the NRC is an 
administrative function and is not included in the bases of any 
Technical Specifications to define or establish a margin of safety. 
Therefore, the proposed changes do not reduce the margin of safety 
as defined in the bases of any 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: Greenfield Community College, 
1 College Drive, Greenfield, Massachusetts 01301.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One 
International Place, Boston, Massachusetts 02110-2624.
    NRC Project Director: Seymour H. Weiss.

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

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

[[Page 64129]]

Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of amendment request: October 16, 1998.
    Description of amendment request: The amendment would change 
Technical Specification (TS) 3.3.1, ``Reactor Protective System 
Instrumentation--Operating'' and TS 3.3.2, ``Reactor Protective System 
Instrumentation Shutdown'' to clarify an inconsistency between TS 
wording and the design basis as described in the TS Bases and the 
Updated Final Safety Analysis Report.
    Date of publication of individual notice in Federal Register: 
October 27, 1998 (63 FR 57320).
    Expiration date of individual notice: November 27, 1998.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: August 17, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification 5.2.2.f regarding the senior reactor operator licensing 
requirement for the operations manager.
    Date of issuance: November 4, 1998.
    Effective date: November 4, 1998.
    Amendment Nos.: 204 and 234.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revise the facility's Technical Specifications.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48258) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 4, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendment: September 19, 1998.
    Brief description of amendment: This amendment revises Section 
5.4.8 of the Oyster Creek Nuclear Generating Station Updated Final 
Safety Analysis Report (UFSAR) such that it incorporates the use of a 
freeze seal as a temporary part of the reactor coolant pressure 
boundary.
    Date of Issuance: November 4, 1998.
    Effective date: November 4, 1998.
    Amendment No. 201.
    Facility Operating License No. DPR-16. Amendment revised the UFSAR.
    Date of initial notice in Federal Register: September 30, 1998 (63 
FR 52307).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated November 4, 1998. .
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

    Date of application for amendment: July 23, 1998, as supplemented 
September 25, 1998. The September 25, 1998, supplement did not change 
the initial proposed no significant hazards consideration 
determination.
    Brief description of amendment: The amendment establishes that the 
existing Safety Limit Minimum Critical Power Ratio in Technical 
Specification 2.1.A is applicable for Cycle 17.
    Date of Issuance: November 5, 1998.
    Effective date: November 5, 1998, to be implemented within 30 days.
    Amendment No.: 202.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 26, 1998 (63 FR 
45525).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated November 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: January 29, 1997, as 
supplemented February 11, 12, March 7, 10, 11, 19, 20, April 29, June 
30, and July 10,1997, June 20, June 22, July 24, September 15, and 
October 1, 1998.
    Brief description of amendments: The amendments change the design 
basis of the cooling water system emergency intake line flow capacity. 
The changes also reclassify the intake canal for use during a seismic 
event, which would be an additional source of cooling water available 
during a design-basis earthquake. The amendments also reflect the 
completion of license conditions that were implemented as part of 
interim amendments 128/120 dated March 25, 1997, to reflect 
compensatory measures taken by Northern States Power until a 
seismically qualified emergency cooling water source could be provided.
    Date of issuance: November 4, 1998.
    Effective date: November 4, 1998, with full implementation within 
30

[[Page 64130]]

days. Implementation of the USAR update shall be no later than June 1, 
1999, as stated in License Condition 3.
    Amendment Nos.: 140 and 131.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the licenses.
    Date of initial notice in Federal Register: October 1, 1998 (63 FR 
52772). The October 1, 1998, submittal provided revised USAR pages 
reflecting the change to the cooling water system emergency intake 
design bases. This information was within the scope of the October 1, 
1998, Federal Register notice and did not change the staff's initial 
proposed no significant hazards considerations determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 4, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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 amendments: March 6, 1998.
    Brief description of amendments: The proposed changes modify the 
technical specifications (TS) to eliminate reference to shutdown 
cooling (SDC) system isolation bypass valve inverters. This allows the 
licensee to replace the inverters with transfer switches.
    Date of issuance: October 26, 1998.
    Effective date: October 26, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 2--143; Unit 3--134.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50939).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713.

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

    Date of amendments request: December 30, 1997, as supplemented by 
letter dated April 9, 1998.
    Brief Description of amendments: The amendments change the 
Technical Specifications to revise the surveillance requirements for 
the Auxiliary Building and Service Water Building batteries to remove 
the existing 1.75 volt minimum individual cell voltage associated with 
the ``service test'' acceptance criterion and replace it with a 
reference to the battery load profile specified in the Final Safety 
Analysis Report, Section 8.3.2.
    Date of issuance: November 3, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1--139; Unit 2--131.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17234). The April 9, 1998, letter provided clarifying information that 
did not change the scope of the December 30, 1997, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 3, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

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

    Date of application for amendment: August 6, 1998.
    Brief description of amendment: Change Technical Specifications 
(TS) Surveillance and Bases Sections 3.3.2, ``ESFAS Instrumentation,'' 
and 3.7.5, ``AFW System'' to clarify the intent of the surveillance 
testing requirements for the turbine driven auxiliary feedwater pump, 
which is consistent with the wording and intent of the Westinghouse 
Improved TS.
    Date of issuance: October 26, 1998.
    Effective date: October 26, 1998.
    Amendment No.: 13.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50941).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 26, 1998.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.

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

    Date of application for amendments: November 6, 1996, as 
supplemented April 15, July 14, and October 16, 1998. The supplemental 
submittals contained clarifying information only, and did not change 
the initial no significant hazards consideration determination.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Sections 3.4.1.4, 4.4.1.4, 3.4.1.5, 
3.4.1.6, 4.4.1.6.1, 4.4.1.6.2, 4.4.1.6.3, 3/4.4.2 and 3/4.4.3 for Unit 
1, and 3.4.1.4, 4.4.1.4, 3.4.1.5, 3/4.4, 3.4.1.6, 4.4.1.6.1, 4.4.1.6.2, 
and 4.4.1.6.3 for Unit 2, modifying the requirements for isolated loop 
startup to permit filling of a drained isolated loop via backfill from 
the reactor coolant system through partially opened loop stop valves.
    Date of issuance: October 30, 1998.
    Effective date: October 30, 1998.
    Amendment Nos.: 215 and 196.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64396).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

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

[[Page 64131]]

amendments. The Commission has determined for each of these amendments 
that the application for the amendment complies with the standards and 
requirements of the Atomic Energy Act of 1954, as amended (the Act), 
and the Commission's rules and regulations. The Commission has made 
appropriate findings as required by the Act and the Commission's rules 
and regulations in 10 CFR Chapter I, which are set forth in the license 
amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By December 18, 1998, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these

[[Page 64132]]

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

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: October 23, 1998, as 
supplemented October 26, 1998.
    Brief description of amendments: The amendments clarify the 
conditions that constitute operable Individual Rod Position Indication 
(IRPI) system channels, provide for an allowed out of service time for 
inoperable IRPI indicator channels, and provide compensatory measures 
to be taken when any channel is determined to be inoperable.
    Date of issuance: October 30, 1998.
    Effective date: October 30, 1998.
    Amendment Nos.: 139 and 130.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
October 30, 1998.
    Attorney for licensee: J.E. Silberg, Esquire, Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    NRC Project Director: Cynthia A. Carpenter.

    Dated at Rockville, Maryland, this 10th day of November 1998.

    For the Nuclear Regulatory Commission.
William H. Bateman,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-30691 Filed 11-17-98; 8:45 am]
BILLING CODE 7590-01-P