[Federal Register Volume 59, Number 119 (Wednesday, June 22, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10622]


[[Page Unknown]]

[Federal Register: June 22, 1994]


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UNITED STATES 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 May 27, 1994, through June 10, 1994. The 
last biweekly notice was published on June 8, 1994 (59 FR 29623).

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, 11555 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 
20555. The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By July 22, 1994, 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 20555 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 20555, 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 20555, and at the local public document 
room for the particular facility involved.

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: April 26, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to change the Table 3.5-1 High 
Containment Pressure ( Hi Level), Safety Injection Setting Limit from 
less than or equal to 2.0 psig to less than or equal to 5.0 psig.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    It is proposed that the High Containment Pressure (Hi Level) 
actuation setting of [less than or equal to] 2.0 psig be revised to 
[less than or equal to] 5.0 psig. This additional operating 
flexibility will decrease the frequency of Containment venting 
necessary to relieve containment of
    non-condensible gases which build up during normal operation.
    Based upon a statistical analysis of the containment pressure 
channel uncertainty for a 30 month operating cycle, a margin must be 
allowed between the Technical Specification limit (plant setting) 
and the Safety Analysis limit so that the Safety Analysis limit(s) 
will not be exceeded under the worst circumstances. For a Technical 
Specification value of [less than or equal to] 5.0 psig, the 
corresponding Safety Analysis limit must be increased to 10 psig to 
provide margin for the channel statistical allowance. A safety 
evaluation performed pursuant to 10 CFR 50.59 is on file which 
supports a change in the Safety Analysis limit from 7.3 psig 
(current value) to 10.0 psig. Key conclusions of the Safety 
Evaluation are that neither the probability nor the consequences of 
an accident or malfunction of equipment important to safety 
previously evaluated in the Safety Analysis report would be 
increased.
    Thus, assurance is provided that appropriate protective actions 
in accordance with the Technical Specifications will be taken so 
that Safety Analysis limits are not exceeded.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The proposed change in the Technical Specification limit 
together with the change in the Safety Analysis limit provides 
adequate margin to accommodate instrument channel uncertainty over a 
30 month operating cycle. Plant equipment, which would be set at the 
Technical Specification limit, will therefore provide protective 
functions to assure that safety analysis limits are not exceeded. 
This would prevent the possibility of a new or different kind of 
accident from that previously evaluated from occurring.
    3. There has been no reduction in the margin of safety.
    The proposed change to the Technical Specification limit would 
decrease the frequency of containment purges necessary to vent the 
build up of non-condensible gases during normal operation. This 
would result in a decrease in the amount of radioactivity discharged 
to the environment (due to decay), decrease the potential for high 
Containment pressure alarms and increase the margin for an ESF trip. 
The change to the Safety Analysis limits, justified by a safety 
Evaluation performed in accordance with 10 CFR 50.59, assures 
sufficient margin exists to accommodate channel instrument 
uncertainty over the maximum operating cycle length. This margin is 
necessary so that safety functions will occur and Safety Analysis 
limits will be preserved.
    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 L. Boyle

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

    Date of amendment request: May 24, 1994
    Description of amendment request: The proposed amendments would 
transfer the boron concentration in Technical Specification (TS) 3.9.1 
for the reactor coolant system and the refueling canal during MODE 6, 
and the boron concentration in TS 3.9.12 for the spent fuel pool from 
the TS to the Core Operating Limits Report (COLR). The application is 
submitted in response to the guidance in Generic Letter 88-16 which 
addresses the transfer of fuel cycle-specific parameter limits from the 
TS to the COLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The following analysis, performed pursuant to 10 CFR 50.91, 
shows that the proposed amendment will not create a significant 
hazards consideration as defined by the criteria of 10 CFR 50.92.
    1. This amendment will not significantly increase the 
probability or consequence of any accident previously evaluated.
    No component modification, system realignment, or change in 
operating procedure will occur which could affect the probability of 
any accident or transient. The relocation of boron concentration 
values to the COLR is an adminsitrative change which will have no 
effect on the probability or consequences of any previously-analyzed 
accident. The required values of boron concentration will continue 
to be determined through use of approved methodologies.
    2. This amendment will not create the possibility of any new or 
different accidents not previously evaluated.
    No component modification or system realignment will occur which 
could create the possibility of a new event not previously 
considered. The administrative change of relocating parameters to 
the COLR, in this case boron concentration, cannot create the 
probability of an accident.
    3.This amendment will not involve a significant reduction in a 
margin of safety.
    Required boron concentrations will remain appropriate for each 
cycle, and will continue to be calculated using approved 
methodologies. There is no 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: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: David B. Matthews, Director

