[Federal Register Volume 59, Number 234 (Wednesday, December 7, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-29925]


[[Page Unknown]]

[Federal Register: December 7, 1994]


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NUCLEAR REGULATORY COMMISSION
 

Biweekly Notice

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 14, 1994, through November 25, 
1994. The last biweekly notice was published on November 23, 1994 (59 
FR 60377).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By January 6, 1995, 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, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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 amendments request: October 25, 1994
    Description of amendments request: The proposed change would delete 
the remainder of Appendix B, Environmental Technical Specifications, 
including section 2/3.3.1, Water Level in the Discharge Canal, and 
Section 2/3.4, Meteorology.
    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:
    Hydraulic - Water level in the Discharge Canal
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The deletion of the discharge canal 
specification of 4.5 [plus or minus] 1 ft mean sea level (msl) with 
daily monitoring still leaves the operating restrictions delineated 
in Updated Final Safety Analysis Report (UFSAR) Section 2.4.8.3.3 
(4.5 [plus or minus] 2 ft msl) and the National Pollutant Discharge 
Elimination System (NPDES) permit requirements to minimize the 
impact of the discharge canal on the local groundwater supply. As 
stated in this UFSAR section, the effect on the local ground water 
regime will be minimal within this band. Level recorders in the 
control room facilitate the continued monitoring the discharge canal 
level in excess of the Appendix B Environment Technical 
Specification (ETS) listed daily surveillance. This change in no way 
affects the design or operation of equipment that could initiate or 
mitigate any accident previously evaluated.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The discharge canal level is an environmental concern 
with the effects of a spill of radioactive liquids (UFSAR Section 
2.4.12.3) being a path to the intake canal, due to the areas natural 
gradient. With the estimated travel time for the liquid to reach the 
canal (intake) at 60 years, and the large flow rate the degree of 
dilution is such that this does not pose a threat to local wells. 
The amendment would not affect the operation or design of any plant 
equipment; therefore, no new credible accidents are created. In 
addition, the proposed amendment would not affect the capability of 
the response systems to mitigate the consequences of any accident 
previously evaluated; therefore, no new or different accident would 
result from this change.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety. The existence of these ETS does not 
provide a margin of safety related to the nuclear operation of the 
site. No safety limits are affected by this change. Therefore, this 
amendment would not result in a reduction in any margin of safety.
    Meteorology
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The deletion of the meteorology specification 
still leaves the program delineated in UFSAR Section 2.3.3, Onsite 
Meteorological Measurements Program, and reporting/recording of the 
hourly meteorological data required to support Technical 
Specification 6.9.1.10.a, Semiannual Radioactive Effluent Release 
Report. This program is based on the meteorological monitoring 
program described in Regulatory Guide 1.23, and NUREG-0654. While 
the existing ETS 30 day reporting requirement for extended out-of-
service time and shiftly manual acquisition of data during batch or 
accidental releases are not otherwise covered, the program does 
contain the Regulatory Guide 1.23 reference to 90% data recovery and 
a backup phone line is available for data retrieval. This proposed 
amendment in no way affects the design or operation of equipment 
that could initiate or mitigate any accident previously evaluated.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The meteorological monitoring program is specified in 
UFSAR Section 2.3.3. The amendment would not affect the operation or 
design of any plant equipment; therefore, no new credible accidents 
are created. In addition, the proposed amendment would not affect 
the capability of the response systems to mitigate the consequences 
of any accident previously evaluated nor would the amendment reduce 
the effectiveness of the Emergency Response Plan; therefore, no new 
or different accident would result from this change.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety. The meteorological ETS does not 
provide any additional margin of safety related to the operation of 
a nuclear plant. The meteorological program established in the UFSAR 
covers the requirements stated in Appendix E to 10 CFR 50 by 
providing meteorological systems adequate for determining the 
magnitude of, and for continuously assessing the impact of, the 
release of radioactive materials to the environment. The 
meteorological instrumentation is used to measure environmental 
parameters which may affect distribution of fission products and 
gases following a Design Basis Accident (DBA); however, it is not a 
primary success path for the mitigation of a DBA. No safety limits 
are affected by this change. Therefore, this amendment would not 
result in a reduction in any margin of safety.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

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 amendments request: October 28, 1994
    Description of amendments request: The proposed changes would 
revise the Technical Specifications to increase the surveillance test 
intervals and allowable out-of-service times for selected 
instrumentation addressed in Section 3/4.3. The proposed changes would 
permit specified channel functional tests to be conducted quarterly 
rather than weekly or monthly. Specifically, the proposed changes would 
revise the surveillance test intervals and allowable out-of-service 
times for the reactor protection system instrumentation, isolation 
actuation instrumentation, emergency core cooling system actuation 
instrumentation, control rod withdrawal block instumentation, control 
room emergency ventilation system instrumentation, anticipated 
transient without scram - recirculation pump trip system 
instrumentation, end-of-cycle recirculation pump trip system 
instrumentation, and reactor core isolation cooling system actuation 
instrumentation, in accordance with NRC-approved General Electric 
Company Licensing Topical Reports and NUREG-1433, Standard Technical 
Specifications, General Electric Plants, BWR/4.''
    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 Brunswick Steam Electric Plant, Units 1 and 2, 
in accordance with the proposed amendment, would not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The generic Licensing Topical Report, NEDC-30851P-A, assessed 
the impact of changing RPS surveillance test intervals (STIs) and 
allowable out-of-service times (AOTs) on the RPS failure frequency, 
the scram frequency and equipment cycling. Specifically, Section 
5.7.4, ``Significant Hazards Assessment'' of NEDC-30851P-1 states:
    ``Fewer challenges to the safeguards system, due to less 
frequent testing of the RPS, conservatively results in a decrease of 
approximately one percent in core damage frequency''. This decrease 
is based upon the following:
    * Based on the plant-specific experience presented in Appendix 
J, the estimated reduction in scram frequency (0.3 scrams/yr) 
represents a 1 to 2 percent decrease in core damage frequency based 
on the BWR plant specific Probabilistic Risk Assessments (PRAs) 
listed in Table 5-8.
    * The increase in core damage frequency due to less frequent 
testing is less than one percent. This increase is even lower (less 
than 0.01 percent) when the changes resulting from the 
implementation of the Anticipated Transients Without Scram (ATWS) 
rule are considered. Therefore, this increase is more than offset by 
the decrease in CDF due to fewer scrams.
    * The effect of reducing unnecessary cycles on RPS equipment, 
although not easily quantifiable also results in a decrease in core 
damage frequency.
    * The overall impact on core damage frequency of the changes in 
allowable out-of-service time is negligible.''
    From this generic analysis, the BWR Owners' Group concluded and 
CP&L concurs that the proposed changes do not significantly increase 
the probability or consequences of an accident previously evaluated, 
since the increase in probability of a a scram failure due to RPS 
unavailablity is insignificant. The overall probability of an 
accident is decreased as the time the RPS instrumentation logic 
operates undisturbed is increased, resulting in fewer inadvertent 
scrams during testing and repair. The proprietary plant-specific 
analysis contained in this submittal (Enclosure 6) demonstrates 
that, although BSEP Units 1 and 2 differ from the generic plant 
analyzed in LTR NEDC-30851P-A, the net effect of the plant-specific 
differences does not alter the generic conclusions.
    The generic Licensing Topical Reports, NEDC-30851P-A, Supplement 
2 and NEDC 31677P-A, assessed the impact of changing STIs and AOTs 
for BWR Isolation Instrumentation. Section 4.0, ``Summary of 
Results,'' of NEDC-30851P-A, Supplement 2 states:
    ``The results indicate that the effects on probability of 
failure to initiate isolation are very small and the effects on 
probability or frequency of failure to isolate are negligible in 
nearly every case. In addition, the results indicate that increasing 
the AOT to 24 hours for tests and repairs has a negligible effect on 
the probability of failure of the isolation function. These combined 
with changes to the testing intervals and allowable out-of service 
times for RPS and ECCS instrumentation provide a net improvement to 
plant safety and operations.''
    and Section 5.6, ``Assessment of Net Effect of Changes,'' of 
NEDC-31677P-A states:
    ``A reduction in core damage frequency (CDF) of at least as much 
as estimated in the ECCS instrumentation analysis can be expected 
when the isolation actuation instrumentation STIs are changed from 
one month to three months. The chief contributor to this reduction 
is the channel functional tests for the MSIVs. Inadvertent closure 
of the MSIVs will cause an unnecessary plant scram. This reduction 
in CDF more than compensates for any small incremental increase (10% 
or 1.0E-07/year) in calculated isolation function failure frequency 
when the STI is extended to three months.''
    From this generic analysis, the BWR Owners' Group concluded and 
CP&L concurs that the proposed changes do not significantly increase 
the consequences of an accident previously evaluated, since the 
increase in probability of an isolation failure due to isolation 
instrumentation unavailability is insignificant. For those 
parameters common to RPS, the overall probability of an accident is 
actually decreased as the time the RPS instrumentation logic 
operates undisturbed is increased, resulting in less inadvertent 
scrams during testing and repair. The plant-specific evaluation 
provided with this submittal (Enclosure 8) demonstrates that the 
conclusions of the generic analyses are applicable to BSEP Units 1 
and 2.
    The generic Licensing Topical Report, NEDC-30936P-A (Parts 1 and 
2), assessed the impact of changing STIs and AOTs for all BWR ECCS 
Actuation Instrumentation. Section 4.0, ``Technical Assessment of 
Changes,'' of NEDC-30936P-A (Part 2) states:
    ``The results indicate an insignificant (less than 5E-7 per 
year) increase in water injection function failure frequency when 
STIs are increased from 31 days to 92 days, AOTs for repair of the 
ECCS actuation instrumentation are increased from one hour to 24 
hours, and AOTs for surveillance testing are increased from two to 
six hours. For all four BWR models the increase represents less than 
4% increase in failure frequency. However, when other factors which 
influence the overall plant safety are considered, the net result is 
judged to be an improvement in plant safety.''
    From this generic analysis, the BWR Owners' Group concluded and 
CP&L concurs that the proposed changes do not significantly increase 
the probability or consequences of an accident previously evaluated, 
since the increase in probability of a water injection failure due 
to ECCS instrumentation unavailability is insignificant and the net 
result is judged to be an improvement in plant safety. The plant-
specific analysis contained in this submittal (Enclosure 7) 
demonstrates that, although BSEP Units 1 and 2 differ from the 
generic model analyzed in LTR NEDC-30936P-A, the net effect of the 
plant-specific differences does not alter the generic conclusions. 
The generic Licensing Topical Reports, NEDC-30851P-A, Supplement 1, 
and GENE-770-06-1-A assessed the impact of changing Control Rod 
Block STIs and AOTs on Rod Block failure frequency. GENE-770-06-1-A 
also assessed the impact of changing STIs and AOTs on ATWS-RPT and 
EOC-RPT failure frequency. Section 5 (Brookhaven National 
Laboratory's Technical Evaluation Report - Attachment 2 to the NRC 
SER) of NEDC-30851P-A, Supplement 1 states:
    1``The BWR Owners' Group proposed changes to the Technical 
Specifications concerning the test requirements for BWR control rod 
block instrumentation. The changes consist of increasing the 
surveillance test intervals from one to three months. These test 
interval extensions are consistent with the already approved changes 
to STIs for the Reactor Protection System. The technical analysis 
reviewed and verified as documented herein indicates that there will 
be no significant changes in the availability of the control rod 
block function if these changes are implemented. In addition, there 
will be a negligible impact on the plant core melt frequency due to 
the decreased testing.''
    and Section 2.0, ``Summary'' of GENE-770-06-1-A states:
    ``Technical bases are provided for selected proposed changes to 
the instrumentation STIs and AOTs that were identified in the BWROG 
Improved BWR Technical Specification activity. These STI and AOT 
changes are consistent with approved changes to the RPS, ECCS, and 
isolation actuation instrumentation. These proposed changes do not 
result in a degradation to overall plant safety.''
    Based on the generic analysis in NEDC-30851P-A, Supplement 1, 
the BWR Owners' Group concluded and CP&L concurs that the proposed 
changes to Control Rod Withdrawal Block instrumentation do not 
significantly increase the probability or consequences of an 
accident previously evaluated. Also, based on the generic assessment 
in GENE-770-06-1-A, the BWR Owners' Group concluded and CP&L concurs 
that the proposed changes to the ATWS-RPT and EOC-RPT 
instrumentation do not significantly increase the probability or 
consequences of an accident previously evaluated.
    Bases contained in GE Topical Report GENE-770-06-2P-A, assessed 
the impact of changing STIs and AOTs on BWR RCIC failure frequency. 
Section 2.0, ``Summary'' of GENE-770-06-2P-A states:
    ``The STI and AOT changes to the RCIC actuation instrumentation 
are justified based on their small effect on the water injection 
function unavailability and consistency with comparable changes to 
actuation instrumentation for the other ECCS subsystems.''
    On this basis, the BWR Owners' Group concluded and CP&L concurs 
that the proposed changes to RCIC instrumentation do not 
significantly increase the probability or consequences of an 
accident previously evaluated.
    2) Operation of Brunswick Steam Electric Plant, Units 1 and 2, 
in accordance with the proposed amendment, would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes do not alter the physical characteristics 
or function of any plant systems or components and they do not 
introduce any new mode of operation. Therefore, system and component 
performance would not be challenged in a manner that could create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3) Operation of Brunswick Steam Electric Plant, Units 1 and 2, 
in accordance with the proposed amendment, would not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed and approved the generic studies 
contained in the LTRs and has concurred with the BWR Owners' Group 
that the proposed changes do not significantly affect the 
probability of failure or availability of the affected Instrument 
Systems. The proposed changes to AOTs provide realistic times to 
complete the required actions without increasing the overall 
instrument failure frequency. Likewise, the extended STIs do not 
result in significant changes in the probability of instrument 
failure. Furthermore, the proposed changes will reduce the 
probability of test-induced plant transients and equipment failures. 
Finally, instrument setpoint drift will remain within present 
tolerances, thereby assuring that the margin of safety, as 
demonstrated by applicable safety analyses, remains unchanged. 
Therefore, it is concluded that the proposed changes would not 
result in a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

