[Federal Register Volume 62, Number 107 (Wednesday, June 4, 1997)]
[Notices]
[Pages 30629-30652]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10604]


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NUCLEAR REGULATORY COMMISSION

Biweekly Notice


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from May 12, 1997, through May 22, 1997. The last 
biweekly notice was published on May 21, 1997 (62 FR 27792).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By July 7, 1997, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first

[[Page 30630]]

prehearing conference scheduled in the proceeding, but such an amended 
petition must satisfy the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(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 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of amendment request: April 22, 1997 (supersedes October 15, 
1996, request)
    Description of amendment request: The proposed amendment would 
revise the Big Rock Point Technical Specifications to correct several 
administrative and editorial inconsistencies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are clarifications within the Technical 
Specifications, and do not alter the technical content of the 
technical specifications. Plant operation or configuration is not 
affected. The postulated doses received by the general public and 
plant personnel as a direct result of accidents previously 
described, are not affected. Plant operation or configuration is not 
affected. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes are either clarifications to correct 
inconsistencies within the Technical Specifications, or corrections 
of typographical errors. The proposed changes do not alter the 
facility in any way, therefore the proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change[s] [do] not affect any margin of safety as 
defined by the Plant Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: John N. Hannon

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of amendment request: April 30, 1997
    Description of amendment request: The proposed amendment would 
alter the company name in the Facility Operating License DPR-6 and 
Technical Specifications for the Big Rock Point Plant. Specifically, 
the proposed amendment would revise the company name from ``Consumers 
Power Company'' to ``Consumers Energy Company.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

[[Page 30631]]

    The proposed changes alter the company name in the Facility 
Operating License and Technical Specifications to reflect the change 
from ``Consumers Power Company'' to ``Consumers Energy Company''. 
The company will continue to own all of the same assets, will 
continue to serve the same customers, and will continue to honor all 
existing obligations and commitments.
    Since the proposed changes do not alter the technical content of 
any Facility Operating License or Technical Specifications 
requirements, they do not alter the design, function, or operation 
of any plant structure, system, or component.
    Therefore, the changes will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes alter the company name in the Facility 
Operating License and Technical Specifications to reflect the change 
from ``Consumers Power Company'' to ``Consumers Energy Company''. 
The company will continue to own all of the same assets, will 
continue to serve the same customers, and will continue to honor all 
existing obligations and commitments.
    Since the proposed changes do not alter the technical content of 
any Facility Operating License or Technical Specifications 
requirements, they do not alter the design, function, or operation 
of any plant structure, system, or component.
    Therefore, the changes will not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Since the proposed changes do not alter the technical content of 
any Facility Operating License or Technical Specifications 
requirements, they do not alter the design, function, or operation 
of any plant structure, system, or component.
    Therefore, the changes will 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: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: John N. Hannon

Duke Power Company, Docket No. 50-413, Catawba Nuclear Station, 
Unit 1, York County, South Carolina