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

    Date of amendment request: May 12, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification Sections 3.1 and 4.1 for Protective 
Instrumentation, the associated bases, and tables to increase the 
surveillance test intervals (STIs) and add allowable out-of-service 
times (AOTs). All proposed STI and AOT changes are in accordance with 
General Electric Company Licensing Topical Reports (LTRs) which have 
been previously reviewed and approved by the NRC staff. Also, AOTs are 
clarified in accordance with the most recently approved BWR Owners' 
Group letters which were used in the development of NUREG-1433 
``Standard Technical Specifications, General Electric Plants, BWR/4.'' 
The Technical Specification changes will permit specified Channel Tests 
to be conducted quarterly rather than weekly or monthly. The amendment 
will enhance operational safety by reducing 1) the potential for 
inadvertent plant scrams, 2) excessive test cycles on equipment, and 3) 
the diversion of plant personnel and resources on unnecessary testing.
    Two additional technical changes are proposed. The first change 
involves extending the Channel Calibration interval for average power 
range monitor (APRM) scram instrumentation from weekly to quarterly. 
GPUN has evaluated the effect of drift on the setpoint over the longer 
interval for this instrumentation and found it to be acceptable. The 
second change would add a quarterly Channel Calibration requirement for 
High Drywell Pressure (for Core Cooling) and Turbine Trip Scram 
instrumentation. This would be a new requirement not currently 
incorporated in the Technical Specifications.
    Nineteen editorial changes have been incorporated in 
Instrumentation Sections 3.1 and 4.1 to provide clarity and 
consistency. These items are editorial only and do not alter the 
meaning or intent of any requirements. Examples of editorial changes 
are: 1) capitalize definitions where used, 2) punctuation and 
grammatical corrections, 3) ensuring consistency in STI nomenclature, 
and 4) reformat of tables. A table note and its associated footnote 
were deleted which involved a 1985 licensing condition which is no 
longer applicable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analyses of the issue of no significant hazards 
consideration, which is presented below:
    NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION OF TECHNICAL 
CHANGES
    1. The operation of the Oyster Creek Nuclear Generating Station, 
in accordance with the proposed amendment, will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The generic analysis contained in LTR NEDC-30851P-A assessed the 
impact of changing RPS [reactor protection system] STIs and adding 
AOTs on RPS failure frequency, scram frequency and equipment 
cycling. Specifically, Section 5.7.4, ``Significant Hazards 
Assessment,'' of NEDC-30851P-A states that:
    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 [core damage frequency] 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 times is negligible.
    The BWR Owners' Group concluded that the proposed changes do not 
significantly increase the probability or consequences of an 
accident previously evaluated since the increase in probability of a 
scram failure due to RPS unavailability is insignificant. The 
overall probability of an accident is decreased as the time RPS 
logic operates as designed is increased resulting in less 
inadvertent scrams during testing and repair. The plant-specific 
evaluation performed by GPUN and GE demonstrates that while the 
Oyster Creek RPS differs from the generic model analyzed in the RPS 
LTR (NEDC-30851P-A), the net effect of the differences do not alter 
the generic conclusions. The AOTs proposed for RPS instrumentation 
are based on improved wording developed for use in NUREG 1433, 
``Standard Technical Specifications, General Electric Plants, BWR/
4,'' which ensures a loss of function does not occur. In addition, 
the change to the APRM Scram Channel Calibration surveillance 
interval from weekly to quarterly has been evaluated by GPUN to 
determine the effect on setpoint drift. The results of the 
evaluation show acceptable performance of this scram parameter 
ensuring that the safety analysis remains valid. The clarification 
that a Channel Calibration is not applicable to Turbine Trip Scram 
instrumentation is appropriate since this trip parameter senses 
turbine stop valve position via limit switches which are fixed in 
position and adjusted, as necessary, during valve maintenance. This 
trip parameter and its switch adjustment methods are similar to the 
Main Steamline Isolation Valve [MSIV] Scram for which the Technical 
Specifications require only a Channel Test.
    LTR NEDC-30936P-A (Parts 1 and 2) contains an assessment of the 
impact of changing STIs and AOTs for BWR ECCS Actuation 
Instrumentation. Section 4.0, ``Technical Assessment of Changes,'' 
of NEDC-30963P-A (Part 2) states that:
    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 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 
evaluation performed by GPUN and GE demonstrates that while the 
Oyster Creek ECCS differs from the generic model analyzed in LTR 
NEDC-30936P-A, the net effect of the plant-specific differences do 
not alter the generic conclusions. The addition of a quarterly 
Channel Calibration STI for the High Drywell Pressure ECCS 
initiation parameter is consistent with the calibration interval 
requirement for other similar instrumentation at Oyster Creek and 
ensures the regular performance of calibrations. This is a new 
requirement not currently contained in the Technical Specifications 
and experience performing the High Drywell Pressure (Core Cooling) 
instrument calibration at a quarterly interval has proven adequate 
for instrument performance monitoring.
    LTRs NEDC-30851P-A, Supplement 2 and NEDC-31677P-A contain 
generic analyses assessing the impact of changing STIs and AOTs for 
BWR Isolation Actuation Instrumentation which are common or not 
common to RPS and ECCS instrumentation. Section 4.0, ``Summary of 
Results,'' of NEDC-30851P-A, Supplement 2 states that:
    The results indicated 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 indicated 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 allowed 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 that:
    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.
    Based on this generic analysis, the BWR Owners' Group concluded 
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. The proposed 
wording of the AOTs is based on the clarifications used in the 
development of NUREG 1433, ``Standard Technical Specifications, 
General Electric Plants, BWR/4,'' which ensures a loss of function 
does not occur where applied to isolation actuation instrumentation.
    LTR NEDC-30851P-A, Supplement 1 contains a generic analysis 
assessing the impact of changing control rod block STIs on Rod Block 
failure frequency. Section 5 (Brookhaven National Laboratory 
Technical Evaluation Report - Attachment 2 to the NRC SER) of NEDC-
30851P-A, Supplement 1 states that:
    The BWROG 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 form 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.
    Bases contained in GE Topical Report GENE-770-1-A assessed the 
impact of changing STIs and AOTs on failure frequency for selected 
systems. Section 2.0, ``Summary,'' of GENE-770-06-1-A states that:
    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.
    The BWR Owners' Group concluded from the generic analysis in 
NEDC-30851P-A, Supplement 1 and the bases in GENE-770-06-1-A that 
the proposed changes do not significantly increase the probability 
or consequences of an accident previously evaluated. GPUN's 
utilization of GENE-770-06-1-A is limited to the identified AOTs for 
Control Rod Block instrumentation analyzed in NEDC-30851P-A since 
the Control Rod Block LTR did not explicitly address AOTs.
    2. The operation of Oyster Creek Nuclear Generating Station, in 
accordance with the proposed amendment, will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The addition of allowable out-of-service times (AOTs) consistent 
with wording developed for use in Improved Standard Technical 
Specifications to ensure no loss of function and the revision of 
surveillance test intervals (STIs) does not alter the function of 
RPS, ECCS, Isolation or Rod Block instrumentation nor involve any 
type of plant modification. No new modes of plant operation are 
involved with the changes.
    Adding a quarterly Channel Calibration STI for High Drywell 
Pressure instrumentation (for Core Cooling) establishes a 
requirement in the Technical Specifications which is not currently 
incorporated. This is an additional requirement beyond that already 
in place for this instrumentation and will not alter its operation 
since by their nature STIs ensure proper instrument performance. The 
clarification that a Channel Calibration is not applicable to 
Turbine Trip Scram instrumentation is appropriate since this trip 
parameter senses turbine stop valve position via limit switches 
which are fixed in position and adjusted during valve maintenance. 
This trip parameter and its switch adjustment methods are similar to 
the Main Steamline Isolation Valve Scram for which the Technical 
Specifications require only a Channel Test. Revising the Channel 
Calibration STI for APRM Scram instruments from weekly to quarterly 
allows these instruments to benefit from the Channel Test STI change 
provided by the generic analysis in the RPS LTR. The benefits 
include a significant reduction in the number of half-scram states 
the plant will undergo reducing the potential for inadvertent plant 
trips. The effect of setpoint drift over the longer interval has 
been evaluated and found acceptable.
    The proposed changes will not alter the physical characteristics 
of any plant systems or components and all safety-related systems 
and components remain within their applicable design limits. Thus, 
system and component performance is not adversely affected by these 
changes, thereby assuring that the design capabilities of those 
systems and components are not challenged in a manner not previously 
assessed so as to create the possibility of a new or different kind 
of accident.
    3. The operation of the Oyster Creek Nuclear Generating Station, 
in accordance with the proposed amendment, will not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed and approved the generic studies 
contained in the GE Licensing Topical Reports and has concurred with 
the BWR Owners' Group that the proposed changes do not significantly 
affect the availability of RPS, ECCS Actuation, Isolation Actuation 
and Control Rod Block instrumentation. The proposed addition of 
allowable out-of-service times for instruments addressed by the LTRs 
provides reasonable times for making repairs and performing tests. 
The lack of sufficient out-of-service time provided in current 
Technical Specifications, increases the potential for an inadvertent 
scram or equipment actuation. The proposed AOTs provide realistic 
times to complete required actions without increasing overall 
instrument failure frequency and ensure that no loss of function 
occurs, therefore, there is no significant reduction in the margin 
of safety.
    The LTRs demonstrate that extending surveillance test intervals 
does not result in significant changes in the probability of 
instrument failure. Where Channel Calibration frequency has not 
changed, assurance exists that setpoints will not be affected by 
drift. In the case of the APRM Scram Channel Calibration, the 
proposed change to quarterly from weekly has been evaluated and 
found acceptable. Expected instrument performance over the extended 
interval will assure that applicable safety analyses will continue 
to be met. In addition, other instrumentation was evaluated for 
drift effects of setpoints and was found acceptable. The addition of 
a quarterly Channel Calibration interval for High Drywell Pressure 
(for Core Cooling) is consistent with Channel Calibration STIs for 
most other instrumentation at Oyster Creek and has been the interval 
used to achieve an adequate level of instrument performance 
monitoring. The clarification that a Channel Calibration is not 
applicable to Turbine Trip Scram instrumentation ensures consistency 
in the establishment of surveillance requirements. This trip 
parameter senses turbine stop valve position via limit switches 
which are fixed in position and adjusted during valve maintenance. 
This trip parameter and its switch adjustment methods are similar to 
the Main Steamline Isolation Valve Scram for which the Technical 
Specifications require only a Channel Test. These proposed changes, 
when coupled with the reduced probability of test-induced plant 
transients and equipment failures, do not result in a reduction in 
the margin of safety.
    No Significant Hazards Consideration Evaluation For Editorial 
Changes
    The above nineteen proposed changes are editorial in nature and 
are typical example I.c.2.e.i in 51FR7744. Therefore, they do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The editorial changes described above do not change the design 
or operation of any structure, system or component relied upon to 