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

    Date of amendment request: November 4, 1994
    Description of amendment request: The requested amendment will 
change the testing frequency of the turbine overspeed protection valves 
from monthly to quarterly.
    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 requested change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The requested change will have no influence on the 
probability or consequences of an accident previously evaluated. The 
accident of concern to this requested change is a turbine overspeed 
with missile generation impacting safety related components or 
structures. The evaluation in WCAP-11525 shows that the probability 
of a missile ejection incident will not be affected with the 
requested frequency reduction to the turbine overspeed protection 
valve surveillance test. There is no change to the consequences of 
the event as the postulated accident event is unchanged. 
Accordingly, the requested change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The requested change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The change affects the test interval for the turbine 
overspeed protection valves and does not change the design, 
operation, or failure modes of the valves and other components in 
the turbine overspeed protection system. Therefore, the requested 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The requested change does not involve a significant reduction 
in the margin of safety. The probability of turbine overspeed with 
an extension of the testing interval has been determined to be 
within applicable acceptance criteria. The change does not affect 
the design, operation, or failure modes of the valves or other 
components in the turbine overspeed protection system. Accordingly, 
the requested change will 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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

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

    Date of amendment request: November 7, 1994
    Description of amendment request: The proposed amendment would 
allow an increase to the allowable nominal fuel enrichment from 4.2 to 
5.0 weight percent Uranium-235 (w/o U-235). The changes include: (1) 
increasing the allowable storage enrichment in Region 1 and allowing 
the use of Integral Fuel Burnable Absorbers (IFBAs) for reactivity 
equivalencing, (2) revising the Region 2 discharge burnup curve to 
include nominal fuel enrichments up to 5.0 w/o U-235, and (3) making 
editorial 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:
    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to Section 5 of Technical Specifications do 
not affect any accident initiators or precursors and do not change 
or alter the design assumptions for the systems or components used 
to mitigate the consequences of an accident. The fuel enrichment 
increase will not affect reactor operation or the core design 
methods. The physical characteristics of the fuel assemblies are not 
changed, and fuel assembly movement will continue to be controlled 
by approved fuel handling procedures. Reload core designs will 
continue to be performed on a cycle by cycle bases as part of the 
reload safety evaluation process, using NRC approved codes and 
methods. Each reload design is evaluated to confirm that the cycle 
core design adheres to the limits that exist in the accident 
analyses and Technical Specifications to ensure that reactor 
operation is acceptable.
    The proposed changes are consistent with the analysis performed 
in the ``Criticality Analysis of Byron and Braidwood Station Fuel 
Storage Racks.'' The analysis was revised in June 1994 to include 
boraflex gaps and shrinkage. The revised analysis is provided in the 
proposed Technical Specification amendment. The analysis methodology 
has been previously accepted by the NRC and is consistent with the 
appropriate standards to establish the Keff limit for storage 
racks and to calculate the maximum Keff. The reanalysis 
addresses the most limiting postulated accident of a misloaded fuel 
assembly and has shown that having at least 300 ppm of soluble boron 
offsets any positive reactivity impacts for any of the postulated 
accidents. The concentration of boron in the spent fuel pool water, 
which is administratively controlled, is sufficient to maintain 
Keff less than or equal to 0.95. The analysis is bounding for a 
dropped fuel assembly on top of a rack or between rack modules, loss 
of cooling systems, and reduction the fuel pool temperature to less 
than 50 deg.F. The proposed changes do not impact any other accident 
previously evaluated in the [Updated Final Safety Analysis Report] 
UFSAR. There is no postulated accident that could cause reactivity 
to increase beyond the analyzed conditions in the spent fuel racks.
    There is no impact on the ability of the Spent Fuel Pool cooling 
system to maintain the bulk pool temperature within limits. The 
UFSAR analysis performed to calculate the maximum fuel cladding 
temperature and spent fuel pool cooling include assumptions which 
bound the use of more highly enriched fuel assemblies. Although fuel 
enrichment is not a specific assumption in any of these analyses, 
the heat load of a typical core offload may change with higher 
enrichments. The average burnup of the offload will be increased 
since few assemblies will be used per cycle; however, the new heat 
load will continue to be [bound] by the UFSAR analysis because the 
spent fuel pool racks have been analyzed for a total core offload 
with all fuel assemblies having 4.5 years of operating time.
    The radiological consequences analysis continues to bound the 
licensed fuel burnup and enrichment at Byron and Braidwood stations. 
The radiological consequences analysis results are a function of the 
core inventory of radioactive isotopes. Since the maximum fuel 
burnup limits and fuel peaking factors will not be exceeded, the 
assumed fission product inventory will remain valid; therefore, the 
limits of 10 CFR [Part] 100 continue to be met. Additionally, Byron 
and Braidwood addressed the issue of the impact on the radiation 
levels at the pool surface to the worker during non-accident 
conditions. These conditions are not changed as [a] result of this 
submittal, because the average fuel assembly burnup limit (isotopic 
inventory) and maximum power produced in each fuel assembly will not 
be changed by the increased fuel enrichment.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not affect the design or operation of 
any system, structure, or component in the plant. There are no 
changes to parameters governing plant operation; no new or different 
type of equipment will be installed. Each reactor core design will 
continue to meet all design requirements; operation of the core will 
not be affected. No modifications to the spent fuel pool are being 
pursued and the fuel parameters used in the analysis remain 
bounding. The method and manner in which the fuel will be stored in 
the spent fuel pool has not changed. The proposed changes ensure 
that 17X17 (Optimized Fuel Assembly, VANTAGE 5, VANTAGE +, and 
PERFORMANCE +) fuel assemblies can be safely stored, maintaining a 
Keff less than or equal to 0.95 under full water density 
conditions, in both Regions 1 and 2 of the spent fuel pool. All 
design criteria and criticality acceptance criteria continue to be 
met. The reanalysis addresses the most limiting postulated accident 
(misloaded fuel assembly) and has shown that having at least 300 ppm 
of soluble boron offsets any positive reactivity impacts for any of 
the postulated accidents. The level of boron in the spent fuel pool 
water, which is administratively controlled, is sufficient to 
maintain Keff less than or equal to 0.95. The reanalysis to 
increase the storage enrichment of fuel in Regions 1 and 2 of the 
spent fuel pool does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Additionally, approval of this amendment will not create a new 
accident with regards to the new fuel storage vault which is 
designed to handle the increased enrichment. The Byron and Braidwood 
new fuel vaults were previously analyzed using NRC accepted 
criticality analysis methodology in June 1989. This analysis was 
performed to increase the storage enrichment of the New Fuel Vault 
to 5.0 w/o U-235. The New Fuel Vault analysis was submitted to the 
NRC and is the current licensing basis.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the margin of safety for any 
Technical Specification. All reactor design criteria will continue 
to be met. The methodologies used in the accident analyses have been 
accepted previously by the NRC and all criticality acceptance 
criteria have been met under all assumed conditions (normal and 
accident). The design basis for preventing criticality outside the 
reactor is that, including uncertainties, there is a 95 percent 
probability at a 95 percent confidence level that the effective 
neutron multiplication factor, Keff, of the fuel assembly array 
will be less than 0.95 as recommended by ANSI 57.2-1983 and OT 
Position Paper for Review and Acceptance of Spent Fuel Storage and 
Handling Applications, dated April 14, 1978. The analyses for both 
Regions 1 and 2 fuel storage were verified to meet the above design 
basis.
    The criticality analysis for Regions 1 and 2 has been revised to 
allow for storage of fuel assemblies with enrichments up to 5.0 w/o 
U-235. The proposed Technical Specification changes include those 
changes necessary to maintain Keff less than or equal to 0.95, 
including conservative allowances for uncertainties and biases, when 
the pool is flooded with unborated water. The proposed changes 
include a requirement for fuel assemblies with enrichments above 4.2 
w/o U-235 to contain sufficient integral fuel burnable absorbers 
such that the maximum reference fuel K infinity is less than or 
equal to 1.470 in unborated water at 68 deg.F due to restrictions on 
spent fuel storage. Should a postulated accident occur which causes 
a reactivity increase in the Byron and Braidwood Spent Fuel Pools, 
Keff will be maintained less than or equal to 0.95 due to the 
presence of at least 300 ppm of soluble boron in the spent fuel 
pool. The proposed changes do not affect any plant safety parameters 
or setpoints.
    The proposed changes ensure that the design basis for preventing 
criticality in the fuel storage areas is preserved, and fuel cycle 
designs will continue to be analyzed using NRC accepted codes and 
methods to ensure the design bases are satisfied.
    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: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power Company (CYAPCO), and Northeast 
Nuclear Energy Company (NNECO), Docket Nos. 50-213 and 50-245, 
Haddam Neck Plant, and Millstone Nuclear Power Station, Unit 1, 
Middlesex County, and New London County, Connecticut