    Date of amendment request: May 8, 1997
    Description of amendment request: The proposed amendment would add 
a phrase to the footnote to Section 3.4.1.2 of the Technical 
Specifications that would permit all reactor coolant pumps (RCPs) to be 
deenergized for up to 4 hours during Mode 3 on a one-time basis. 
Currently, the RCPs are permitted to be deenergized for up to 1 hour 
during Mode 3. The proposed change would allow the licensee to perform 
a natural circulation test using the new steam generators (installed in 
late 1996).
    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 activity does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed natural circulation test would be performed in Mode 
3 with the reactor subcritical. This transient is bounded by the 
transient analyzed in UFSAR [Updated Final Safety Analysis Report] 
Section 15.2.6, Loss of Non-Emergency AC Power to the Station 
Auxiliaries. For this ANS [American Nuclear Society] Condition II 
event, the reactor is assumed to be operating at 102% power, the 
turbine driven auxiliary feedwater pump is assumed unavailable and 
each steam generator is assumed to have 18% of the steam generator 
tubes plugged. By contrast, the planned natural circulation test 
would be performed with the reactor subcritical, less than 0.1% of 
the tubes plugged in each steam generator, and all support systems 
such as auxiliary feedwater, operable for the test. Therefore, the 
proposed natural circulation test would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2) The activity does not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    The proposed change does not involve a physical alteration of 
the unit (i.e., no new or different equipment will be installed), 
nor will the function of equipment be changed. The change will allow 
for a one time performance of a natural circulation test in Mode 3 
which will provide useful data on the natural circulation 
capabilities of the new Babcock and Wilcox International (BWI) steam 
generators that were recently installed at Catawba Unit 1. The test 
data will be utilized to validate analysis and simulator models. 
Plant operators will also receive valuable experience from 
performance of the test. The test will be conducted using written 
and approved procedures. An Emergency procedure (EP/1/A/5000/ECA-
0.1) is also available to the Operators for this test. This test is 
bounded by the Loss of Non-Emergency AC Power to the Station 
Auxiliaries event in Section 15.2.6 of the Catawba UFSAR. For these 
reasons, the planned natural circulation test will not create the 
possibility of a new or different type of accident from any 
previously evaluated.
    3) The activity does not involve a significant reduction in the 
margin of safety.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (the fuel and fuel cladding, the 
Reactor Coolant System pressure boundary, and the containment) to 
limit the level of radiation doses to the public. As demonstrated by 
the bounding UFSAR analysis in Section 15.2.6, none of the fission 
product barriers are adversely impacted by the proposed one-time 
change. The proposed change does not alter the manner in which 
safety limits, limiting safety system setpoints, or limiting 
conditions for operation are determined. For these reasons, the 
activity does not involve a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
proposed amendments involve no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: February 24, 1997, as supplemented on 
April 24, 1997.
    Description of amendment request: The licensee proposed changes to 
Technical Specification Section 6.9.1.7, Core Operating Limits Report, 
to reflect use of the Westinghouse Best Estimate Large Break Loss-of-
Coolant Accident (LOCA) methodology for large break LOCA analysis, 
including supporting documents.
    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:
    Question 1 Does the proposed license amendment involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    The plant conditions assumed in the analysis are bounded by the 
design conditions for all equipment in the plant. Therefore, there 
will be no increase in the probability of a Loss of Coolant Accident

[[Page 30632]]