prevent or mitigate the consequences of any accident evaluated. 
These editorial changes also do not add new structures, systems or 
components which may have an effect on existing elements of the 
facility. The changes proposed correct, clarify and/or retain 
existing requirements.
    2. Create the possibility of a new or different kind of accident 
form any accident previously evaluated.
    Since neither physical changes to the facility nor changes in 
its operation are involved in the proposed editorial changes to the 
Technical Specifications, there is no possibility for creation of a 
new or different kind of accident.
    3. Involve a significant reduction in the margin of safety.
    Facility configuration and operation are unaffected by the 
proposed editorial changes. As a result no changes in margin of 
safety occur.
    The editorial changes described and evaluated above are purely 
administrative to achieve consistency or correct an error in the 
Technical Specifications.
    The NRC staff has reviewed the licensee's analyses and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied for both the technical issues and editorial changes. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, New Jersey 
08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: February 10, 1994
    Description of amendment request: The revision proposed by 
Technical Specification Change Request (TSCR) No. 230 to the Technical 
Specifications would revise specification 3.7.2.c, ``Unit Electric 
Power System,'' to eliminate testing of an emergency diesel generator 
(EDG) when the redundant EDG is inoperable.
    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 serves to assure that an EDG is always 
available to perform on demand and the lower number of demands for 
performance reduce the probability of equipment failure. The 
required action no longer requires a ``test'' be performed. 
Therefore, the word ``test'' has been deleted from TS 3.7.2.c. The 
change is administrative. Since the proposed amendment does not 
affect the design or performance of the diesel generators or their 
ability to perform their design function, the change will not result 
in an increase in the consequences or probability of an accident 
previously analyzed. The proposed change will increase diesel 
generator reliability, thereby increasing overall plant safety.
    2. Operation of the facility in accordance with the proposed 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. Accidents 
involving loss of off-site power and single failure have been 
previously evaluated. The change does not introduce any new mode of 
plant operation or new accident precursors, involve any physical 
alterations to plant configurations, or make any changes to system 
setpoints which could initiate a new or different kind of accident.
    3. Operation of the facility in accordance with the proposed 
amendment does not involve a significant reduction in a margin of 
safety. This change does not result in a reduction in the margin of 
safety since there is no margin of safety associated with the 
supplemental immediate and daily testing of the operable EDG. If a 
margin of safety were presumed to exist, no reduction would result 
because of the proposed amendment: no physical modification to the 
plant or change to procedurally prescribed operator actions resulted 
from the proposed amendment.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 25, 1994
    Description of amendment request: The proposed amendment would 1) 
allow entry through an operable personnel air lock hatch to perform 
surveillance testing, repair an inoperable hatch, or perform other 
necessary activities inside containment, 2) update plant Technical 
Specifications to reflect a previous change to the list of containment 
boundary valves, 3) add a new exception to allow quarterly surveillance 
testing of the excess flow check valves, 4) add a new exception to 
allow periodic preventive maintenance on control room ventilation 
lasting up to 30 minutes per calendar quarter without a written report 
of such inoperability, and 5) make related administrative changes to 
reflect and clarify items 1 through 4 above.
    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 amendment would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Containment air lock hatch entry, surveillance testing of the 
excess flow check valves, and preventive maintenance of control room 
ventilation are of short duration and do not alter any associated 
remedial action completion times, or the requirements of Technical 
Specification 3.0.A. If necessary, prompt operator action to restore 
containment integrity, excess flow check valve position, or control 
room ventilation is assured by plant operators, or individual(s) 
procedurally dedicated to perform such restoration. The subject 
containment boundary valves are manual containment isolation valves, 
and the current specification allows them to be repositioned under 
administrative control without compensatory measures to isolate the 
penetration. The boundary valves to be added remain closed during 
power operation, and are opened only after the reactor is shut down 
and cooldown has begun. The boundary valves to be deleted are open 
only during plant heatup.
    The staff therefore concludes that implementation of the proposed 
change will not involve a significant increase in the probability or 
consequences of an 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 proposed change to containment air lock hatch entry, 
surveillance testing of the excess flow check valves, and preventive 
maintenance of the control room ventilation system, will not affect 
equipment reliability when such equipment is required to be 
operable. The Limiting Conditions for Operation and Remedial Actions 
for these items remain unchanged to govern operability of the 
equipment. The containment boundary valves being added are closed 
when the reactor is at power, and are opened only after the reactor 
is shut down. The boundary valves being deleted are open only during 
plant heatup. The subject boundary valves are manual containment 
isolation valves, and the current specification allows them to be 
repositioned under administrative control without compensatory 
measures to isolate the penetration.
    The staff therefore concludes that implementation of the 
proposed change will not create any new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change would allow excess flow check valves to be 
exercised through approximately 1.5 inches of valve travel on a 
quarterly basis without declaring the valves inoperable or taking 
compensatory measures. Such testing constitutes approximately 15 
minutes per calendar quarter, during which time containment 
isolation can easily be reestablished. Similarly, access through an 
operable air lock hatch would allow the hatch to be open for only a 
short period of time and while under control of an individual 
dedicated to operating the hatch. The proposed change also permits 
the control room ventilation system to be inoperable for 30 minutes 
per calendar quarter, without a written report of such 
inoperability. Because of the short time during which these systems 
are unavailable, and because operation is easily reestablished, 
there is no significant reduction in a margin of safety. The 
containment boundary valves being added are closed when the reactor 
is at power, and are opened only after the reactor is shut down. The 
boundary valves being deleted are open only during plant heatup. The 
subject boundary valves are manual containment isolation valves, and 
the current specification allows them to be repositioned under 
administrative control without compensatory measures to isolate the 
penetration.
    The staff therefore concludes that implementation of the proposed 
change would 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, 83 Edison Drive, Augusta, Maine 04336
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: April 29, 1994
    Description of amendment request: The amendment would change the 
requirement for reactor operators (RO) in Table 6.2-1 from 2 to 3 for 
the RUN, STARTUP/HOT STANDBY and HOT SHUTDOWN conditions. In addition, 
two typographical corrections are made to page 6-4.
    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:
    NNECO has reviewed the proposed changes in accordance with 
10CFR50.92 and concluded that the changes do not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed changes do not involve a significant 
hazards consideration because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    Accident analyses for Millstone Unit No. 1 do not require a 
specific number of operators. Increasing the Technical Specification 
minimum to require a third RO does not decrease the effectiveness of 
the shift staff in response to normal or abnormal conditions. In 
fact, the third RO enhances the ability of the operating crew to 
mitigate complex transients which could occur during beyond design 
basis events. The shifts have trained and functioned at the higher 
staffing level for several years.
    The typographical corrections to page 6-4 provide a clearer 
representation of the required actions, and do not affect the intent 
nor implementation of the specification.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of a previously analyzed 
accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The addition of a third RO to the minimum shift-crew composition 
required by Technical Specification Table 6.2-1 does not affect the 
operation of the unit, nor does it change any of the operating 
procedures, off-normal procedures, or EOPs [emergency operating 
procedures]. Staffing the control room with an additional operator 
enhances the capability of the operating crew to mitigate 
transients. Therefore, addition of a third RO to the minimum shift-
crew composition cannot create the possibility of a new or different 
accident.
    3. Involve a significant reduction in the margin of safety.
    The proposed addition of a third reactor operator is to ensure 
that sufficient operating staff is available to respond to complex 
transients involving multiple equipment failures. Ensuring that 
sufficient resources are available to cope with beyond design basis 
event scenarios provides an increase in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 6, 1994
    Description of amendment request: The proposed amendment would 
modify the Limiting Conditions for Operation (LCO) for the Millstone 
Unit 2 Technical Specifications 3.8.2.3 and 3.8.2.4 and the 
surveillance requirement of TS 4.8.2.3.2.c.3. These changes relate to 
the amperage requirements and the charging capability of the DC 
distribution systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes do not involve an SHC [significant hazards 
consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    In 1993, revised battery and battery charger sizing calculations 
demonstrated that a charger capacity of 400 amperes is sufficient to 
provide the continuous DC loads, and is capable of recharging a 
fully discharged station battery in a timely manner consistent with 
the design basis discussed in Section 8.5.3.1 of the Millstone Unit 
No. 2 FSAR [Final Safety Analysis Report]. The calculations 
determined that the largest continuous load was 154 amperes; 
therefore, 400 amperes of charging capacity could provide 246 
amperes to recharge a battery.
    The calculations conservatively demonstrated that this charging 
capacity could recharge a battery in 10.37 hours. This recharging 
time is well within the 12-hour recharging time discussed in Section 
8.5.3.1 of the Millstone Unit No. 2 FSAR. Additionally, this 
recharging time is more conservative than the 24-hour recharging 
time stated in Section 8.3.2 of the original Safety Evaluation for 
Millstone Unit No. 2. Therefore, the proposed changes do not involve 
a significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed LCO and surveillance changes do not alter the 
existing DC bus configuration, as described in Section 8.5.3.1 of 
the Millstone Unit No. 2 FSAR. This bus configuration has been 
previously analyzed, and was found acceptable. The proposed changes 
also meet the recharging time specified in the design basis. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    In 1993, revised battery and battery charger sizing calculations 
demonstrated that a charger capacity of 400 amperes is sufficient to 
provide the continuous DC loads, and is capable of recharging a 
fully discharged station battery in a timely manner consistent with 
the design basis discussed in Section 8.5.3.1 of the Millstone Unit 
No. 2 FSAR. The calculations determined that the largest continuous 
load was 154 amperes; therefore, 400 amperes of charging capacity 
could provide 246 amperes to recharge a battery. The calculations 
conservatively demonstrated that this charging capacity could 
recharge a battery in 10.37 hours. This recharging time is well 
within the 12-hour recharging time discussed in Section 8.5.3.1 of 
the Millstone Unit No. 2 FSAR. Additionally, this recharging time is 
more conservative than the 24-hour recharging time stated in Section 
8.3.2 of the original Safety Evaluation for Millstone Unit No. 2. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 6, 1994
    Description of amendment request: The proposed amendment would 