    Date of amendment request: October 31, 1994
    Description of amendment request: The proposed amendments would 
renew the existing license conditions for both plants to implement and 
maintain Integrated Implementation Schedule (IIS) Program Plans (the 
Program Plan). The Program Plans provide a methodology to be followed 
for scheduling plant modifications and engineering evaluations.
    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:
    YAPCO and NNECO have reviewed the proposed changes in accordance 
with 10 CFR 50.92 and conclude that the changes do not involve a SHC 
[significant hazards consideration]. The basis for this conclusion 
is that the three criteria of 10 CFR 50.92(c) are not compromised. 
The proposed changes do not involve an SHC because the changes would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    Operation of the facilities in accordance with these proposed 
changes would require the implementation of the IIS methodology 
described in the Program Plans. As such, it requires that CYAPCO and 
NNECO establish an administrative means for tracking, prioritizing, 
and scheduling NRC-required plant modifications and engineering 
evaluations, and licensee identified plant improvement projects. 
This methodology is intended to enhance plant safety by more 
effectively controlling the number and scheduling of plant 
modifications, thereby assuring that issues required for safe 
operation of the plants receive priority and are completed in a 
timely manner. Because the license conditions address only an 
administrative scheduling mechanism, it does not affect directly the 
design or operation of the plant. Therefore, no accident analyses 
are affected and the proposed changes do not increase the 
probability or consequences of any previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed license conditions establish a requirement related 
to scheduling of modifications and engineering evaluations. Because 
the license conditions address only an administrative scheduling 
mechanism, they do not affect directly the design or operation of 
the plants. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from those 
previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The proposed license conditions renew administrative 
requirements intended to enhance public safety and reliable plant 
operation. The proposed license conditions do not affect any 
accident analyses, directly modify the plant configurations, or 
change the way the plants are operated. The methodologies are 
intended to enhance plant safety by more effectively controlling the 
number and scheduling of plant modifications, thereby assuring that 
issues required for safe operation of the plants receive priority 
and are completed in a timely manner. Because the license conditions 
address only an administrative scheduling mechanism, they do not 
affect directly the design or operation of the plants. Therefore, 
the proposed changes do not involve a reduction in any margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, Connecticut 06457, for the Haddam Neck Plant, and 
the Learning Resource Center, Three Rivers Community-Technical College, 
Thames Valley Campus, 574 New London Turnpike, Norwich, CT 06360, for 
Millstone Unit 1.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: August 5, 1994, as supplemented on 
November 17, 1994.
    Description of amendment request: This amendment is an additional 
followup to the amendment request of May 29, 1992, published in the 
Federal Register on July 8, 1992 (57 FR 30242), which changed the 
Technical Specifications (TSs) Section 1.0, Definitions, to accommodate 
a 24-month fuel cycle and which proposed the extension of the test 
intervals for specific surveillance tests. This amendment proposes 
extending the surveillance intervals to 24 months for the following 
additional surveillance tests:
    (1) Charging Flow Instrumentation
    (2) Containment Sump, Recirculation Sump, and Reactor Cavity 
Continuous Level Instrument Channels
    (3) Auxiliary Feedwater Flow Rate Channel
    (4) Control Room Air Filtration System
    (5) Post Accident Containment Venting System
    (6) Liquid Rad-Waste Flow Channel
    (7) Steam Generator Blowdown Flow Channel
    (8) Liquid Waste Distillate Tank Level Channels
    (9) Primary Water Storage Tank Level Instrumentation
    (10) Flow Rate Monitors; Plant Vent (Unit 2) and Stack Vent (Unit 
1)
    (11) Stack Vent Noble Gas Activity Monitor (R-60)
    (12) High Pressure Water Fire Protection System
    (13) Fire Protection System Diesel Engine
    (14) Electrical Tunnel, Diesel Generator Building, and Containment 
Fan Cooler Fire Protection Spray Systems; (A) System Functional Test 
and (B) Spray Header Visual Inspection
    (15) Penetration Fire Barriers
    (16) Smoke Detectors/Electrical Penetration Area Inside Containment
    (17) Functional Testing of Containment Sump Pumps
    The changes requested by the licensee are in accordance with 
Generic Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Charging Flow Instrumentation
    The proposed change does not involve a significant hazards 
consideration since:
    1. A significant increase in the probability or consequences of 
an accident previously evaluated will not occur.
    It is proposed that the channel calibration frequency for the 
Charging Flow instrumentation be changed from 18 months (+25%) to 
every 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists to accommodate the 
channel statistical error resulting from a 30 month operating cycle. 
The existing margin provides assurance that plant protective actions 
will occur as required. It is therefore concluded that changing the 
surveillance interval from 18 months (+25%) to 24 months (+25%) will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any accident previously evaluated has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds the current margin. Plant 
equipment will provide protective functions to assure that Safety 
Analysis limits are not exceeded. This will prevent the possibility 
of a new or different kind of accident from any previously evaluated 
from occurring.
    3. A significant reduction in a margin of safety is not 
involved.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds current margin. This margin, which is 
equivalent to the existing margin, is necessary to assure that 
protective safety functions will occur so that Safety Analysis 
limits are not exceeded.(2) Containment Sump, Recirculation Sump, 
and Reactor Cavity Continuous Level Instrument Channels
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the calibration and test frequency for the 
Containment Sump, Recirculation Sump and Reactor Cavity continuous 
level monitoring instrument channels be revised from every 18 months 
(+25%) to 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists to accommodate the 
channel statistical error resulting from a 30 month operating cycle. 
The existing margin provides assurance that plant protective actions 
will occur as required. It is therefore concluded that changing the 
surveillance interval from 18 months (+25%) to 24 months (+25%) will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds current margin. Plant equipment 
will provide protective functions to assure that Safety Analysis 
limits are not exceeded. This will prevent the possibility of a new 
or different kind of accident from any previously evaluated from 
occurring.
    3. There has been no reduction in the margin of safety.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds current margin. This margin is necessary to 
assure that protective safety functions will occur so that safety 
analysis limits are not exceeded.
    (3) Auxiliary Feedwater Flow Rate Channel
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the calibration frequency for the Auxiliary 
Feedwater Flow Rate channel be revised from 18 months (+25%) to 24 
months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists between the existing 
technical specification limits and the licensing basis Safety 
Analysis limit to accommodate the channel statistical error 
resulting from a 30 month operating cycle. The existing margin 
between the Technical Specification limit and the Safety Analysis 
limit provides assurance that plant protective actions will occur as 
required. It is therefore concluded that changing the surveillance 
interval from 18 months (+25%) to 24 months (+25%) will not result 
in a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds the current margin between the 
existing Technical Specification limit and the Safety Analysis 
limit. Plant equipment, which will be set at (or more conservatively 
than) Technical Specification limits, will provide protective 
functions to assure that safety analysis limits are not exceeded. 
This will prevent the possibility of a new or different kind of 
accident from any previously evaluated from occurring.
    3. There has been no reduction in the margin of safety.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds the margin which exists between the current 
Technical Specification limit and the licensing basis Safety 
Analysis limit. This margin, which is equivalent to the existing 
margin, is necessary to assure that protective safety functions will 
occur so that Safety Analysis limits are not exceeded.
    (4) Control Room Air Filtration System
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the surveillance frequency for the Control 
Room Air Filtration System be changed from every 18 months (+25%) to 
every 24 months (+25%).
    For the flow tests, data from 1986 to date indicates that the 
Control Room Filtration System performed in an acceptable manner 
when surveilled on an 18 month (+25%) basis. The only discrepancy 
was due to a hardware error and was independent of the time between 
surveillances. Per Generic Letter 91-04, this past test history 
provides an adequate basis to conclude that an extended operating 
cycle would have minimal impact upon the flow characteristics of the 
Control Room Filtration System. The modification of the filtration 
system in 1993 only enhanced system performance.
    With regard to the absorbance properties of the charcoal, 
previous test data highlights a problem occurring during the 1986-
1987 period which subsequent testing confirms was adequately 
resolved.
    With the 1993 modification which increased the carbon bed 
thickness from 1'' to 4'', performance can only be enhanced.
    Therefore, it is concluded that a significant increase in the 
probability or consequences of an accident previously evaluated will 
not be incurred by changing the surveillance interval from 18 months 
(+25%) to 24 months (+25%).
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    A review of past historical surveillance data over 7 years 
indicates no failures which were time dependent. The modification, 
which was performed in 1993, can only enhance performance of the 
system. New fans, an increased charcoal bed thickness, and new HEPA 
[high-efficiency particulate air] filters will increase the 
reliability of the system. Thus, it is concluded that the 
possibility of a new or different kind of accident than that 
previously evaluated has not been created.
    3. There has been no significant reduction in the margin of 
safety.
    Past test data validated the acceptability of the previous air 
filtration system for an extended surveillance interval. The 
modification performed in 1993 will only enhance the reliability and 
performance of the air filtration system. Thus, it is concluded that 
a significant reduction in the margin of safety is not involved.
    (5) Post Accident Containment Venting System
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the surveillance frequency for the Post 
Accident Containment Venting system be revised from every 18 months 
(+25%) to 24 months (+25%).
    A review of past test history from 1986 to date indicates that 
the Post Accident Containment Venting System performed in a 
satisfactory manner when the surveillance period was 18 months 
(+25%). There was one discrepant condition noted in the 1989 test, 
which, based upon subsequent tests in 1991 and 1993, does not appear 
to have been age related. The 1989 observation concerning a gasket 
is considered to be a one time only event and unlikely to reoccur as 
a result of extending the surveillance interval from 18 months 
(+25%) to 24 (+25%).
    An added consideration, in terms of safety significance, is the 
fact that the Post Accident Containment Venting system is diverse 
and redundant to the post accident hydrogen recombiners which are 
themselves redundant and the primary means of reducing the post 
accident hydrogen concentration within containment. The venting 
system is not relied upon for containment pressure control.
    Due to the satisfactory past test history of the venting system, 
together with its secondary role as a means of controlling post 
accident hydrogen concentration, it is concluded that a significant 
increase in the probability or consequences of an accident 
previously evaluated will not be incurred by changing the 
surveillance interval from 18 months (+25%) to 24 months (+25%).
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    A review of past historical surveillance data over 7 years 
indicates no failures which are considered to be time dependent. 
Although one discrepant condition was observed in the 1989 test it 
was not repeated in subsequent surveillances. Per Generic Letter 91-
04, this constitutes a sufficient basis for revising the 
surveillance interval from 18 months (+25) to 24 months (+25%). This 
extension in the operating interval is not expected to have an 
impact upon the availability of the system. Thus, it is concluded 
that the possibility of a new or different kind of accident 
previously evaluated has not been created.
    3. There has been no reduction in the margin of safety.
    As past test data validates the presumption that an extended 
operating cycle will not impact the availability of the Post 
Accident Containment Venting Systems, it is concluded that a 
significant reduction in the margin of safety is not involved.
    (6) Liquid Rad-Waste Flow Channel
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the channel calibration frequency for the 
Liquid Rad-Waste Flow Channel be revised from every 18 months 
(+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists to accommodate the 
channel statistical error resulting from a 30 month operating cycle. 
The existing margin provides assurance that plant protective actions 
will occur as required. It is therefore concluded that changing the 
surveillance interval from 18 months (+25%) to 24 months (+25%) will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds current margin. Plant equipment 
will be set to provide protective functions to assure that Safety 
Analysis limits are not exceeded. This will prevent the possibility 
of a new or different kind of accident from any previously evaluated 
from occurring.
    3. There has been no reduction in the margin of safety.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds the allowable operating margin. This margin, 
which is equivalent to the existing margin, is necessary to assure 
that protective safety functions will occur so that Safety Analysis 
limits are not exceeded.
    (7) Steam Generator Blowdown Flow Channel
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the channel calibration frequency for the 
Steam Generator Blowdown Flow channel be revised from every 18 
months (+25%) to every 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists to accommodate the 
channel statistical error resulting from a 30 month operating cycle. 
The existing margin provides assurance that plant protective actions 
will occur as required. It is therefore concluded that changing the 
surveillance interval from 18 months (+25%) to 24 months (+25%) will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds the current margin. Plant 
equipment will provide protective functions to assure that Safety 
Analysis limits are not exceeded. This will prevent the possibility 
of a new or different kind of accident from any previously evaluated 
from occurring.
    3. There has been no reduction in the margin of safety.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds the allowable operating margin. This margin, 
which is equivalent to the existing margin, is necessary to assure 
that protective safety functions will occur so that Safety Analysis 
limits are not exceeded.
    (8) Liquid Waste Distillate Tank Level Channels
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the channel calibration frequency for the 
Liquid Waste Distillate Tank level of tanks 13 and 14 be revised 
from every 18 months (+25%) to every 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists to accommodate the 
channel statistical error resulting from a 30 month operating cycle. 
The existing margin provides assurance that plant protective actions 
will occur as required. It is therefore concluded that changing the 
surveillance interval from 18 months (+25%) to 24 months (+25%) will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a channel 
statistical allowance which exceeds current margin. Plant equipment 
will be set to provided protective functions to assure that Safety 
Analysis limits are not exceeded. This will prevent the possibility 
of a new or different kind of accident from any previously evaluated 
from occurring.
    3. There has been no reduction in the margin of safety.
    The above change in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds the allowable operating margin. This margin, 
which is equivalent to the existing margin, is necessary to assure 
that protective safety functions will occur so that safety analysis 
limits are not exceeded.
    (9) Primary Water Storage Tank Level Instrumentation
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the channel calibration frequency for the 