(LOCA). The consequences of a LOCA are not being increased. That is, 
it is shown that the emergency core cooling system is designed so 
that its calculated cooling performance conforms to the criteria 
contained in 10 CFR 50.46 paragraph (b). No other accident is 
potentially affected by this change. Therefore, neither the BiWeekly 
probability nor the consequences of an accident previously evaluated 
is increased due to the proposed change.
    Question 2 Does the proposed license amendment create the 
possibility of a new or different kind of accident from any accident 
previously evaluated?
    No new modes of plant operation are being introduced. The 
parameters assumed in the analysis are within the design limits of 
existing plant equipment. All plant systems will perform as designed 
in response to a potential accident. Therefore, the proposed license 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Question 3 Does the proposed amendment involve a significant 
reduction in the margin of safety?
    The analysis in support of the proposed license amendment 
realistically models the expected response of the Turkey Point Units 
3 & 4 nuclear core during a postulated LOCA. Uncertainties have been 
accounted for as required by 10 CFR 50.46. A sufficient number of 
loss of coolant accidents with different break sizes, different 
break locations and other variations in properties have been 
calculated to provide assurance that the most severe postulated loss 
of coolant accidents were analyzed. It has been shown by the 
analysis that there is a high level of probability the criteria 
contained in 10 CFR 50.46 paragraph (b) would not be exceeded. 
Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Florida International 
University, University Park, Miami, Florida 33199
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: February 17, 1997 as revised May 1, 
1997.
    Description of amendment request: The proposed amendment would 
change the Crystal River Unit 3 (CR-3) Technical Specifications (TS) to 
implement 10 CFR Part 50, Appendix J, ``Primary Reactor Containment 
Leakage Testing for Water-Cooled Reactors,'' Option B. This option 
allows to change from prescriptive testing requirements to performance-
based testing requirements based on the leakage rate testing history of 
the containment and components. The proposed TS changes include 
revision to TS 3.6.1, 3.6.3, and addition of ``Containment Leakage Rate 
Testing Program'' to TS 5.0. The licensee did not propose any 
deviations from methods approved by the Commission and endorsed in the 
applicable regulatory guide. This notice supersedes the previous notice 
dated February 28, 1997 (62 FR 9214)
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The TS amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to the TS are to implement Option B of 10 
CFR 50, Appendix J, at CR-3. The proposed changes will result in 
increased intervals between containment leakage tests based on the 
leakage rate testing history. The proposed changes do not involve a 
change to the plant design or operation and does not change the 
testing methodology.
    NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' 
provides the technical basis of 10 CFR 50, Appendix J, Option B. 
NUREG-1493 contains a detailed evaluation of the expected leakage 
from containment and the associated consequences. The increased risk 
due to increasing the intervals between containment leakage tests 
was also evaluated. The NUREG used a statistical approach to 
determine that the increase in the expected dose to the public due 
to decreasing the testing frequency is extremely low. NUREG-1493 
also concluded that a small increase is justifiable in comparison to 
the benefits from decreasing the testing frequency. The primary 
benefit is in the reduction in occupational radiation exposure.
    Criterion 2
    Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The TS amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed TS amendment incorporates the performance-based 
testing approach authorized by 10 CFR 50 Appendix, J, Option B. 
Decreasing the testing frequency allowed by this change does not 
involve a change to plant design or operation. Safety related 
equipment and safety functions are not altered as a result of this 
change. Decreasing the testing frequency does not affect testing 
methodology. As a result, the proposed change does not affect any of 
the parameters or conditions that could contribute to the initiation 
of any accidents.
    Criterion 3
    Does not involve a significant reduction in the margin of 
safety.
    This TS amendment does not involve a significant reduction in 
the margin of safety.
    The proposed TS amendment does not change the methodology of the 
containment leakage rate testing program or program acceptance 
criteria. The proposed TS change does affect the frequency of 
containment leakage rate testing. With an increased interval between 
tests, a small possibility exists that an increase in leakage could 
go undetected for a longer period of time. Based on the operational 
experience at CR-3, it has been demonstrated that the leak-tightness 
of the containment building has consistently been significantly 
below the allowable leakage limit. Adequate controls are in place to 
ensure that required maintenance and modifications are performed.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629
    Attorney for licensee: R. Alexander Glenn, Corporate Counsel, 
Florida Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, 
Florida 33733-4042
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: March 27, 1997, as supplemented April 3, 
and May 1, 1997.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) for the Crystal River Nuclear 
Plant Unit 3 (CR3) relating to the Once Through Steam Generator's 
(OTSG's) tube inspection acceptance criteria. Specifically, the 
licensee proposed to:
    (1) revise TS 3.4.12 (d) to specify 150 gallons per day limit on 
primary-to-secondary leakage through either OTSG;
    (2) add TS 5.6.2.10.2 e. to define inspection requirements and 
disposition criteria for applicable tubes in the ``B'' OTSG first span;

[[Page 30633]]