provide additional Technical Specification requirements regarding non-
Quality Assurance (QA) equipment utilized to achieve feedwater 
isolation in response to a main steam line break (MSLB) inside 
containment. Specifically the amendment would incorporate additional 
sections numbered 3/4.7.1.6, titled ``Plant Systems - Main Feedwater 
Isolation Components (MFICs);'' 3/4.8.2.1A, titled '' Onsite Power 
Distribution Systems - A.C. Distribution - Operating;'' and 3/4.8.2.5, 
titled ``Onsite Power Distribution Systems (Turbine Battery) - D.C. 
Distribution - Operating.'' In addition, the proposed amendment would 
modify the Index and the Bases to reflect the additional requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed changes in accordance with 
10CFR50.90 and has concluded that the changes do not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(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.
    Currently, the Millstone Unit No. 2 Technical Specifications 
contain response time requirements for the feedwater isolation 
valves to ensure rapid isolation of feedwater to the steam 
generators and to maintain the peak containment pressure below the 
containment design pressure of 54 psig. However, clear Action 
Statements specifying operability requirements for the non-QA 
equipment associated with feedwater isolation are not included 
within the Millstone Unit No. 2 Technical Specifications. NNECO's 
proposal to add sections 3/4.7.1.6, 3/4.8.2.1A, and 3/4.8.2.5 into 
the Millstone Unit No. 2 Technical Specifications will incorporate 
additional requirements regarding components that are credited to 
provide feedwater isolation in the event of an MSLB inside 
containment. These proposed changes will impose additional 
limitations, restrictions, and controls not currently in place in 
the Millstone Unit No. 2 Technical Specifications.
    Additionally, NNECO's proposals to modify the Bases and the 
Index of the Millstone Unit No. 2 Technical Specifications will: 1) 
provide personnel with information concerning the additional 
requirements, and 2) correct an editorial error. These proposed 
changes to the Bases and the Index do not alter the manner in which 
equipment is operated, nor do they affect equipment availability.
    Based on the above, the proposed license amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    NNECO's proposal to add section 3/4.7.1.6, 3/4.8.2.1A, and 3/
4.8.2.5 into the Millstone Unit No. 2 Technical Specifications will 
incorporate additional requirements regarding components that are 
credited to provide feedwater isolation in the event of an MSLB 
inside containment. These proposed changes will impose additional 
limitations, restrictions, and controls not currently in place in 
the Millstone Unit No. 2 Technical Specifications.
    Additionally, NNECO's proposals to modify the Bases and the 
Index of the Millstone Unit No. 2 Technical Specifications will: 1) 
provide personnel with information concerning the additional 
requirements, and 2) correct an editorial error. These proposed 
changes to the Bases an the Index do not alter the manner in which 
equipment is operated, nor do they affect equipment availability.
    Based on the above, the proposed license amendment cannot create 
the possibility of a new or different kind of accident from any 
previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    NNECO's proposal to add sections 3/4.7.1.6, 3/4.8.2.1A, and 3/
4.8.2.5 into the Millstone Unit No. 2 Technical Specifications will 
incorporate additional requirements regarding components that are 
credited to provide feedwater isolation in the event of an MSLB 
inside containment. These proposed changes will impose additional 
limitations, restrictions, and controls not currently in place in 
the Millstone Unit No. 2 Technical Specifications.
    Additionally, NNECO's proposals to modify the Bases and the 
Index of the Millstone Unit No. 2 Technical Specifications will: 1) 
provide personnel with information concerning the additional 
requirements, and 2) correct an editorial error. These proposed 
changes to the Bases and the Index do not alter the manner in which 
equipment is operated, nor do they affect equipment availability.
    Therefore, this proposed license amendment does not involve a 
significant reduction in a margin of safety. In fact. The margin of 
safety will be increased due to the imposition of restriction on the 
non-QA equipment credited for feedwater isolation in the event of an 
MSLB inside containment.
    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, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 6, 1994
    Description of amendment request: The proposed amendment modifies 
the monthly operational test of the reactor trip bypass breakers to 
monthly staggered, such that each breaker is tested every 62 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes do not involve an SHC [significant hazards 
consideration] because the changes would not:
    1. Involve a significant increase in probability or consequences 
of an accident previously evaluated.
    Revising the technical specifications to require a staggered 
monthly surveillance operational test of the reactor trip bypass 
breakers (such that each breaker is tested every 62 days) will only 
make operational testing of the reactor trip bypass breakers 
consistent with operational testing of the trip breakers and the 
automatic trip and interlock logic. It will also reduce cycling of 
the reactor trip bypass breakers by eliminating the requirement to 
test both bypass breakers during the monthly surveillance, thereby 
reducing maintenance and surveillance time. The proposed changes do 
not affect any of the design basis accidents nor are there any 
malfunctions associated with these changes.
    Additionally, this technical specification bases change only 
clarifies both the meaning of a reactor trip breaker and trip 
breaker train which have been included for completeness and clarity 
concerning the reactor trip breaker system.
    2. Create the possibility of a new or different kind of accident 
previously evaluated.
    Revising the technical specifications to require a staggered 
monthly surveillance operational test of the reactor trip bypass 
breakers (such that each breaker is tested every 62 days) will only 
make operational testing of the reactor trip bypass breakers 
consistent with operational testing of the reactor trip breakers and 
the automatic trip and interlock logic. There are no new failure 
modes associated with the proposed changes. Since the plant will 
continue to operate as designed, the proposed changes will not 
modify the plant response to the point where it can be considered a 
new accident.
    3. Involve a significant reduction in a margin of safety.
    Revising the technical specifications to require a staggered 
monthly surveillance operational test of the reactor trip bypass 
breakers (such that each breaker is tested ever 62 days) will only 
make operational testing of the reactor trip bypass breakers 
consistent with operational testing of the reactor trip breakers and 
the automatic trip and interlock logic. It will also reduce cycling 
of the reactor trip bypass breakers by eliminating the requirement 
to test both bypass breakers during the monthly surveillance, 
thereby reducing maintenance and surveillance time. The proposed 
changes do not have any adverse impact on the protective boundaries 
nor do they affect the consequences of any accident previously 
analyzed. The surveillance requirements will still ensure that the 
reactor trip breakers and the reactor trip bypass breakers are 
tested and within the limits. Therefore, the proposed changes will 
not impact the margin of safety as designated in the bases of any 
technical specification.
    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, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 18, 1994
    Description of amendment request: The amendment would change 
operability requirements for the Fuel Building Exhaust Filter System to 
require it to be operable whenever irradiated fuel is in the spent fuel 
pool, which has had less than 60 days of decay time. Surveillance 
requirements for the Fuel Building Exhaust Filter System would be 
changed to require that the system be tested and verified operable at 
no greater than 31 days prior to its required usage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ...The proposed change does not involve an SHC [significant 
hazards consideration] because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed modification will revise the period of time during 
which the Fuel Building Exhaust Filter System must be operable.
    The propose[d] change will require that the system is operable 
whenever irradiated fuel, which has decayed less than 60 days, is in 
the spent fuel pool. Currently, the system is required to be 
operable whenever a load is moved over the pool or fuel is being 
moved in the pool.
    The modification has no effect on the probability of a fuel 
handling accident. The consequences of a fuel handling accident has 
been evaluated at two intervals. The first time is the minimum decay 
time. At this time (t=100 hours) with irradiated fuel in the pool, 
the Fuel Building Exhaust Filter System is required, per the 
existing and the proposed Technical Specification, to be operable. 
Therefore, the consequences of an accident are identical to that 
described in the FSAR [Final Safety Analyses Report]. The second 
scenario evaluated is when the filters are initially isolated (t=60 
days). The resultant offsite dose, assuming no filtration and lower 
core inventory due to decay, are significantly lower than was 
calculated at t=100 hours. Therefore, the existing accident analysis 
in FSAR Section 15.7.4 is limiting and the proposed modification 
will not impact the probability or consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed change does not impact any system or component 
which could cause a fuel handling accident. The Fuel Building 
Exhaust Filter System is used for accident mitigations. It's failure 
cannot, in any way, create the possibility of a new or different 
kind of accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the Fuel Building Exhaust Filter System 
has been analyzed at the two most critical times. The first analysis 
was done when the fuel is first placed in the pool, and the second 
analysis was done when the filtration system is isolated. The first 
event resulted in no change in assumptions in the analysis presented 
in the FSAR, ergo no change in dose. The second event has been 
analyzed and doses have decreased, when compared to the first event. 
The system will be verified operable per the performance of 
Surveillance Requirement 4.9.12a prior to fuel or load movement over 
the pool. Therefore, there is no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 13, 1994
    Description of amendment request: This amendment would revise 
Technical Specifications Surveillance Requirement 4.8.1.1.2e.8, which 
requires that an emergency diesel generator be retested within 5 
minutes after completing a 24-hour endurance run.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed TS change would revise the Emergency Diesel 
Generator (EDG) surveillance criteria to allow the hot restart test 
to be performed independent of the Engineered Safety Features (ESF) 
load sequencing test and the 24 hour endurance run. The proposed 
surveillance requirements would continue to demonstrate that the 
objectives of each of these tests are met. Specifically, the EDG's 
are shown to be capable of starting the ESF loads in the required 
sequence, operating at full load for an extended period of time, and 
restarting from a full load temperature condition. Therefore, the 
proposed changes would not adversely affect the EDG's ability to 
support mitigation of the consequences of any previously evaluated 
accident. The proposed changes to the surveillance requirements do 
not affect the initiation or progression of any accident sequence.
    Therefore, the proposed change does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This proposed TS change does not require physical changes to the 
plant or equipment, and does not impact any design or functional 
requirements of the Emergency Diesel Generators (EDGs). The proposed 
change affects surveillance test criteria such that increased 
scheduling flexibility is allowed while the test objectives 
associated with demonstrating EDG operability continue to be met. 
The proposed changes do not allow any plant configurations that are 
presently prohibited by the Technical Specifications.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The proposed TS change does not involve a change to the physical 
design or functional requirements of the Emergency Diesel Generators 
(EDGs). Surveillance testing in accordance with the proposed 
Technical Specification will continue to demonstrate the ability of 
the EDG's to perform their intended function of providing electrical 
power to ESF systems needed to mitigate design basis transients, 
consistent with the plant safety analyses. The margin of safety 
demonstrated by the plant safety analyses is therefore not affected 
by the proposed change.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Charles L. Miller