Primary Water Storage Tank Level instrumentation be changed from 
every 18 months (+25%) to 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
surveillance has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists to accommodate the 
channel statistical error resulting form a 30 month surveillance. 
The existing margin provides assurance that plant protective actions 
will occur as required. It is therefore concluded that changing the 
surveillance interval from 18 months (+25%) to 24 months (+25%) will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The proposed change in surveillance interval will result in a 
channel statistical allowance which can be accommodated over a 30 
month operating cycle. Plant equipment, which will be set at (or 
more conservatively than) Technical Specification limits, will 
provide protective functions to assure that Safety Analysis limits 
are not exceeded. This will prevent the possibility of a new or 
different kind of accident from any previously evaluated from 
occurring.
    3. There has been no significant reduction in the margin of 
safety.
    The above changes in surveillance interval resulting from an 
increased operating cycle will not result in a channel statistical 
allowance which exceeds current margin. This margin is necessary to 
assure that protective safety functions will occur so that Safety 
Analysis limits are not exceeded.
    (10) Flow Rate Monitors; Plant Vent (Unit 2) and Stack Vent 
(Unit 1)
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the calibration frequency for the flow rate 
monitors for the Plant Vent (Unit 2) and the Stack Vent (Unit 1) be 
revised from every 18 months (+25%) to every 24 months (+25%).
    A statistical analysis of channel uncertainty for a 30 month 
operating cycle has been performed. Based upon this analysis it has 
been concluded that sufficient margin exists to accommodate the 
statistical error resulting from a 30 month operating cycle. The 
existing margin provides assurance that plant protective actions 
will occur as required. It is therefore concluded that changing the 
surveillance interval from 18 months (+25%) to 24 months (+25%) will 
not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The proposed change in operating cycle length due to an 
increased surveillance interval will not result in a statistical 
allowance which exceeds the current margin. Plant equipment will be 
calibrated to provide data to assure that safety analysis limits are 
not exceeded. This will prevent the possibility of a new or 
different kind of accident from an previously evaluated from 
occurring.
    3. There has been no reduction in the margin of safety.
    The proposed change in the surveillance interval resulting from 
an increased operating cycle will not result in a channel 
statistical allowance which exceeds the allowable operating margin. 
This margin, which is equivalent to the existing margin, is 
necessary to assure that protective safety functions will occur so 
that safety analysis limits are not exceeded.
    (11) Stack Vent Noble Gas Activity Monitor (R-60)
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the calibration frequency for the stack vent 
noble gas activity monitor be revised from every 18 months (+25%) to 
every 24 months (+25%).
    The current monitor replaced the previous monitor and therefore 
there is only one refueling cycle surveillance data available which 
proved to be satisfactory. The vendor recommends a calibration 
period based on user experience. Insofar as the 18 month (+25%) 
surveillance has proven to be acceptable, extension to a 24 month 
(+25%) cycle is consistent with the vendor's recommendation. Any 
additional uncertainty generated due to the extended surveillance is 
bounded by the uncertainty inherent in a grab sample taken once per 
24 hours which is the required compensatory action should the 
monitor be inoperable. Since setpoints for alarms are not critical 
to either plant operation or safety, since extensive margin is 
reflected between the setpoint and applicable limits, it is 
concluded that any additional uncertainty involved in a longer 
surveillance cycle will not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    This monitor measures the activity of potentially radioactive 
gaseous effluent through the stack vent. The alarm setpoints are set 
at a point sufficiently above expected radioactivity levels to avoid 
unnecessary alarms and, at the same time, far below discharge 
limits. The purpose of the monitor is to annunciate in the event an 
unexpected spike in radioactivity level should occur so that 
corrective action can be taken prior to exceeding a discharge limit. 
The margin that exists between the discharge limit and the setpoint 
is more than sufficient to accommodate any drift that could be 
practically expected in a 24 month (+25%) operating cycle.
    In this capacity, the monitor does not have setpoints which are 
critical to plant operation or safety. Readings are not used in a 
quantitative manner nor is accuracy important. It is important that 
the instrument remain operable and respond to step changes in 
radioactivity level over the operating cycle. It is therefore 
concluded that an extended operating cycle will not result in the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. There has been no reduction in the margin of safety.
    Sufficient margin exists between plant setpoints and applicable 
limits to accommodate any realistic drift projected to occur over a 
30 month operating cycle. Furthermore, instrument indications are 
not used in a quantitative manner nor is instrument accuracy of 
importance. Therefore, it is concluded that no significant reduction 
in the margin of safety will result from an extended operating 
cycle.
    (12) High Pressure Water Fire Protection System
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the system functional test of the High-
pressure Water Fire Protection System be changed from every 18 
months (+25%) to every 24 months (+25%).
    This system is a static system which is not normally required to 
operate. The main fire pumps are on standby and are not in operation 
except for testing. Thus, almost no wear is induced as a function of 
time except that which results from being in standby status which is 
minimal and slow acting. Under these circumstances, extending the 
operating cycle between surveillances would be expected to have 
negligible affect upon system operability. It is therefore concluded 
that there would be no significant increase in the probability or 
consequences of an accident as a result of an extended interval 
between surveillances.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Extension of the plant operating cycle will primarily extend the 
time the pumps are in standby capacity. The potential for system 
deterioration is minimal under these circumstances. Any 
deterioration that does occur will be slow acting with respect to 
time. A significant deterioration would be detected by a monthly 
pump operating test. Thus, an extended operating cycle is not 
expected to create the possibility of a new or different kind of 
accident form [from] any previously analyzed.
    3. There has been no reduction in the margin of safety.
    Extension of the operating cycle by several months only serves 
to extend the period of time when the pumps are in standby status. 
Any deterioration under these circumstances will be slow acting. 
Significant deterioration would be detected by the monthly operating 
test. Therefore, it is concluded that an extended interval between 
surveillances will involve no significant reduction in the margin of 
safety.
    (13) Fire Protection System Diesel Engine
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the Fire Protection System Diesel Engine 
Functional test be changed from every 18 months (+25%) to every 24 
months (+25%).
    Except for periodic testing, the diesel is in a standby state 
and not subject to operational stress. Periodic testing imposes 
limited wear as evidenced by the absence of major repairs during 
past maintenance. Extension of the operating cycle for several 
months is expected to have virtually no impact upon diesel 
operability. Monthly testing would detect any degradation. Thus it 
is concluded that there would be no significant increase in the 
probability or consequences of an accident as a result of an 
extended interval between surveillances.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Extension of the plant operating cycle will, for the most part, 
only extend the time spent by the pumps in standby capacity. The 
potential for system deterioration is minimal under these 
circumstances. Any deterioration that does occur will be slow acting 
with respect to time. Significant deterioration in performance would 
be detected by the monthly pump operating test. Thus, an extended 
operating cycle is not expected to create the possibility of a new 
or different kind of accident from any previously analyzed.
    3. There has been no reduction in the margin of safety.
    Extension of the operating cycle by several months only serves 
to extend the period of time when the pumps are in standby status. 
Any deterioration under these circumstances will be slow acting and 
significant deterioration would be detected by the monthly operating 
test. Therefore, it is concluded that an extended interval between 
surveillances will involve no significant reduction in the margin of 
safety.
    (14) Electrical Tunnel, Diesel Generator Building, and 
Containment Fan Cooler Fire Protection Spray Systems; (A) System 
Functional Test and (B) Spray Header Visual Inspection
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    (a) It is proposed that the functional test surveillance 
interval for the Electrical Tunnel, Diesel Generator Building and 
Containment Fan Cooler Fire Protection Spray Systems be changed from 
every 18 months (+25%) to every 24 months (25%).
    (b) It is proposed that the Spray Header visual inspection 
interval be revised from every 18 months (25%) to 24 months (+25%).
    Extension of the surveillance interval for Electrical Tunnel and 
Diesel Generator Building Fire Protection System functional tests 
will have virtually no impact upon the operability of these systems. 
These systems are accessible during normal operation and other 
sections of the Technical Specifications (4.14.A.1.g.(i) and 
4.14.B.1.a(i)) require that the system valve tests be conducted on 
an annual (12 month) basis. These annual tests would reveal any 
system deterioration prior to the conclusion of the proposed 
extended surveillance interval.
    For the Fan Cooler Fire Protection System as well as the Spray 
Header itself, evaluation of surveillance data from the past five 
refueling outages indicates minor discrepancies which would not have 
impaired system operability.
    It is therefore concluded that extension of the proposed 
surveillance interval will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    For the Electrical Tunnel and Diesel Generator Building, 
extension of the surveillance interval will have a negligible affect 
as other portions of the Technical Specifications require the same 
surveillance on an annual basis. For the spray header and the fan 
cooler fire protection system, historical surveillance data 
validates operability over an 18 month (+25%) interval which lends 
confidence to conclude that operability will be maintained over a 24 
month (+25%) interval. It is therefore concluded that the 
possibility of a new or different kind of accident from any accident 
previously evaluated has not been introduced.
    3. There has been no reduction in the margin of safety.
    Extension of the surveillance for two systems will have minimal 
impact as the Technical Specifications impose more frequent testing 
for system valves on an annual basis. For the Spray Header and Fan 
Cooler Fire Protection System, as well as the fire protection system 
for the Diesel Generator Building and Electrical Tunnel, it can be 
stated that these systems are static existing mainly in a standby 
capacity under which little deterioration would be expected. Past 
surveillance data validates system reliability. It is therefore 
concluded that increasing the time interval between inspections 
would not involve a significant reduction in the margin of safety.
    (15) Penetration Fire Barriers
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the visual inspection frequency of the 
penetration fire barriers listed in the Technical Specifications be 
changed from every 18 months (+25%) to every 24 months (+25%).
    The fire barrier penetration seals are static devices existing 
in standby status. Normal environmental conditions exist during 
normal plant operations. The only deterioration expected would be 
that due to aging in a normal ambient which would be minimal to non-
existent. Evaluation of unacceptable seals detected during 
surveillances indicates that initial seal installation was faulty 
and aging was not the cause. Surveillances during four refueling 
outages confirm this evaluation. Accordingly, it is not expected 
that the proposed change in surveillance interval will involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Past surveillances indicate that time is not a predominate 
failure mechanism. In the few unacceptable seals detected, the 
initial installation procedure has been identified as the cause of 
the problems. Since the seals are static devices which exist in a 
standby condition and experience normal ambient conditions during 
normal operation, this would be the expected conclusion. In 
addition, the fire barriers are just one means of fire protection. 
Other means of fire protection exist such as fire alarms, sprinklers 
and heat detectors which provide defense in depth. Thus, it is 
concluded that the proposed change in the surveillance interval will 
not create the possibility of a new or different kind of accident 
from that previously evaluated.
    3. There has been no reduction in the margin of safety.
    Aging has not been identified as a principle contributor to seal 
failures. In addition, there exists additional means of fire 
protection which provides defense in depth. Therefore, the proposed 
change in surveillance intervals is not expected to involve a 
significant reduction in the margin of safety.
    (16) Smoke Detectors/Electrical Penetration Area Inside 
Containment
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the surveillance interval for the smoke 
detectors located in the electrical penetration area inside 
containment be revised from every 18 months (+25%) to every 24 
months (+25%).
    Based on data taken from six surveillances from 1984 through and 
including 1993, these devices have proven to be highly reliable. No 
test failures were observed during this period. Based on the 
guidance contained in Generic Letter 91-04, this demonstration of 
reliable performance provides an adequate basis to conclude that the 
proposed extension in the surveillance interval will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Only 3 of the 5 detectors are required during normal operation. 
Past surveillance data from six refueling outages indicate that it 
is reasonable to expect all 5 detectors will remain operable over 
the extended operating cycle which provides margin. It is therefore 
concluded that the possibility of a new or different kind of 
accident from any accident previously evaluated has not been 
created.
    3. There has been no reduction in the margin of safety.
    The proven reliability of these devices indicates that a 
significant reduction in the margin of safety would not be involved 
in extending the operating cycle to 24 months (+25%).
    (17) Functional Testing of Containment Sump Pumps
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the functional test of the Containment Sump 
Pump be changed from every 18 months (+25%) to every 24 months 
(+25%).
    No credit is taken within the FSAR for the Containment Sump 
Pumps as a means of mitigating the consequences of an accident. 
During normal operation the pumps serve as a means of quantifying 
leakage inside Containment and therefore serve a safety function in 
terms of accident prevention. However, in this capacity they are 
only one of several systems which are capable of serving this 
function and their failure would not result in a loss of this 
capability.
    In addition, evaluation of surveillance data back to 1986 
indicates, with one exception, that the devices are very reliable. 
In one instance, the pumps did not actuate or cause operation within 
the setpoint tolerance but did operate as required. This was 
determined not to be a time dependent event.
    It is therefore concluded that extending the interval between 
refueling surveillances will not result in a significant increase in 
the probability or consequences of an accident.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    Past surveillances indicate that time is not a predominate 
failure mechanism. Also, there exists a Technical Specification 
requirement to perform almost the same surveillance on a monthly 
basis in addition to every refueling outage. This monthly test 
diminishes any potential risk in extending the operating cycle. It 
is therefore concluded that the possibility of a new or different 
kind of accident from any previously analyzed has not been created.
    3. There has been no reduction in the margin of safety.
    Past surveillance data indicates that pump operation is 
reliable. In addition, there are alternate means of providing the 
safety function fulfilled by these pumps. Also, a monthly test is 
required which would detect any malfunction prior to the end of an 
extended operating cycle. It is therefore concluded that extending 
the operating cycle by several months will not result in a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: Michael J. Case, Acting

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: November 2, 1994
    Description of amendment request: The proposed amendments will 
upgrade existing TS 3/4.6.2.1 and TS 3/4.6.2.3 by adapting the combined 
specification for Containment Spray and Cooling Systems, contained in 
the Standard Technical Specifications for Combustion Engineering 
Plants, to the St. Lucie units. The changes account for plant-specific 
differences and include all related requirements of NUREG-1432, Rev. O, 
specification 3.6.6A. Accordingly, the proposal is consistent with the 
Commission's Final Policy Statement on Technical Specifications 
Improvements (58 FR 39132).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10 CFR 50.92, a determination may be made that a 
proposed license amendment involves no significant hazards 
consideration if 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. Each standard is discussed as 
follows:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment will upgrade the existing Limiting 
Conditions for Operation (LCOs) associated with the Containment 
Cooling and Spray Systems to be consistent with NUREG-1432, Standard 
Technical Specifications for Combustion Engineering Plants. The 
Containment Cooling and Spray Systems are not initiators of 
accidents previously evaluated, but are included as part of the 
success paths associated with mitigating various accidents and 
transients. The redundancy afforded by Containment Cooling and Spray 
Systems in conjunction with the requirements of the proposed LCO 
assures that the safety function of these systems can be 
accomplished considering single failure criteria. Neither the design 
nor the safety function of the Containment Cooling and Spray Systems 
have been altered, and the proposed amendment does not change the 
applicable plant safety analyses. Therefore, operation of the 
facility in accordance with the proposed amendment will not involve 
a significant increase in the probability 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 accident previously evaluated.
    The proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. The changes are 
administrative in nature in that they do not involve the addition of 
new equipment or the modification of existing equipment, nor do they 
otherwise alter the design of St. Lucie Unit 1 & 2 systems. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The safety function of the Containment Cooling System is to 
provide containment heat removal during normal operation and 
accident conditions. The safety function of the Containment Spray 
System is to provide containment heat and iodine removal during 
accident conditions. The proposed amendment, in conjunction with the 
redundancy afforded by the Containment Cooling and Spray system 
design, assures that these safety functions can be accomplished 
considering single-failure criteria. The bases for required actions 
and the action completion times specified for inoperable Containment 
Cooling and Spray trains are consistent with the corresponding 
specifications in NUREG-1432. The safety analyses for applicable 
accidents and transients remain unchanged from those previously 
evaluated and reported in the Updated Final Safety Analysis Report. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Based on the above discussion and the supporting Evaluation of 
Technical Specification changes, FPL has determined that the 
proposed license amendment involves no significant hazards 
consideration.
    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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Mohan Thadani, Acting