Philadelphia Electric Company, Public Service Electric and Gas 
Company,Delmarva Power and Light Company, and Atlantic City 
Electric Company,Docket No. 50-277, Peach Bottom Atomic Power 
Station, Unit No. 2,York County, Pennsylvania

    Date of application for amendment: May 13, 1994
    Description of amendment request: The proposed amendment would 
extend the Type A test (i.e., Containment Integrated Leak Rate Test) 
interval on a one-time basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The accidents which are potentially negatively impacted by the 
proposed change are any Loss of Coolant Accident (LOCA) inside 
primary containment as described in the PBAPS [Peach Bottom Atomic 
Power Station], Units 2 and 3 UFSAR [Updated Final Safety Analysis 
Report].
    The proposed change increases the surveillance interval of the 
10 CFR [Part] 50, Appendix J Type A test (i.e., Containment 
Integrated Leakage Rate Test (CILRT)) from 42 months to 66 months. 
This test is performed to determine that the total leakage from 
containment does not exceed the maximum allowable primary 
containment leakage rate (i.e., designated La) at a calculated peak 
containment internal pressure (Pa), as defined in 10 CFR [Part] 50, 
Appendix J. The primary containment limits the leakage of 
radioactive material during and following design bases accidents in 
order to comply with the offsite dose limits specified in 10 CFR 
[Part] 100. Accordingly, the primary containment is not an accident 
initiator, it is an accident mitigator. No physical or operational 
changes to the containment structure, plant systems, or components 
would be made as a result of the proposed change. Therefore, the 
probability of occurrence of an accident previously evaluated is not 
increased.
    The failure effects that are potentially created by the proposed 
one-time TS change have been considered. The relevant components 
important to safety which are potentially affected are the 
containment structure, plant systems, and containment penetrations. 
There are no physical or operational changes to any plant equipment 
associated with the proposed TS change. Therefore, the probability 
or consequences of a malfunction of equipment important to safety is 
not increased.
    The proposed change introduces the possibility that primary 
containment leakage in excess of the allowable value (i.e., La) 
would remain undetected during the proposed 24 month extension of 
the interval between the second and third Type A test. The types of 
mechanisms which could cause degradation of the primary containment 
can be categorized into two types. These are: 1) degradation due to 
work which is performed as part of a modification or maintenance 
activity on a component or system (i.e., activity-based), or; 2) 
degradation resulting from a time-based failure mechanism.
    A review of activity-based failure mechanisms has determined 
that the potential from degradation due to activity based mechanisms 
is minimal.
    Regarding the potential for primary containment degradation due 
to a time-based mechanism, we have concluded that the PBAPS Local 
Leak Rate Test (LLRT) program would identify most types of 
penetration leakage. The LLRT program involves measurement of 
leakage from Type B and Type C primary containment penetrations as 
defined in 10 CFR [Part] 50, Appendix J.
    The 10 CFR [Part] 50, Appendix J, Type B tests are intended to 
detect local leaks and to measure leakage across pressure containing 
or leakage-limiting boundaries other than valves, such as 
containment penetrations incorporating resilient seals, gaskets, 
expansion bellows, flexible seal assemblies, door operating 
mechanism penetrations that are part of the containment system, 
doors, and hatches. 10 CFR [Part] 50, Appendix J, Type C testing is 
intended to measure reactor system primary containment isolation 
valve leakage rates. The frequency of the Type B and Type C testing 
is not being altered by the proposed TS change. [However, in an 
April 18, 1994 letter, the licensee has requested a 60-day extension 
of the Type B and Type C testing.] The acceptance criterion for Type 
B and Type C leakage is 0.6 La (i.e., 0.3 % wt/day) which, when 
compared to the Type A test acceptance criterion of 0.75 La (i.e., 
0.375 % wt/day), is a significant portion of the Type A test 
allowable leakage.
    The proposed TS change only extends the interval between two 
consecutive Type A tests. The Type B and Type C tests will be 
performed as required. The Type B and Type C tests will continue to 
be used to confirm that the containment isolation valves and 
penetrations have not degraded. Containment system components that 
would not be tested are the containment structure itself and small 
diameter instrumentation lines. Time-based degradation of any of the 
instrumentation lines would most likely be identified by faulty 
instrument indication or during instrument calibrations that will be 
performed during the PBAPS, Unit 2 refueling outage 10. In examining 
the potential for a time-based failure mechanism that could cause 
significant degradation of the containment structure, we concluded 
that the risk, if any, of such a mechanism is small since the design 
requirements and fabrication specifications established for the 
containment structure are in themselves adequate to ensure 
containment leak tight integrity.
    Based on the above evaluation, we have concluded that the 
proposed TS change will have a negligible impact on the consequences 
of any accident previously evaluated. To support this conclusion, a 
review of the PBAPS, Unit 2 CILRT history was performed. This review 
identified that the only failure mechanism that has been detected 
during the past CILRTs is an activity based component failure, and 
that there is no indication of any time-based degradation that would 
not be identified during performance of Type B and Type C tests.
    Although this review concluded that the risk of undetected 
primary containment degradation is not increased, the Individual 
Plant Examination (IPE) for PBAPS, Units 2 and 3, was also reviewed 
in order to assess the impact of exceeding the primary containment 
allowable leakage rate, if a non-mechanistic activity type (i.e., 
time-based) failure were to occur. The IPE included an evaluation of 
the effect of various containment leakage sizes under different 
scenarios. The IPE results showed that a containment leakage rate of 
35% wt/day would represent less than a 5% increase in risk to the 
public being exposed to radiation. This evaluation was based on a 
study performed by Oak Ridge National Laboratory for light water 
reactors that evaluated the impact of leakage rates on public risk. 
As stated earlier, the current value of La for PBAPS, Unit 2, is 
0.5% wt/day, which is significantly less than the 35% wt/day 
discussed in the IPE evaluation.
    Therefore, the proposed TS change involving a one-time extension 
of the Type A test interval and performing the third Type A test 
after the second Appendix J 10-year service period will not involve 
an increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change is an increase of a surveillance test 
interval and does not make any physical or operational changes to 
existing plant systems or components. Primary containment acts as an 
accident mitigator not initiator. Therefore, the possibility of a 
different type of accident than any previously evaluated or the 
possibility of a different type of equipment malfunction is not 
introduced.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.
    The total primary containment leakage rate ensures that the 
total containment leakage volume will not exceed the value assumed 
in the safety analyses at the peak accident pressure. As an added 
conservatism, the measured overall leakage rate is further limited 
to less than or equal to 0.75 La during performance of periodic 
tests to account for possible degradation of the containment leakage 
barriers between leakage tests. There is the potential that 
containment degradation could remain undetected during the proposed 
24 month surveillance interval extension and result in the 
containment leakage exceeding the allowable value assumed in safety 
analysis. A review of activity-based failure mechanisms has 
determined that the potential from degradation due to activity based 
mechanisms is minimal.
    Regarding the potential for primary containment degradation due 
to a time-based mechanism, we have concluded that the PBAPS Local 
Leak Rate Test (LLRT) program would identify most types of 
penetration leakage. The LLRT program involves measurement of 
leakage from Type B and Type C primary containment penetrations as 
defined in 10 CFR [Part] 50, Appendix J.
    The 10 CFR [Part] 50, Appendix J, Type B tests are intended to 
detect local leaks and to measure leakage across pressure containing 
or leakage-limiting boundaries other than valves, such as 
containment penetrations incorporating resilient seals, gaskets, 
expansion bellows, flexible seal assemblies, door operating 
mechanism penetrations that are part of the containment system, 
doors, and hatches. 10 CFR [Part] 50, Appendix J, Type C testing is 
intended to measure reactor system primary containment isolation 
valve leakage rates. The frequency of the Type B and Type C testing 
is not being altered by the proposed TS change.
    Finally, a review of the results of previous PBAPS, Unit 2 CILRT 
results concluded that the only failure mechanism which has been 
detected during the past CILRTs is activity-based and that there is 
no indication of time-based failures that would not be identified 
during performance of Type B and Type C tests. Therefore, we have 
concluded that the proposed extended test interval would not result 
in a non-detectable PBAPS, Unit 2 primary containment leakage rate 
in excess of the allowable value (i.e., 0.5% wt/day) established by 
the TS and 10 CFR [Part] 50, Appendix J.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Charles L. Miller

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 amendment request: March 11, 1994
    Description of amendment request: The proposed amendment would 
reduce the allowed outage time for the residual heat removal (RHR) 
suction relief valves (SRVs) in accordance with the guidance of Generic 
Letter (GL) 90-06.
    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 probability or consequences of an accident previously 
evaluated is not significantly increased.
    This change decreases the allowed outage time of a Low 
Temperature Overpressure Protection (LTOP) system. There is no 
hardware, software, or operating methodology change, so there is no 
increase in probability or consequences. Since the time allowed for 
one train of this equipment to be inoperable is shorter, the 
probability of an overpressure event not being mitigated has also 
been reduced. The consequences will not change unless the system or 
operation of the system changes.
    2. [The proposed change will not] [c]reate the possibility of a 
new or different kind of accident from any previously analyzed.
    As this proposed change will not involve any changes to 
hardware, software, or operating practices, it cannot create any 
possibility of new or different kinds of accidents from those 
previously analyzed. The RHR SRVs are intended to provide protection 
against a rupture of a pressure boundary from an over-pressure 
condition which has the potential to result in core uncovery. The 
original design basis of the plant complies with the requirements of 
10 CFR 50 Appendix G and uses the RHR SRVs to meet the fracture 
toughness requirements of 10 CFR 50 Appendix G. This change only 
increases the availability of this protection and does not create 
any new or different kinds of accidents.
    3. [The proposed amendment does not] [i]nvolve a significant 
reduction in a margin of safety.
    SCE&G already has administrative controls in place to minimize 
the possibility of an overpressure event occurring as well as to 
assure that there are two trains of LTOP equipment operable during 
the modes when the potential exists for this event. There are 
controls to preclude the inadvertent start-up of a Reactor Coolant 
Pump or Charging Pump and controls to ensure that both RHR Suction 
Isolation Valves for each train are open and remain open except for 
testing and maintenance. This alignment is maintained until the RHR 
System is realigned for its ECCS function. These controls are 
proceduralized in plant operating procedures.
    This change does not involve a significant reduction in a margin 
of safety as nothing is changed which affects the margin in a 
negative direction. The decrease in AOT actually increases the 
margin since the allowed time for one train to be inoperable has 
been reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: William H. Bateman