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

    Date of amendment request: November 7, 1994
    Description of amendment request: The proposed amendment would 
change the number of diesel generators (emergency power supply) 
required to be operable during Mode 5 with the loops filled and Mode 6 
with greater than or equal to 23 feet of water above the reactor vessel 
flange. In addition, changes to certain system specifications that are 
affected by the changes for the emergency power supply were also 
proposed.
    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 accidents previously 
evaluated.
    The equipment which is affected by the technical specification 
changes proposed here are not precursors to any accident postulated 
to occur in Modes 5 and 6. Therefore, the probability of an accident 
is not increased. A design review has demonstrated the ability of 
the required systems to perform their accident mitigation functions 
for the postulated accidents during Mode 5 and 6 operation. 
Therefore, it is concluded that an increase in the consequences of 
the postulated accidents will not result from the proposed Technical 
Specifications.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The system design, function, and performance is not affected by 
these specifications. No new equipment interactions are created. 
Calculations and Failure Modes and Effects Analyses (FMEA) have been 
conducted for selected mechanical systems and show there are no 
failures which would cause situations where applicable accidents 
would not be mitigated or which would cause new accidents. On this 
basis, the proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The electrical power system specifications support the equipment 
required to be operable, commensurate with the current level of 
safety, including the equipment requiring a diesel backed power 
source. The design review results demonstrate that operation in 
Modes 5 and 6, in accordance with the proposed Technical 
Specification changes, is acceptable from an accident mitigation 
standpoint. The basic Modes 5 and 6 plant system functions are not 
changed. On this basis, the proposed change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 7, 1994
    Description of amendment request: The proposed amendment would 
permit both containment personnel airlock doors to be open while moving 
fuel during refueling operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to Technical Specification 3.9.4, 
Containment Building Penetrations, would allow the containment 
personnel airlock to be open during fuel movement and core 
alterations. The containment personnel airlock is closed during fuel 
movement and core alterations to prevent the escape of radioactive 
material in the event of a fuel handling accident. The containment 
personnel airlock is not an initiator to any accident. Whether the 
containment personnel airlock doors are open or closed during fuel 
movement and core alterations has no affect on the probability of 
any accident previously evaluated.
    The proposed change does alter assumptions previously made in 
evaluating the radiological consequences of the fuel handling 
accident inside the reactor containment building. The proposed 
change allows for the containment personnel airlock to be open 
during refueling. The radiological consequences described in this 
change are bounded by those given in the South Texas Project Safety 
Evaluation Report and General Design Criteria 19. All doses for the 
proposed change are less than the acceptance criteria, therefore, 
there is no significant increase in the consequences of an accident 
previously analyzed.
    The proposed change will significantly reduce the dose to 
workers in the containment in the event of a fueling handling 
accident by accelerating the containment evacuation process. The 
proposed change will also significantly decrease the wear on the 
containment personnel airlock doors and, consequently, increase the 
reliability of the containment personnel airlock doors in the event 
of an accident.
    Since the probability of a fuel handling accident is unaffected 
by the airlock door positions, and the increased doses do not exceed 
acceptance limits, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change affects a previously evaluated accident, 
e.g., a fuel handling accident inside containment. The existing 
accident has been modified to account for the containment personnel 
airlock doors being opened at the time of the accident. It does not 
represent a significant change in the configuration or operation of 
the plant and, therefore, does not create the possibility of a new 
or different type of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The margin of safety is reduced when the offsite and control 
room doses exceed the acceptance criteria in the STP SER. As 
previously discussed in the response to question 1, the offsite and 
control room doses are below the acceptance criteria. Therefore, 
this proposed change does not significantly reduce the 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, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 8, 1994
    Description of amendment request: The proposed amendment would 
require only one of the two battery chargers associated with each Class 
1E 125 VDC Channel I and Channel IV to be operable.
    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 accidents previously 
evaluated.
    A single charger is able to maintain the operability of Channel 
I or Channel IV at the design loading with a single failure 
condition. The proposed change does not alter equipment or 
assumptions made in previously evaluated accidents. The consequences 
of previously evaluated accidents are not increased. On this basis, 
the proposed change does not involve a significant increase in the 
probability or consequences of accidents previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change involves only the operability requirement 
for the second battery charger in Channel I and Channel IV. The 
failure modes and operating modes would then be identical for all 
four STPEGS Class 1E DC channels. Failure modes and effects analyses 
already performed for DC Channels II and III would thus become 
applicable to Channels I and IV also. The change proposed by this 
Technical Specification revision is bounded by the failure modes and 
effects analysis provided as Table 8.3-8 of the STPEGS UFSAR 
[Updated Final Safety Analysis Report]. On this basis, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change involves only the operability requirement 
for the second battery charger in Channel I and Channel IV. The 
number and capacity of DC channels required is not affected by the 
proposed change. The electrical loads supported by these DC channels 
are not changed and the duration of their function is not impacted. 
On this basis, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW, Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 8, 1994
    Description of amendment request: The proposed amendment would 
permit the substitution of an extended range neutron flux monitor for 
one of the source range neutron flux monitors during refueling 
operations.
    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 as previously 
evaluated.
    During refueling operations, the Source Range channels are used 
only for monitoring changes in core reactivity, and does not provide 
inputs for automatically actuated equipment. The same function could 
be performed by an Extended Range channel. The combination of the 
present Channel Check and the proposed Channel Calibration are 
sufficient to ensure that the detectors are capable of monitoring 
core reactivity changes. By providing the intended redundant core 
reactivity monitoring, neither the possibility or consequences of an 
accident previously evaluated are increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    During refueling operations, the Source Range Monitors are used 
simply as monitoring instrumentation. Extended Range Monitors are 
capable of performing this function. The combination of the present 
Channel Check and the proposed Channel Calibration are sufficient to 
ensure that the detectors are capable of monitoring core reactivity 
changes.
    The proposed change would require revision of STP refueling 
procedures. However, the physical movement of fuel assemblies is 
within the scope of current refueling procedures. No new mechanism 
for fuel misloading or damage or boron dilution would be created by 
the change. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change provides core reactivity monitoring 
comparable to that provided by the use of the Source Range channels. 
The Extended Range channel is capable of detecting core reactivity 
changes and provides the intended redundancy. The combination of the 
present Channel Check and the proposed Channel Calibration are 
sufficient to ensure that the detectors are capable of monitoring 
core reactivity changes. No margin of safety is compromised by this 
change.
    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, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: October 24, 1994
    Description of amendment request: The proposed amendment would 
modify Technical Specifications Table 4.1-3 surveillance requirements 
for new emergency feedwater flow instrumentation. Specifically, the 
currently installed analog feedwater flow transmitters would be 
replaced by new, digital-type flow transmitters. The new digital flow 
emergency feedwater flow transmitters are continuously self-checking 
and have a recommended calibration interval of 9 years. The licensee 
proposes to verify flow whenever the system operates and send one 
transmitter back to the manufacturer for recalibration every 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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The staff's analysis is 
presented below:
    1. The proposed change does not involve a significant increase in 
the
     probability or consequences of an accident previously evaluated.
    Performance of Technical Specifications Table 4.1-3 (items 10 a and 
b) ensures the emergency feedwater flow transmitters are operable when 
required. The proposed change will continue to ensure operabililty and 
therefore will not increase 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.
    Emergency feedwater flow transmitter operability verification is 
maintained, with no change to the system's configuration. Thus, there 
is no unique operating condition that could adversely affect system 
functional performance.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    There is no change to any Final Safety Analysis Report Chapter 14 
(Safety Analysis) event. There is no change to the demonstration of 
component operability; thus, the proposed change does not involve a 
significant reduction in a margin of safety.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, Maine 04578
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, 329 Bath Road, Brunswick, Maine 04011
    NRC Project Director: Walter R. Butler