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

    Date of amendment request: May 16, 1994 (TS 94-03)
    Description of amendment request: The proposed change would remove 
Table 3.3-2, ``Reactor Trip System Instrumentation Response Times,'' 
and Table 3.3-5, ``Engineered Safety Features Response Times,'' from 
the technical specifications and incorporate the limits into the 
Updated Final Safety Analysis Report. In addition, references to these 
tables in Specifications 3.3.1.1, 3.3.2.1, and 4.3.1.1.3 (for Unit 1) 
and 3.3.1, 3.3.2, and 4.3.1.1.3 (for Unit 2) would be removed. A 
footnote would be added to Specification 4.3.1.1.3 indicating that 
neutron detectors are exempt from response time testing. These changes 
have been proposed in accordance with Generic Letter 93-08. A change to 
the Bases would indicate that the response time limits would be 
maintained in the Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not alter the response time limit 
requirements for the reactor trip or engineered safety feature 
actuation systems or surveillance testing and frequency. Placing 
these limits in the Updated Final Safety Analysis Report (UFSAR) 
will ensure the plant design basis is maintained in accordance with 
10 CFR 50.59. Since no actual changes to response time limits or 
surveillance requirements are involved, the probability or 
consequences of an accident are not increased.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed changes does not affect any plant equipment, 
functions, or setpoint by relocating response time limits to the 
UFSAR. Therefore, the possibility of a new or different kind of 
accident is not created.
    3. Involve a significant reduction in a margin of safety.
    The proposed change will continue to require SQN to maintain the 
plant functions at the required setpoints necessary for the design 
basis and to support the accident analysis. The margin of safety is 
not reduced because there is no change to plant functions and the 10 
CFR 50.59 process will continue to ensure the plant design basis is 
appropriately maintained.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: May 18, 1994 (TS 94-05)
    Description of amendment request: The proposed change would add a 
note to the action statement for Limiting Condition for Operation 
3.7.7, ``Control Room Emergency Ventilation System,'' indicating that 
the provisions of TS 3.0.3 are not applicable while performing actions 
associated with a tornado warning.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The control room emergency ventilation system (CREVS) was 
designed to ensure control room habitability during accident 
conditions. The design basis of SQN does not include an accident 
creating a contaminated air condition concurrent with a tornado. The 
ability of the CREVS to perform its design function has not been 
affected by this change. The proposed change will not increase the 
possibility or consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    An accident involving a contaminated air condition and a tornado 
have been analyzed as part of the SQN design basis. Both accidents 
are assumed to occur independently. This change does not create a 
new or different accident not previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The design basis of the CREVS is not impacted by this TS change. 
There is no change in any assumptions made in the Final Safety 
Analysis Report. Therefore, there is no reduction in the margin of 
safety as a result of this change.
    The NRC 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: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: March 19, 1994; superseded May 16, 1994 
(TS 93-04)
    Description of amendment request: The proposed change would clarify 
and consolidate the technical specifications (TS) regarding the dual 
function of the containment vacuum relief system (i.e., the vacuum 
relief and containment isolation functions). The proposed changes would 
revise TS 3/4.6.6, ``Vacuum Relief Valves,'' to indicate the actions 
that would be required should one or more vacuum relief (VR) lines be 
incapable of performing its containment isolation function or incapable 
of performing its VR function. In addition, the testing requirements 
would be revised to add specific requirements and reflect the inservice 
test (IST) program by relocating the testing requirements from TS 
4.6.3.2.d and Table 3.6-2 to the new TS 4.6.6 (and to Sequoyah's IST 
program). Other proposed changes affect Bases 3/4.6.6 section and TS 
index pages to reflect the proposed changes indicated above. This 
proposed change was originally noticed on May 12, 1993 (58 FR 28060), 
which is superseded by this notice.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    TVA's proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed change does not increase the probability of an accident 
since the proposed change does not affect any plant systems, 
equipment, or components. The dual design functions of SQN's 
containment vacuum relief (VR) system (i.e., provide containment VR 
and containment isolation) are not affected. The consequences of an 
event are not significantly increased by the proposed increase in 
allowed outage time from 4 hours to 72 hours for returning an 
inoperable VR system to operable status. The probability of an event 
during the relatively short duration of the TS completion times, in 
conjunction with the redundancy provided in the design of the 
system, provide sufficient assurance that the VR lines are available 
for mitigating an accident or abnormal event.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    No physical modification is being made to any plant hardware or 
plant operating setpoints, limits, or operating procedures as a 
result of this change. TVA's proposed change provides a TS 
improvement that clarifies the TS requirements associated with the 
dual design function of SQN's VR system. The proposed change removes 
the potential for creating a conflict between Specification 3/4.6.3, 
``Containment Isolation Valves,'' and Specification 3/4.6.6, 
``Vacuum Relief Valves.''
    The proposed change does not alter any accident analysis or any 
assumptions used to support the accident analyses. The containment 
leakage assumptions used to determine offsite dose limits for 
compliance with 10 CFR 100 are not affected. The analysis that 
supports the containment VR system also remains unchanged. The 
proposed 72-hour and 1-hour completion times for returning an 
inoperable VR line to operable status are consistent with the NUREG-
1431 and NUREG-1433. Consequently, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety provided by the design of SQN's containment 
VR system remains unchanged. TVA's proposed change does not affect 
the VR function or the containment isolation function that currently 
exists in SQN TSs. The proposed change eliminates the potential for 
conflicting requirements within SQN TSs and ensures that the proper 
action is taken to preserve these dual design functions while the 
plant is in Modes 1, 2, 3, or 4. TVA's proposed change provides a TS 
improvement that combines these functional requirements into a 
single specification. Both VR and containment isolation requirements 
will continue to be provided. Accordingly, the proposed change does 
not involve a reduction in the margin of safety.
    The NRC 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: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: February 14, 1993
    Brief description of amendments: The proposed amendments would 
revise the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2, 
technical specifications (TS) by (1) changing the allowable value for 
Unit 2 overtemperature N-16 and pressurizer pressure-low setpoints, (2) 
deleting Equation 2.2-1 from TS 2.2.1, and (3) administrative changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of a previously evaluated 
accident.
    Overtemperature N-16, Unit 2
    Incorporation of the increased temperature uncertainties 
reported by Rosemount will change the Allowable Value of the 
Overtemperature N-16 trip function. The change does not affect the 
Safety Analysis Limits assumed in the accident analysis. Because the 
change only impacts the Allowable Value for a setpoint and does not 
affect any system designs or operations, the change does not 
increase the probability of an accident. Although the Allowable 
Value is changed in the conservative direction, the change assures 
that, considering the newly identified transmitter uncertainty, the 
trip actuates prior to the conditions assumed in the accident 
analyses. As such, there is no impact on the consequences of any 
accidents previously evaluated.
    Pressurizer Pressure - Low, Unit 2
    The added uncertainties change the Allowable Value of the Unit 2 
Pressurizer Pressure-Low Reactor Trip function. The change does not 
affect the Safety Analysis Limits assumed in the accident analysis. 
Because the change only impacts the Allowable Value for a setpoint 
and does not affect the system design or operations, the change does 
not increase the probability of an accident. Although the Allowable 
Value is changed in the conservative direction, the change assures 
that, considering the newly identified transmitter uncertainty, the 
trip actuates prior to the conditions assumed for the accident 
analyses. As such, there is no impact on the consequences of any 
accidents previously evaluated.
    Equation 2.2-1
    The changes to Specifications 2.2.1 and 3.3.2, to Tables 2.2-1 
and 3.3-3, and to the bases sections will require recalibration of 
the channel and removal of any accumulated errors in any function 
whose ``as found'' setpoint is found to be less conservative than 
its allowable value. These changes delete a potentially less 
conservative option and will result in actual channel operation 
closer to the nominal setpoint and within the allowable value band. 
These changes will in effect validate one of the assumptions made in 
the accident analysis and will not increase the probability or 
consequences of any accident evaluated in the Safety Analysis 
Report.
    Administrative Changes
    The changes to combine the Unit 1 and Unit 2 line items into a 
dual Unit line if the Trip Setpoint and Allowable Value values are 
the same is administrative and meant as a human factors improvement 
for operator convenience. The change does not affect the operation 
of any equipment, the operating point of any equipment, nor any 
equipment hardware and thus does not increase the probability or 
consequences of any accident evaluated in the Safety Analysis 
Report.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed accident.
    Overtemperature N-16 and Pressurizer Pressure - Low, Unit 2
    As the proposed amendment changes only the Unit 2 Allowable 
Values of the Overtemperature N-16 reactor trip and the Pressurizer 
Pressure-Low reactor trip and does not have any physical effect on 
the transmitter or circuitry, there are no new or different types of 
accident introduced.
    Equation 2.2-1
    Deletion of this equation and its associated action statements, 
definitions and values does not introduce any physical changes to 
any systems, structures, or components. The change merely assures 
that setpoints which are less conservative than their Allowable 
Value are recalibrated prior to being declared operable. These 
changes do not introduce any new credible failure modes which may 
create the possibility of a new or different accident.
    Administrative Changes
    Combining line items for Unit 1 and Unit 2 into a dual Unit 
entry for administrative purposes does not introduce any new 
credible failure modes which may create the possibility of a new or 
different accident.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    Overtemperature N-16 and Pressurizer Pressure - Low, Unit 2
    Incorporation of the added temperature uncertainties of the 
Rosemount transmitters assures that the safety analysis limits 
assumed in the accident analyses for Overtemperature N-16 and 
Pressurizer Pressure-Low reactor trip functions for Unit 2 are met. 
There is no change in the acceptance criteria or the results of 
these analyses due to this change. Thus there is no effect on the 
margin of safety.
    Equation 2.2-1
    Deletion of Equation 2.2-1, related actions and associated 
definitions and values, merely eliminates one option to assure that 
the safety analysis assumptions are met. This option is not 
presently in use and the accident analyses assumptions have been and 
will continue to be met using the other option (to re-calibrate 
channels prior to restoring operability). Thus the margin of safety 
is unaffected.
    Administrative Changes
    Combining the Unit 1 and Unit 2 line items of Table 2.2-1 for 
RTS [Reactor Trip Systems] functions and of Table 3.3-3 for ESFAS 
[Engineered Safety Features Actuation System] functions into dual 
unit entries does not change the Trip Setpoint or the Allowable 
Value for the functions. The margin of safety is unaffected.
    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, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: April 19, 1994
    Description of amendment request: The proposed amendment revises 
Technical Specification 6.2.2.g to reflect a title designation change 
within the Wolf Creek Nuclear Operating Corporation (WCNOC) 
organization. The title of Supervisor Operations is being changed to 
Assistant Manager Operations. The title change does not represent any 
change in reporting relationships, job responsibilities, or overall 
organizational commitments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
This change involves an administrative change to the WCNOC 
organization and to the position title and as such has no effect on 
plant equipment or the technical qualification of plant personnel.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any previously evaluated. This 
change is administrative in nature and does not involve any change 
to installed plant systems or the overall operating philosophy of 
Wolf Creek Generating Station.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety. This change does not involve any changes in 
overall organizational commitments. A position title change alone 
does not reduce any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: Theodore R. Quay

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

    Date of amendment request: April 19, 1994
    Description of amendment request: The proposed amendment revises 
Technical Specification Table 3.6-1, ``Containment Isolation Valves,'' 
by deleting reference to two (2) valves. The Technical Specification 
change reflects a planned modification which removes the essential 
service water (ESW) containment air cooler return line isolation valve 
bypass valves and associated piping.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    After the design modification is completed the ESW Containment 
Penetrations will be provided with stainless steel isolation valves, 
which will be provided with automatic SIS [safety injection signal] 
actuation signals to open automatically to provide required cooling 
water flow to the Containment Air Coolers following a LOCA [loss-of-
coolant accident] or MSLB [main steamline break]. Replacement of the 
current carbon steel isolation valves with stainless steel valves 
and removing the unnecessary bypass lines and bypass isolation 
valves will reduce the amount of seat leakage currently experienced 
with these valves.
    The probability of occurrence of a previously evaluated accident 
is not increased because this modification does not introduce any 
new potential accident initiating conditions. The consequences of an 
accident previously evaluated is not increased because the ability 
of containment to restrict the release of any fission product 
radioactivity to the environment will not be degraded by this 
modification.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed modification will reduce the number of containment 
isolation valves and replace several carbon steel isolation valves 
with stainless steel valves, which will be less susceptible to 
erosion and corrosion. Thus, potential system leakage will be 
reduced by this modification, while valve reliability will be 
enhanced. The new valves are designed to the original ESW System 
requirements, and removal of the bypass lines and bypass isolation 
valves will not result in a malfunction of any other plant 
equipment. Therefore, this proposed modification will 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 removal of the bypass lines and bypass isolation valves will 
not adversely affect containment isolation capability for credible 
accident scenarios. Due to a previous design change, the bypass 
lines are no longer required to ensure adequate cooling flow to the 
Containment Air Coolers. In addition, the operability and 
reliability of the remaining isolation valves will be enhanced by 
replacing the current carbon steel valves with stainless steel 
valves.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: Theodore R. Quay

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.