North Atlantic Energy Service Corporation, Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: October 7, 1994
    Description of amendment request: The proposed amendment would 
remove from the Technical Specifications certain audit responsibilities 
of the Nuclear Safety Audit Review Committee and certain review 
responsibilities of the Station Operation Review Committee relating to 
the Emergency Plan and Security Plan and their implementing procedures. 
The proposed changes are consistent with the guidance of Generic Letter 
93-07.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)) because the proposed changes do not affect the manner by 
which the facility is operated and do not change any facility design 
feature or equipment. Since there is no change to the facility or 
operating procedures, there is no affect upon the probability or 
consequences of any accident previously analyzed.
    B. The changes do not create the possibility of a new or different 
kindof accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because they do not affect the manner by which the 
facility is operated. The proposed changes merely affect audit and 
review responsibilities and their deletion or relocation to other 
controlled documents does not introduce new or different accident 
scenarios.
    C. The changes do not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)) because the proposed changes do not 
affect the manner by which the facility is operated or involve 
equipment or features which affect the operational characteristics of 
the facility.
    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:  Exeter Public Library, 47 
Front Street, Exeter, NH 03833.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One 
International Place, Boston, MA 02110-2624.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: October 14, 1994
    Description of amendment request: The proposed change clarifies the 
low pressure coolant injection (LPCI) requirements as required by 
Technical Specification 4.5.A.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed change in accordance with 10 CFR 
50.92 and concluded that the change does not involve a significant 
hazards consideration (SHC). The basis for this conclusion is that 
the three criteria of 10 CFR 50.92(c) are not compromised. The 
proposed change does not involve an SHC because the change would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The LPCI flow surveillance requirement to demonstrate that three 
pumps can deliver 15,000 gpm does not relate to any previously 
analyzed accident. There are no accident scenarios which rely upon 
three pumps or 15,000 gpm. The existing scenarios are more limiting 
in that they rely on, at the most, two LPCI pumps. The actual 
testing of the pumps in accordance with the [inservice testing] IST 
program and Technical Specification 4.13 will not change. The 
testing of the pumps currently performed demonstrates that LPCI will 
function to mitigate the postulated accidents. Therefore, the 
elimination of the requirement to demonstrate that three pumps can 
deliver 15,000 gpm will not involve an increase in the probability 
or consequence of any previously evaluated accident.
    The elimination of a requirement to test the LPCI header 
instrumentation can not result in an increase to the probability or 
consequence of an accident, since no such instrumentation exists, or 
is required.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change will remove a requirement to perform a 
mathematical evaluation that provides no safety benefit. There is no 
change in the test methodology currently performed. All four LPCI 
pumps are tested. Deleting the requirement to verify that three LPCI 
pumps can produce 15,000 gpm flow does not create the possibility of 
a new or different type of accident.
    Deleting the requirement to test the LPCI spray header 
instrumentation can not create the possibility of a new or different 
kind of accident, since there is no LPCI header instrumentation. 
This change corrects an error which was introduced by an earlier 
License Amendment.
    3. Involve a significant reduction in the margin of safety.
    This change to the LPCI testing requirements does not change any 
of the actual testing, or individual component requirements which 
exist for the LPCI system. The change to remove the three pump, 
15,000 gpm flow requirement eliminates the need to calculate a value 
which provides no relevant information in ascertaining the ability 
of the LPCI system to perform its required safety function. The 
existing testing ensures performance of the LPCI subsystem in 
accordance with the accident analysis requirements. The intent of 
the Technical Specification Surveillance Requirement remains 
unchanged. Elimination of the requirement to test the LPCI header 
instrumentation corrects an error introduced in an earlier License 
Amendment. No LPCI header instrumentation exists, therefore, no 
credit was taken for such instrumentation in determining the margin 
of safety.
    This change can not involve a significant reduction in the 
margin of safety since there are no changes to the surveillance 
requirements for any of the individual components of the LPCI 
system.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment requests: October 17, 1994, as supplemented 
October 27, 1994
    Description of amendment requests: The proposed amendments would 
change the submittal frequency of the Radioactive Effluent Release 
Report from semiannual to annual in accordance with 10 CFR 50.36a.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated
    The proposed license amendments are requested to implement a 
revision to 10 CFR 50.36a. The requested amendment[s] does not alter 
any administrative controls over radioactive effluents, nor do they 
affect any accident evaluations. Also, the requested amendments do 
not involve any physical alterations to the plant with respect to 
radioactive effluents. The proposed changes would only affect the 
reporting requirements concerning routine data for radioactive 
effluents.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by any of the proposed 
amendments.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed
    The proposed license amendments are requested to implement a 
revision to 10C FR 50.36a. The requested amendment[s] does not alter 
any administrative controls over radioactive effluents, nor do they 
involve any physical alterations to the plant with respect to 
radioactive effluents. Also, the requested amendments do not change 
the method by which any safety-related system performs its function. 
The proposed changes would only affect the reporting requirements 
concerning routine data for radioactive effluents.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be 
created.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety
    The proposed license amendments are requested to implement a 
revision to 10 CFR 50.36a. The requested amendment[s] does not alter 
any administrative controls over radioactive effluents, nor do they 
involve any physical alterations to the plant with respect to 
radioactive effluents. The proposed changes would only affect the 
reporting requirements concerning routine data for radioactive 
effluents. The operation of systems and equipment remains unchanged.
    Therefore, a significant reduction in the margin of safety would 
not be involved.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: October 25, 1994
    Description of amendment request: The amendment would add to the 
Susquehanna Units 1 and 2 Technical Specifications, isolation signals 
to Table 3.6.3-1 for the containment isolation valves on the sample 
lines for the containment radiation monitoring (CRM) and wetwell sample 
lines. This change is based on the licensee's design change for 
installation of a new CRM and wetwell sample 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. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The addition of the new CRM and Wetwell Sample System does not 
affect any of the postulated initiating events identified in Chapter 
6 and 15 of the FSAR, the Design Assessment Report, the current 
Reload Analysis or the NRC Safety Evaluation Report (NUREG 0776).
    The new CRM and Wetwell Sample System with separate containment 
sample lines is isolated from the primary containment under accident 
conditions. The power and control-power to the CRM from the Class 1E 
Division I and Division II sources is through electrical isolation 
schemes so that failure(s) in the CRM under accident conditions is 
isolated from the Class 1E systems.
    The addition of a new CRM and Wetwell Sample System with 
separate sample lines and isolation valves does represent a change 
in the probability of occurrence of a malfunction of equipment. The 
addition of the auxiliary relay to the Division I and Division II 
CAC System containment isolation logic does represent the source of 
another potential malfunction in the logic due to the additional 
relay in the circuit. However, the increase in probability due to 
the additional relay is considered to be so small or insignificant 
that the change is within the error bounds associated with the 
original design calculations and does not constitute a significant 
increase in probability of the overall system malfunction.
    Thus, the addition of a new CRM and Wetwell Sample System does 
not significantly increase the probability of occurrence or the 
consequences of an accident or malfunction of equipment important to 
safety, as previously evaluated in the SAR.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Chapter 6 and 15 of the FSAR, the Design Assessment Report, the 
current Reload Analysis and NUREG-0776 were reviewed to determine if 
the proposed action had the potential of creating a postulated 
initiating event which was not within the spectrum of events which 
transient or anticipated operational occurrences and accident 
conditions were analyzed. The review did not identify a postulated 
initiating event which would create the possibility for an accident 
of a different type.
    A random single failure in the CRM A or CRM B does not create a 
malfunction of a different type. A random single failure in the 
existing containment isolation circuitry, the new isolation valve 
control circuitry or the new valve position indication circuitry for 
the new containment isolation valves does not create a malfunction 
of a different type. The consequences of random single failure of 
the CRM or the CRM and Wetwell containment isolation valve isolation 
signal, control and indication circuitry is the same as the existing 
consequences.
    Thus, the addition of a new CRM and Wetwell Sample System does 
not create a possibility for an accident or malfunction of a new or 
different type.
    3. Involve a significant reduction in a margin of safety.
    The operability of the primary containment isolation valves for 
the sample lines to the new CRMs and Wetwell Sample Rack is governed 
by Technical Specification Section 3/4.6.3 entitled ``Containment 
Systems, Primary Containment Isolation Valves'' with Table 3.6.3-1 
establishing the maximum isolation time. The bases for operability 
of the primary containment isolation valves is to ensure that the 
containment atmosphere is isolated from the outside environment in 
the event of a release of radioactive material to the containment 
atmosphere or pressurization of the containment. This is consistent 
with GDC 54 through 57 of 10 CFR 50, Appendix A. The bases for the 
containment isolation within the time limits specified in Table 
3.6.3-1 is for those isolation valves designed to close 
automatically to ensure that the release of radioactive material to 
the environment is consistent with the assumptions used in the 
analyses for a LOCA. The new CRM and Wetwell Sample Rack sample line 
isolation valves are solenoid valves which close immediately on an 
accident signal. The proposed action does not affect the operability 
requirements of Section 3/4.6.3. The margin of safety as defined in 
the Technical Specification for the containment isolation valves is 
not affected.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: November 11, 1994
    Brief description of amendments: The proposed amendment would 
modify the technical specifications (TS) by deleting accelerated 
testing and special reporting requirements for CPSES Units 1 and 2 
emergency diesel generators. These changes are based on Generic Letter 
94-01, ``Removal of Accelerated Testing and Special Reporting 
Requirements for Emergency Diesel Generators,'' dated May 31, 1994.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Deletion of the requirement for special reporting of EDG 
failures has no relation to probability or consequences of 
accidents. Therefore, deletion of the requirement for special 
reporting of EDG failures does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    There are no initiating events in accidents previously evaluated 
that involve testing of EDGs. Therefore, deletion of accelerated 
testing of EDGs does not involve a significant increase in the 
probability of an accident previously evaluated.
    A reduction in the number of test starts decreases EDG component 
stress and wear and decreases unavailability time for maintenance 
and pre and post run checks. The resulting change in EDG reliability 
and availability is an improvement toward ensuring the EDGs are 
capable of fulfilling their functional requirement to provide 
electric power for safe shutdown of the plant during loss of offsite 
power. Furthermore, implementation of the maintenance rule 
provisions for performance monitoring and root cause analysis for 
failures as a basis for establishing corrective actions establish an 
alternate reliability basis that is at least equivalent to that 
established by accelerated testing. Therefore, deletion of 
accelerated testing of EDGs does not involve a significant increase 
in the consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Deletion of the requirement for special reporting of EDG 
failures introduces no new failure modes for the EDGs or other plant 
systems and therefore has no relation to creation of accidents. 
Therefore, deletion of the requirement for special reporting of EDG 
failures does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The frequency at which EDG testing occurs does not affect the 
potential failure modes of the EDGs, which have already been 
assessed in the CPSES design. Therefore, deletion of accelerated 
testing of EDGs does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Acceptance limits and failure values are not affected by the 
requirement for special reporting of EDG failures. Therefore, 
deletion of the requirement for special reporting of EDG failures 
does not involve a significant reduction in a margin of safety.
    The margin of safety impact associated with accelerated testing 
relates to EDG reliability and availability. A reduction in the 
number of test starts decreases EDG component stress and wear and 
decreases unavailability time for maintenance and pre and post run 
checks. The resulting change in EDG reliability and availability is 
an improvement toward ensuring the EDGs are capable of fulfilling 
their functional requirement to provide electric power for safe 
shutdown of the plant during loss of offsite power. Furthermore, 
implementation of the maintenance rule provisions for performance 
monitoring and root cause analysis for failures as a basis for 
establishing corrective actions establish an alternate reliability 
basis that is at least equivalent to that established by accelerated 
testing. Therefore, deletion of accelerated testing of EDGs 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: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 11, 1994
    Brief description of amendments: The proposed amendment would 
provide for cycle-specific allowances to account for increases in the 
Heat Flux Hot Channel Factor between monthly surveillances.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes provide for the use of cycle-specific 
allowances to account for F2Qc(z) increases 
between surveillances. No hardware or setpoint changes are involved; 
therefore, the changes have no impact on the probability of 
occurrence of any accident previously analyzed.
    The proposed changes ensure that F2Qc(z) 
remains within its limit. Thus, the changes do not increase the 
consequences of any accident previously analyzed.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes provide for the use of a cycle-specific 
allowances to account for F2Qc(z) increases 
between surveillances. The proposed changes do not involve any 
hardware or setpoint changes. Therefore the changes do not create 
the possibility of a new or different kind of accident from any 
accident previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    The proposed changes do not affect the failure values of any 
system or any event acceptance criteria. Higher cycle-specific 
allowances ensure that remains below its limit between surveillances 
and within the bounds considered in the safety analyses. Therefore, 
the proposed changes do not involve a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, NW., Suite 1000, Washington, DC 20036
    NRC Project Director: William D. Beckner
    Wisconsin Public Service Corporation, Docket No. 50-305, 
Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin
    Date of amendment request: November 8, 1994
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
to allow application of a voltage-based repair limit for the steam 
generator (SG) tube support plate (TSP) intersections experiencing 
outside diameter stress corrosion cracking (ODSCC).
    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:
    This 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 
consequences of an accident previously evaluated.
    Testing of model boiler specimens for free span tubing (no TSP 
restraint) at room temperature conditions show burst pressures in 
excess of 5,000 psig for indications of ODSCC with voltage 
measurements as high as 19 volts. Burst testing performed on five 
intersections pulled from the Kewaunee SGs with up to a 2 volt 
indication showed measured tube burst in the range of 9,537 to 9,756 
psig. Burst testing performed on pulled tubes from other plants with 
up to 7.5 volt indications show burst pressures in excess of 6,300 
psi at room temperatures. Correcting for the effects of temperature 
on material properties and the minimum strength levels, tube burst 
capability significantly exceeds the safety factor requirements of 
RG 1.121.
    Tube burst criteria are inherently satisfied during normal 
operating conditions due to the presence of the TSPS. Test data 
indicates that tube burst cannot occur within the TSP, even for 
tubes with through wall EDM notches 0.75 inch long, when the notch 
is adjacent to the TSP. Since tube burst is precluded during normal 
operating conditions, the criterion that must be satisfied to 
demonstrate adequate tube integrity is a safety margin of 1.43 times 
MSLB pressure differential. From Figure 3-2 of EPRI report TR-
100407, the BOC structural limit for 7/8 inch diameter tubing is 9.6 
volts. Applying an allowance of 20% for NDE uncertainty and 50% for 
crack growth rate over an operating cycle results in a voltage 
repair limit of 5.6 volts. The proposed repair limit of 2 volts is 
very conservative when compared to the 5.6 volts taking into account 
the low average growth rates experienced at Kewaunee and the high 
tube burst pressures.
    Relative to the expected leakage during accident condition 
loadings, a plant specific calculation was performed to determine 
the maximum primary-to-secondary leakage during a postulated MSLB 
event. The evaluation considered both pre-accident and accident 
initiated iodine spikes. The results of the evaluation show that the 
accident spike yielded the limiting leak rate. This case was based 
on a 30 rem thyroid dose at the site boundary and initial primary 
and secondary coolant activity levels of 1.0 uCi/gm and 0.1 uCi/gm 
dose equivalent iodine-131, respectively. A leak rate of 34.0 gpm 
was determined to be the upper limit for allowable primary to 
secondary leakage in the SG in the faulted loop. The SG in the 
intact loop was assumed to leak at a rate of 0.1 gpm (150 gpd).
    Application of the voltage-based repair limit will be 
supplemented with a projected EOC MSLB leakage calculation and 
conditional burst probability assessment. The methodology for 
performing these calculations will be consistent with that discussed 
in the draft GL until final guidance is published. Should the 
projected MSLB leakage be exceeded indications will be repaired or 
removed from service until the projected leakage is less than or 
equal to 34.0 gpm.
    Application of the voltage-based repair limit will not adversely 
affect SG tube integrity. Therefore, the proposed amendment will not 
increase the probability or consequences of an accident previously 
evaluated.
    2) Create the possibility of a new or different kind of accident 
from any previously evaluated.
    Implementation of the proposed voltage-based repair limit will 
not reduce the overall safety or functional requirements of the SG 
tube bundles. The tube burst criteria will be satisfied during 
normal operating conditions by the presence of the TSPs. The RG 
1.121 criteria that must be satisfied during accident loading 
conditions is 1.43 times MSLB differential pressure. Conservatively, 
the existing data base of burst testing shows that the tube burst 
margins can be satisfied with bobbin coil signal amplitudes of about 
8.82 volts or less regardless of the depth of tube wall penetration.
    The proposed repair criteria will be supplemented with a reduced 
operating leakage requirement of 150 gpd average through either SG 
to preclude the potential for excessive leakage during operating 
conditions. The 150 gpd restriction will provide for timely 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. The operating leakage limit is 
based on leak-before break considerations, critical crack length and 
predicted leakage.
    The SG tube integrity will continue to be maintained through 
inservice inspections and primary-to-secondary leakage monitoring. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident.
    3) Involve a significant reduction in the margin of safety.
    Application of the voltage-based repair criteria has been 
demonstrated to maintain tube integrity commensurate with the RG 
1.121 criteria. RG 1.121 describes a method acceptable to the staff 
for meeting GDCs 2, 14, 15, 31 and 32. This is accomplished by 
determining the limiting degradation of SG tubing as established by 
inservice inspection, beyond which tubes should be removed from 
service. Upon implementation of the repair criteria, even under the 
worst case conditions, the occurrence of ODSCC at the TSPs is not 
expected to lead to a SG tube rupture event during normal or faulted 
conditions. The most limiting event would be a potential increase in 
leakage during a MSLB event. Excessive leakage during a MSLB is 
precluded by verifying that the expected EOC crack distribution of 
ODSCC indications at TSP locations would result in an acceptably low 
primary-to-secondary leakage. Therefore, the radiological 
consequences from tubes remaining in service is a small fraction of 
the 10 CFR 100 limits.
    The combined effects of a LOCA plus SSE on the SGs were assessed 
as required by GDC 2. This issue was addressed for the Kewaunee SGs 
through the application of leak-before-break (LBB) principles to the 
primary loop piping. Based on the results of this analysis, it is 
concluded that the LBB is applicable to the Kewaunee primary loops 
and, thus, the probability of breaks in the primary loop piping is 
sufficiently low that they need not be considered in the structural 
design basis of the plant. Excluding breaks in the primary loops, 
the LOCA loads from the large branch lines were also assessed and 
found to be of insufficient magnitude to result in SG tube collapse. 
Based on these analysis results, no tubes are expected to collapse 
or deform to the degree that secondary-to-primary in-leakage would 
be increased over currently expected levels. On this basis no tubes 
need to be excluded from the voltage-based repair criteria for 
reasons of deformation resulting from combined LOCA and SSE 
loadings.
    Addressing the RG 1.83 considerations, implementation of the 
voltage-based repair criteria will include a 100% bobbin coil probe 
inspection of all TSP intersections with known ODSCC down to the 
lowest cold leg TSP identified. This will be supplemented by a 
reduced operating leakage limit, enhanced eddy current data analysis 
guidelines, MRPC inspection requirements and a projected EOC voltage 
distribution. It is concluded that the proposed change will not 
result in a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: Leif J. Norrholm

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.

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

    Date of application for amendment: August 8, 1994
    Brief description of amendment request: The proposed amendment 
would modify Technical Specification (TS) 4.0.5.a. to delete the 
requirement to obtain prior written relief from the NRC for inservice 
inspection (ISI) and inservice testing (IST) of components conducted 
pursuant to 10 CFR 50.55a. This change would provide relief from the 
ASME Code requirement in the interim between the submittal of a relief 
request and the NRC's issuance of a safety evaluation regarding the 
relief request. The change would allow the plant to operate in 
accordance with a proposed relief request while the NRC staff completed 
its review of the relief request. The licensee has also proposed to 
modify TS 4.0.5.b. to add a definition for biennial or every-2-year 
inspection and testing activities. The definition of biennial or every 
2 years will be at least once per 731 days.Date of individual notice in 
Federal Register: November 14, 1994 (59 FR 56558)
    Expiration date of individual notice: December 14, 1994
    Local Public Document Room location:  Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
SteamElectric Plant, Unit No. 2, Darlington County, South 
CarolinaDate of application for amendment: June 29, 1994

    Brief description of amendment: The amendment deletes the 
requirement to perform alternate train testing to demonstrate that 
other, similar, safety-related components are operable when components 
are found, or made, inoperable in the safety injection, residual heat 
removal, and containment spray systems. The surveillance requirements, 
which the licensee refers to as accelerated testing requirements, 
affect the safety injection (SI) pumps, residual heat removal (RHR) 
pumps, containment spray (CS), SI, RHR and CS flow paths.
    Date of issuance: November 21, 1994
    Effective date: November 21, 1994
    Amendment No. 153
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39581) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 21, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College, Hartsville, South Carolina 29550

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

    Date of application for amendment: June 29, 1992, as supplemented 
February 22, 1994
    Brief description of amendment: The amendment revised TS Sections 
3/4.3, ``Instrumentation,'' 3/4.4.2, ``Safety/Relief Valves,'' and 
associated Bases to increase the surveillance test intervals and 
allowable out-of-service times for specific safety-related 
instrumentation.
    Date of issuance: November 22, 1994
    Effective date: November 22, 1994
    Amendment No. 67
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17605) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 22, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power 
Station,Units 1 and 2, Rock Island County, IllinoisDate of 
application for amendments: October 15, 1992

    Brief description of amendments: The proposed amendments would 
revise the Dresden and Quad Cities Technical Specification (TS) 3/4.4 
to revise the sodium pentaborate solution concentrations for the 
Standby Liquid Control System (SLCS) storage tanks based on net 
positive suction head test results.
    Date of issuance: November 16, 1994
    Effective date: November 16, 1994
    Amendment Nos.: 130, 124, 151, and 147
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 9, 1992 (57 FR 
58245) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 16, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: For Dresden, The Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; For Quad Cities, 
The Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

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

    Date of application for amendment: October 29, 1993, as 
supplemented on March 28, 1994, and November 8, 1994.
    Brief description of amendment: The amendment revises the 
surveillance intervals for the Volume Control Tank Level Instrument, 
the Containment High Range Radiation Monitors, the Safety Injection 
System Electrical Loading, the Safety Injection System, and the Reactor 
Coolant System Subcooling Margin Monitors to accommodate a 24-month 
fuel cycle. These revisions are being made in accordance wih the 
guidance provided by Generic Letter 91-04, ``Changes in Technical 
Specification Surveillance Intervals to Accommodate a 24-Month Fuel 
Cycle.''
    Date of issuance: November 16, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 178
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37067) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 16, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: August 25, 1994
    Brief description of amendments: The amendments revise the testing 
interval for auxiliary feedwater (AFW) system pumps from monthly to 
quarterly on a staggered test basis. The amendments are consistent with 
the guidance in NUREG-1366, ``Improvements to Technical Specifications 
Surveillance Requirements'' and Generic Letter 93-05, ``Line-Item 
Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation.'' In addition, a note 
is incorporated from NUREG-1431, ``Revised Standard Technical 
Specifications, Westinghouse Plants'' into the TS clarifying that the 
turbine-driven AFW pump cannot be tested until the required pressure 
exists in the secondary side of the steam generator.
    Date of issuance: November 9, 1994
    Effective date: November 9, 1994
    Amendment Nos.: 151 and 133
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49426) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 9, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223

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

    Date of application for amendment: March 19, 1993, augmented August 
18, 1994.
    Brief description of amendment: This amendment allows a reduction 
in Reactor Coolant System design flowrate from the current value of 
370,000 gpm to 355,000 gpm in Technical Specifications Figure 2.1-1 and 
Tables 2.2-1 and 3.2-1.

    Date of issuance: November 25, 1994
    Effective date: November 25, 1994
    Amendment No.: 130
    Facility Operating License No. DPR-67: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1993 (58 FR 
25855) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 25, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of amendment request: April 29, 1994, as supplemented by 
letter dated September 8, 1994.
    Brief description of amendments: The amendments revise the 
technical specifications to permit revision of the maximum allowable 
power range neutron flux high setpoint when one or more main steam 
safety valves are inoperable. In addition, new algorithm used to 
calculate the revised setpoint values is incorporated into the Bases 
for the technical specifications.
    Date of issuance: November 22, 1994
    Effective date: To be implemented within 30 days of issuance
    Amendment Nos.: Unit 1 - Amendment No. 66; Unit 2 - Amendment No. 
55
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29628) The additional information contained in the supplemental letter 
dated September 8, 1994, was clarifying in nature and, thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated November 22, 1994.No significant hazards consideration 
comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

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

    Date of applications for amendment: June 4, 1993, as supplemented 
February 4, 1994, and May 6, 1994.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TS) by changing the requirements of the TS Section 3.6, 
``Primary Systems Boundary,'' adding definitions into Section 1.0, 
``Definitions,'' and revising Bases Section 3/4.6. These changes 
improved clarity and provided consistency of the TS with the Standard 
TS (NUREG-1202). Typographical and administrative corrections were also 
made in Section 3.6.
    Date of issuance: November 17, 1994
    Effective date: November 17, 1994, and to be implemented within 120 
days
    Amendment No.: 203
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39052). The May 6, 1994, application, repeated a TS change included in 
the June 4, 1993, application, and proposed changes to the TS Bases. 
The information in the February 14, 1994, supplement, did not change 
theinitial no significant hazards determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated November 17, 1994. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

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

    Date of application for amendment: July 12, 1994
    Brief description of amendment: The proposed amendment changes the 
requirement to perform the surveillance test for the channel functional 
test Rod Block Monitor, Flow-biased Average Power Range Monitor and 
Recirculation Flow instruments from within 24 hours prior to startup to 
after the reactor is in the RUN mode, but prior to when each system is 
assumed to function in the plant safety analysis.
    Date of issuance: November 18, 1994
    Effective date: November 18, 1994, to be implemented within 30 days 
of issuance
    Amendment No.: 204
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45025) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 18, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401

South Carolina Electric & Gas Company, South Carolina Public 
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of application for amendment: October 29, 1993, as 
supplemented March 11, 1994, May 18, 1994, September 20, 1994, and 
October 20, 1994
    Brief description of amendment: The proposed changes support the 
installation of new steam generators at Summer Station. The changes 
involve:
    (1) alterations to the core operating limits
    (2) changes to various reactor trip setpoints
    (3) deletion of the negative flux rate trip
    (4) removal of references to specific correlations used in the 
departure from nucleate boiling (DNB) analyses
    (5) changes to the steam/feedwater flow mismatch activation 
specification
    (6) changes to shutdown limits
    (7) changes to instrument uncertainty allowances
    (8) a change to the methodology for reactor coolant system (RCS) 
flow determination
    (9) modifications to DNB parameters
    (10) a change to the engineered safety features actuation system 
setpoints for steam generator water levels
    (11) removal of the F* and L* criteria (12) addition of a 
requirement for a first inservice inspection for the new steam 
generators
    Date of issuance: November 18, 1994
    Effective date: November 18, 1994
    Amendment No.: 119
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7968) The May 18, 1994, September 20, 1994, and October 20, 1994 
submittals contained explanatory information and did not change finding 
of nos significant hazards consideration as published in the FEDERAL 
REGISTER. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 18, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

South Carolina Electric & Gas Company, South Carolina Public 
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of application for amendment: July 20, 1994
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) to modify TS Table 2.2-1, Reactor Trip System 
Instrumentation Setpoints, and Table 3.3-4, Engineered Safety Features 
Actuation System Instrumentation Trip Setpoints and several associated 
bases. The change would remove specific rack and sensor allowable drift 
values by removing three columns from the tables.
    Date of issuance: November 18, 1994
    Effective date: November 18, 1994
    Amendment No.: 120
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 19, 1994 (59 
FR 47181) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 18, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

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 5, 1993, as supplemented 
by letter dated September 22, 1994
    Brief description of amendments: These amendments propose to revise 
Technical Specification (TS) 3/4.7.1.1, ``Main Steam Safety Valves,'' 
and the associated Bases to (1) increase the as-found setpoint 
tolerance of Table 3.7-1 for the Main Steam Safety Valves (MSSVs) from 
+/- 1 percent to +2 percent and -3 percent; (2) add a footnote to Table 
3.7-1 to indicate that the setpoint tolerance for the lowest set pair 
of MSSVs will be +1 percent and -3 percent; (3) add a footnote to TS 
3.7.1.1 and revise footnote 1 of Table 3.7-1 to clarify that the MSSVs 
will be left at the lift setting according to Table 3.7-1 within a +/-1 
percent tolerance following inservice testing; (4) add an ACTION 
statement requiring the plant to be in HOT STANDBY within 6 hours and 
in HOT SHUTDOWN within the following 12 hours for the case of less than 
five MSSVs operable per operable steam generator; (5) require the plant 
to be in ``HOT SHUTDOWN within the following 12 hours'' instead of 
``COLD SHUTDOWN within the following 30 hours'' per the existing ACTION 
statement; (6) revise the title of column 1 of Table 3.7-2 to read 
``Number of Operable Safety Valves per Operable Steam Generator,'' 
instead of ``Maximum Number of Inoperable Safety Valves on Any 
Operating Steam Generator, for better readability; and (7) delete the 
ORIFICE SIZE column of Table 3.7-1.
    Date of issuance: November 23, 1994
    Effective date: As of the date of its issuance and must be fully 
implemented no later than 30 days from the date of issuance.
    Amendment Nos.: 114 and 103
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34093) The additional information contained in the supplemental letter 
dated September 22, 1994, served to clarify the amendments, was within 
the scope of the initial notice, and did not affect the Commission's 
proposed no significant hazards consideration determination.The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated November 23, 1994.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713

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

    Date of application for amendments: September 9, 1994 (TS 94-08)
    Brief description of amendments: The amendments add the main steam 
valve vaults to the exclusion areas where containment penetration 
integrity is not required to be verified once every 31 days for 
penetrations that are secured in the closed position.
    Date of issuance: November 22, 1994
    Effective date: November 22, 1994
    Amendment Nos.: 191 and 183
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51630) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 22, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: July 14, 1994
    Brief description of amendments: These amendments modify the 
current Technical Specifications having cycle-specific parameter limits 
in the Core Operating Limits Report.
    Date of issuance: November 15, 1994
    Effective date: November 15, 1994Amendment Nos. 194 and 194
    Facility Operating Licen se Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51630) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 15, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

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

    Date of application for amendment: May 17, 1994
    Brief description of amendment: The amendment revises the TS by 
separating the specification for the Internal Containment Spray (ICS) 
and the Spray Additive Systems into two distinct specifications. The 
amendment also removes the requirement that for a spray train to be 
operable, a spray pump suction flow path from the additive tank is 
needed. In addition, the allowable out-of-service time for the Spray 
Additive System is increased from 48 hours to 72 hours.
    Date of issuance: November 18, 1994
    Effective date: November 18, 1994
    Amendment No.: 113
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37090) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 18, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Dated at Rockville, Maryland, this 30th day of November 1994.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV,Office of Nuclear 
Reactor Regulation
[FR Doc. 94-29925 Filed 12-6-94; 8:45 am]
BILLING CODE 7590-01-F