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

    Date of amendment request: May 11, 1994
    Brief description of amendment request: The amendment would allow 
reduced power operation as a function of reactor coolant system (RCS) 
total flow rate for flow rate reductions of up to 5 percent below the 
currently specified flow rate. Operation will be allowed at total flow 
rates slightly lower than (293,540 gpm X (1.0 plus C1)) if rated 
thermal power (RTP) is reduced by 1.5 percent for each one percent that 
RCS total flow is less than this rate. This change would provide for 
needed operational margin and flexibility without the unnecessary 
penalty of a large power reduction.
    Date of publication of individual notice in Federal Register: May 
25, 1994 (59 FR 27079)
    Expiration date of individual notice: June 24, 1994
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of amendment request: May 27, 1994
    Description of amendment request: The amendment would add a new 
section to Technical Specification Section 6.17 and would require that 
procedures be in place to provide for monitoring and sampling of 
emergency service water (ESW) discharge flow during accident conditions 
when a positive differential pressure cannot be maintained between ESW 
and low pressure coolant injection (LPCI) in the LPCI heat exchangers.
    Date of publication of individual notice in Federal Register: June 
7, 1994 (59 FR 29448)
    Expiration date of individual notice: July 7, 1994
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

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 20555, and at the local public document 
rooms for the particular facilities involved.

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

    Date of application for amendments: February 18, 1994, as 
supplemented by letter dated May 16, 1994
    Brief description of amendments: These amendments modify Technical 
Specification (TS) Figure 3.2-1, ``REACTOR COOLANT COLD LEG vs CORE 
POWER LEVEL,'' of TS 3/4.2.6, ``REACTOR COOLANT COLD LEG TEMPERATURE,'' 
for Units 1 and 3 to include the cold leg temperature between 552 deg.F 
and 562 deg.F at core power levels between 90 percent and 100 percent 
within the AREA OF ACCEPTABLE OPERATION. Also, the proposed amendments 
modify TS 3/4.1.1.4, ``MINIMUM TEMPERATURE FOR CRITICALITY,'' and BASES 
3/4.1.1.4, ``MINIMUM TEMPERATURE FOR CRITICALITY,'' to allow the 
minimum temperature for criticality to be established at 545 deg.F, 
rather than the current value of 552 deg.F, to establish the 
surveillance temperature at 552 deg.F, rather than the current 
557 deg.F, and to clarify the BASES for this TS.
    Date of issuance: June 7, 1994
    Effective date: NPF-41 and NPF-51, prior to startup from the next 
refueling outage; NPF-74, no later than 45 days from the date of 
issuance.
    Amendment Nos.: 77, 63, and 49
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14886) The additional information contained in the May 16, 1994, letter 
was clarifying in nature, was within the scope of the initial notice, 
and did not affect the NRC staff's proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated June 7, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

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

    Date of application for amendments: April 14, 1994, as supplemented 
on May 16, 1994.
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) to relocate the Instrument Response Time 
Tables to the Updated Final Safety Analysis Report in accordance with 
NRC Generic Letter 93-08.
    Date of issuance: May 31, 1994
    Effective date: May 31, 1994
    Amendment Nos.: 171 and 202
    Facility Operating License Nos. DPR-71 and DPR-62. Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1994 (59 FR 
21785) The May 16, 1994, letter provided clarifying information that 
did not change the initial no significant hazards consideration 
determination.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 31, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendments: August 20, 1993, as 
supplemented by letters dated December 27, 1993, March 22, 1994, and 
May 31, 1994.
    Brief description of amendments: The amendments delete Technical 
Specification Section 3/4.6.1.5, ``Primary Containment Structural 
Integrity'' which includes Surveillance Requirements for the Primary 
Containment Tendons and adds a Technical Specification requirement to 
establish, implement, and maintain a comprehensive containment tendon 
program. The containment tendon program is based on Regulatory Guide 
1.35, Rev. 3, and is titled ``Inservice Inspection Program for Post 
Tensioning Tendons.'' The new program will allow the Unit 1 and 2 
containments to be tested as twin containments.
    Date of issuance: June 3, 1994
    Effective date: June 3, 1994
    Amendment Nos.: 100 and 84
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59746) The supplemental information submitted December 27, 1993, 
March 22, 1994, and May 31, 1994, contained clarifying information 
related to the original request, and did not change the no significant 
hazards finding. The Commission's related evaluation of the amendments 
is contained in a Safety Evaluation dated June 3, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: December 20, 1993
    Brief description of amendments: The amendments increase the 
minimum critical power ratio (MCPR) from 1.06 to 1.07 for Quad Cities, 
Units 1 and 2, as a result of the planned implementation of GE 8x8NB-3 
fuel for Cycle 14 of each unit.
    Date of issuance: June 10, 1994
    Effective date: June 10, 1994
    Amendment Nos.: 146 and 142
    Facility Operating License Nos. DPR-29 and DPR-30. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 17, 1994 (59 
FR 10003) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 10, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of application for amendment: February 25, 1994
    Brief description of amendment: The amendment adds a new Technical 
Specification 3/4.7.12, ``Ultimate Heat Sink'' and its associated Bases 
Section 3/4.7.12.
    Date of Issuance: May 31, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 172
    Facility Operating License No. DPR-61. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17596) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated May 31, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, Connecticut 06457.

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: December 6, 1993
    Brief description of amendment: The amendment revises the Technical 
Specifications (TSs) to provide several temporary one-time changes that 
are necessary to support the fuel out, chemical decontamination program 
that is currently scheduled for the upcoming 1995 refueling outage. 
Specifically, the amendment revises the definition of the cold shutdown 
condition in TS 1.2.1 by changing the upper limit of Tavg for the 
cold shutdown condition from 200 deg.F to 250 deg.F. The amendment also 
revises the definition of the hot shutdown condition in TS 1.2.2 by 
changing the lower limit of Tavg for the hot shutdown condition 
from greater than 200 deg.F to greater than 250 deg.F.
    Date of issuance: June 9, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 170
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7687) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 9, 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, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: January 27, 1994
    Brief description of amendments: The amendments would eliminate the 
humidity control functions of the containment purge (VP) system 
humidistats by deleting the surveillance requirement (SR) for periodic 
verification of automatic isolation of the VP system on a high relative 
humidity (RH) test signal and heater failure from the existing SR for 
Catawba Units 1 and 2.
    Date of issuance: May 25, 1994
    Effective date: May 25, 1994
    Amendment Nos.: 118 and 112
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10005) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 25, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendments: April 29, 1993, as supplemented 
May 16, 1994
    Brief description of amendments: The amendments delete License 
Condition 2.C.(20) from Facility Operating License NPF-35 for Unit 1, 
and License Condition 2.C.(11) from Facility Operating License NPF-52 
for Unit 2. These conditions address engine teardown and inspection 
required following the crankshaft failure of an Enterprise emergency 
diesel generator at the Shoreham Nuclear Plant.
    Date of issuance: June 2, 1994
    Effective date: June 2, 1994
    Amendment Nos.: 119/113
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 26, 1993 (58 FR 
30192) The May 16, 1994, letter provided additional information that 
did not change the scope of the April 29, 1993, application and 
proposed initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 2, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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: January 13, 1993, as 
supplemented January 28, February 17, and April 26, 1993.
    Brief description of amendments: The amendments revise Technical 
Specification Table 2.2.1, Sections 3/4.1.2.5, 3/4.1.2.6, 3/4.5.1.1, 3/
4.5.5, and their associated Bases, and Technical Specification 6.9.1.9, 
to relocate the values of certain cycle-dependent limits from the 
Technical Specifications to the Core Operating Limits Report.
    Date of issuance: May 31, 1994
    Effective date: May 31, 1994
    Amendment Nos.: 143 and 125
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41503) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 31, 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.

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

    Date of application of amendments: March 23, 1994, as supplemented 
April 14, May 11, and May 17 (two letters) 1994.
    Brief description of amendments: The amendments relating to the 
March 23, 1994, application revise Technical Specification (TS) 6.9.2, 
``Core Operating Limits Report,'' (COLR) to include a reference to a 
Duke Power Company Topical Report describing an analytical method for 
determining the core operating limits. Specifically, the amendments 
add: ``(4) DPC-NE-1004A, Nuclear Design Methodology Using CASMO-3/
SIMULATE-3P,'' to TS 6.9.2.
    The May 11, 1994, letter added a statement to TS 6.9.2 that the 
approved methods used to determine the core operating limits given in 
TS 6.9.1 are specified in the COLR. The May 11 and 17, 1994, letters 
provided information regarding Duke Power's transition from the EPRI-
NODE-P based methodology to the simulate methodology. Revision 1 to the 
COLR for Oconee 1 Cycle 16 was submitted by letter dated May 17, 1994.
    The April 14, 1994, letter revised the TS Table of Contents to 
delete reference to Table 4.4-1. This table was removed from the TS by 
an amendment issued on September 16, 1993.
    Date of Issuance: June 8, 1994
    Effective date: June 8, 1994
    Amendment Nos.:  206, 206, and 203
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22007) The April 14, May 11, and May 17 (two letters), 1994, letters 
provided additional information that did not change the scope of the 
March 23, 1994, application and the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
June 8, 1994. No significant hazards consideration comments received: 
No
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

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

    Date of amendment request: December 23, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications in accordance with Generic Letter 93-05, ``Line Item 
Technical Specification Improvements To Reduce Surveillance 
Requirements For Testing During Power Operation'' for radiation 
monitors, pressurizer heaters, reactor coolant isolation valves, and 
auxiliary feedwater pumps.
    Date of issuance: June 6, 1994
    Effective date: June 6, 1994
    Amendment No.:  96
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7689) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 6, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of application for amendments: December 30, 1993
    Brief description of amendments: The proposed change would allow a 
one time extension of the allowable outage time for each residual heat 
removal (RHR) pump from 3 to 7 days to allow modifications to the RHR 
system while the plant is in Mode 1.
    Date of issuance: May 31, 1994
    Effective date: May 31, 1994
    Amendment Nos.: 72 and 51
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10007) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 31, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of application for amendments: November 19, 1993
    Brief description of amendments: The amendments modify Technical 
Specification Table 3.3-2, Engineered Safety Features Actuation System 
Instrumentation, modifying the Mode for which Item 6.e, ``Trip of All 
Main Feedwater Pumps, Start Motor-Driven Pumps,'' is required to be 
operable.
    Date of issuance: June 1, 1994
    Effective date: June 1, 1994
    Amendment Nos.: 73 and 52
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67847) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June, 1, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of application for amendments: March 1, 1994
    Brief description of amendments: The amendments modify Technical 
Specification (TS) 3.2.4, ``Quadrant Power Tilt Ratio,'' by adding an 
exception to the requirements of TS 3.0.4.
    Date of issuance: June 1, 1994
    Effective date: June 1, 1994
    Amendment Nos.: 74 and 53
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17599) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 1, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

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

    Date of application for amendment: February 7, 1994
    Brief description of amendment: The amendment revises the plant 
Technical Specifications (TS) to require the Three Mile Island, Unit 1 
(TMI-1) annual radioactive effluent release report for the previous 
calendar year be submitted by May 1 of each year. The current TS 
requires the TMI-1 report be submitted within 60 days after January 1 
of each year. Changing the TMI-1 due date to May 1 enables the licensee 
to combine the reports for TMI-1 and TMI-2 into a single report with a 
common due date.
    Date of Issuance: June 10, 1994
    Effective date: As of its date of issuance to be implemented within 
30 days.
    Amendment No.: 185
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17600) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 10, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

    Date of amendment request: May 27, 1993, as supplemented by letter 
dated April 18, 1994.
    Brief description of amendments: The amendments upgrade the fuel 
used in the South Texas Project reactors to Westinghouse VANTAGE 5 
Hybrid (V5H) design and implement several analytical and operational 
upgrades into the South Texas Project Updated Final Safety Analysis 
Report. The amendments modify related setpoints, limiting conditions 
for operation, surveillance requirements, design features information, 
and associated bases in the following specifications: TS Table 2.2-1, 
``Reactor Trip System Instrumentation Trip Setpoints,'' TS Figure 3.1-
1, ``Required Shutdown Margin for Modes 1 and 2,'' TS Figure 3.1-2, 
``Required Shutdown Margin for Mode 5,'' TS Figure 3.1-2a, ``MTC versus 
Power Level,'' TS 3/4.2.5, ``Power Distribution Limits - DNB 
Parameter,'' TS Table 3.3-4, ``Engineered Safety Features Actuation 
System Instrumentation Trip Setpoints,'' TS 3/4.6.1.1, ``Primary 
Containment - Containment Integrity,'' TS 3/4.6.1.2, ``Containment 
Systems - Containment Leakage,'' TS 3/4.6.1.3, ``Containment Systems - 
Containment Air Locks,'' TS 3/4.6.1.5, ``Containment Systems - Air 
Temperature,'' TS 3/4.7.1.2, ``Plant Systems - Auxiliary Feedwater 
System,'' TS 5.2.1, ``Containment - Configuration,'' TS 5.3.1, 
``Reactor Core - Fuel Assemblies,'' TS 5.6.1, ``Fuel Storage - 
Criticality,'' and adds TS Figure 5.6-7, ``Minimum IFBA Content for In-
Containment Rack Fuel Storage.''
    Date of issuance: May 27, 1994
    Effective date: May 27, 1994, to be implemented prior to completion 
of Unit 1 REO5
    Amendment Nos.:  Unit 1 - Amendment No. 61; Unit 2 - Amendment No. 
50
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36436) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 27, 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

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

    Date of application for amendment: March 9, 1994, as supplemented 
April 13, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications to allow a one-time extension for Type B and C leak rate 
tests. The Commission had previously granted a one-time schedular 
exemption from the requirements in 10 CFR Part 50, Appendix J, 
paragraphs III.D.2.(a) and III.D.3. The exemption extends the maximum 
allowable time between tests by 150 days.
    Date of issuance: June 1, 1994
    Effective date: June 1, 1994
    Amendment No.: 162
    Facility Operating License No. DPR-74. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22009). The April 13, 1994, supplemental letter provided clarifying 
information that was within the scope of the April 28, 1994, notice. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated June 1, 1994.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns 
Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama

    Date of application for amendment: January 14, 1992 (TS 300)
    Brief description of amendments: The amendments add requirements to 
the Browns Ferry Units 1 and 3 Technical Specifications to ensure 
thermal-hydraulic stability, consistent with guidance provided by NRC 
Bulletin 88-07 ``Power Oscillations in Boiling Water Reactors,'' and 
Supplement 1 to that Bulletin.
    Date of issuance: May 31, 1994
    Effective date: May 31, 1994
    Amendment Nos.: 206 and 179
    Facility Operating License Nos. DPR-33 and DPR-68:
    Date of initial notice in Federal Register: April 15, 1992 (57 FR 
13138) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 31, 1994.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611

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: February 8, 1994 (TS 93-14)
    Brief description of amendments: The amendments increase the 
pressure setpoint for the motor driven auxiliary feedwater pumps 
switchover from the condensate storage tank to the essential raw 
cooling water supply.
    Date of issuance: May 27,1994
    Effective date: May 27, 1994
    Amendment Nos.: 183 and 175
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12368) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated May 27, 1994.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendment: July 14, 1993
    Brief description of amendment: This amendment revises Sections 3.6 
and 4.6 of the Technical Specifications to incorporate reactor coolant 
system leakage detection requirements to address Generic Letter 88-01 
``NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in 
BWR Austenitic Stainless Steel Piping.''
    Date of issuance: June 1, 1994
    Effective date: June 1, 1994
    Amendment No.: 139
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12370) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 1, 1994 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301.

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555, 
and at the local public document room for the particular facility 
involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By July 22, 1994, 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 20555 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket No. STN 50-456, Braidwood 
Station, Unit No. 1, Will County, Illinois

    Date of application for amendments: April 25, 1994, as supplemented 
April 28, 1994, April 30, 1994, May 2, 1994, May 4, 1994, and May 6, 
1994.
    Brief description of amendments: The amendment revises Braidwood, 
Unit 1, technical specifications (TSs) in Appendix A to the operating 
license by adding additional surveillance and operating requirements to 
Section 4.4.5.2, ``Steam Generator Tube Sample Selection and 
Inspection; Section 4.4.5.4, ``Acceptance 
Criteria; Section 4.4.5.5, ``Reports; and Section 
3.4.6.2. This amendment is applicable only for 100 calendar days from 
the date of issuance, not counting any time when the Thot 
temperature is below 500 deg.F. These changes revise the existing steam 
generator tube repair criteria to allow usage of the voltage-based 
criteria identified by the staff in draft NUREG-1477 as the interim 
plugging criteria (IPC). Additionally, a footnote is added to TS 3.4.8 
to limit the dose equivalent iodine-131 concentration to 0.35 
microcuries per gram of coolant for the limited time period cited 
above. The Unit 1 Bases are revised to be consistent with the changes 
cited above.
    Date of issuance: May 7, 1994
    Effective date: May 7, 1994
    Amendment No.: 50
    Facility Operating License No. NPF-72. The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes. The NRC published a public 
notice of the proposed amendment, issued a proposed finding of no 
significant hazards consideration and requested that any comments on 
the proposed no significant hazards consideration be provided to the 
staff by the close of business on May 5, 1994. The notice was published 
in the Herald News and the Morris Daily Herald on May 3, 1994. The 
Commission's related evaluation of the amendment, finding of emergency 
circumstances, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated May 7, 1994.
    Attorney for the licensee: Michael I. Miller, Esquire, Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
    NRC Project Director: James E. Dyer

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

    Date of application for amendment: May 19, 1994
    Brief description of amendment: The amendment revises the 
surveillance requirements in TS 3.3.9.3 and 3.3.10.3, to change the 
neutron power limits i.e., 105 neutron counts per second (cps) and 
1E-6 amperes (amps) indications on the source and intermediate range 
instruments, respectively, for verifying overlap between them.
    Date of issuance: May 27, 1994
    Effective date: May 27, 1994
    Amendment No.: 150
    Facility Operating License No. DPR-72. Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendment and the final determination of no 
significant hazards consideration comments are contained in a Safety 
Evaluation dated May 27, 1994.
    Attorney for the Licensee: Harold F. Reis, Esquire, Newman and 
Holtzer, P.C., 1615 L Street, NW., Washington DC 20036
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629
    NRC Project Director: Herbert N. Berkow

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

    Date of application for amendment: May 27, 1994, as supplemented 
June 1, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) by adding a footnote to Tables 3.3-3, 3.3-4 and 
3.3-5 of the Millstone Unit No. 2 TS denoting that the operability of 
the automatic initiation logic for the auxiliary feedwater system will 
rely on operator action for the remainder of Cycle 12.
    Date of issuance: June 7, 1994
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 176
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated June 7, 1994.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
    NRC Project Director: John F. Stolz
    Dated at Rockville, Maryland, this 15th day of June 1994.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/IIOffice of Nuclear Reactor 
Regulation
[Doc. 94-15025 Filed 6-21-94 8:45 am]
BILLING CODE 7590-01F