    (3) revise TS 5.6.2.10.4.a.7 to define ``pit-like Intergranular 
attack indications
    (4) revise TS 5.6.2.10 and 5.7.2 to delete requirements that were 
specific to the interim tube plugging criteria applicable until Refuel 
11.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    FPC Response:
    No. The CR-3 components addressed by this proposed change are 
the Once Through Steam Generators (OTSGs), identified by plant 
tagging procedures as RCSG-1A and RCSG-1B. The OTSGs are straight 
tube, straight shell heat exchangers which allow for heat removal 
and the subsequent production of steam as a result of heat transfer 
from the primary side reactor coolant to the secondary side 
feedwater. Proposed changes are; retaining reduced primary-to-
secondary leak rates approved previously for one cycle only, 
returning inspection result reporting requirements to those 
previously implemented, and establishing new inspection requirements 
for the ``B'' OTSG. Sunset clauses are being removed from pages 
containing requirements effective for one refueling outage and 
subsequent operating cycle only.
    Based on review of Chapter 14 of the CR-3 Final Safety Analysis 
Report (FSAR), FPC performed analyses to assess the consequences of 
a steam generator tube rupture event, including the complete 
severing of a steam generator tube. The analyses concluded that CR-3 
was sufficiently designed to ensure that, in the event of a steam 
generator tube rupture, the radiological doses would not exceed the 
allowable limits prescribed by 10 CFR 100, and would not result in 
additional tube failures and further degradation of the reactor 
coolant pressure boundary.
    Retaining the present primary-to-secondary leakage limit (LCO 
3.4.12, RCS Operational Leakage) that was previously approved for 
the current operating cycle will continue to provide assurance that 
should a significant leak occur, it would be detected and the plant 
will be shut down in a timely manner to reduce the likelihood of a 
potential tube rupture. This value of primary-to-secondary leakage 
applicable to both OTSGs is conservative relative to existing safety 
analyses and would result in lower doses than currently calculated 
and found acceptable. Removing reporting requirements specific to 
use of alternate flaw sizing criteria approved for Refueling Outage 
10 only, and returning to previous reporting requirements applicable 
to both OTSGs, has no effect on operating plant safety. These 
requirements are administrative only and do not affect steam 
generator inspection or disposition of inspection results.
    The proposed change to the ``B'' OTSG inspection criteria 
establishes that future inspections will include 100% inspection of 
the first span of specific tubes which are known to have indications 
of degradation. The degradation of these tubes is attributed to a 
common non-random mechanism.
    The results of inspections of these tubes will be dispositioned 
using the same criteria as all other OTSG tubes for determination of 
the need for plugging or sleeving. Therefore, the proposed change 
will not increase the probability or consequence of an accident 
previously evaluated as all tubes degraded beyond acceptable limits 
will be subject to consistent corrective actions.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    FPC Response:
    No. The purpose of OTSG tube inspection is to identify tubes 
that may have a higher potential for failure due to degradation that 
results in a reduced ability to withstand operating conditions. 
Neither the type of inspection of OTSG tubes nor the process for 
performing inspections will be changed by this amendment. Consistent 
criteria will be applied to disposition inspection results and 
consistent corrective actions will be taken for tubes that exceed 
this criteria. Retaining the lower leakage limit is conservative 
relative to existing analyses. Changes to revise requirements for 
reporting inspection results, and remove ``sunset'' clauses 
addressing the applicability of License Amendment 154 until 
Refueling Outage 11 only, do not alter the design or operation of 
the OTSGs. Therefore, no new or different kind of accident will be 
created as a result of these changes.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in margin of safety?
    FPC Response:
    No. The analyses that have been performed on the effects of OTSG 
tube failures, as reported in the CR-3 FSAR, have demonstrated that 
internal and offsite consequences are within allowable limits. This 
change will not alter the acceptance criteria for inspection 
results. Since this change will assure that a group of tubes with 
existing first span pit-like inter-granular attack indications are 
inspected each inspection period, the likelihood of detecting active 
degradation, as well as the probability of repairing degraded tubes 
prior to the degradation resulting in a through-wall opening or tube 
rupture, is increased. Retaining the currently accepted primary-to-
secondary leakage limit continues to provide assurance that should a 
significant leak occur, it would be detected and the plant will be 
shut down in a timely manner to reduce the likelihood of a potential 
tube rupture, thereby maintaining or improving the existing margin 
of safety. Changes to revise requirements for reporting inspection 
results, and remove ``sunset'' clauses addressing the applicability 
of License Amendment 154 until Refueling Outage 11 only, do not 
alter the design or operation of the OTSGs. Therefore, these changes 
will not involve a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629
    Attorney for licensee: R. Alexander Glenn, Corporate Counsel, 
Florida Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, 
Florida 33733-4042
    NRC Project Director: Frederick J. Hebdon


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

    Date of amendment request: May 8, 1997
    Description of amendment request: The proposed amendment 
incorporates additional analytical methods, GPU Nuclear Topical 
Reports, TR-078, TR-087, TR-091, and TR-092P, previously approved by 
the NRC, to Technical Specifications (TS) Section 6.9.5.2. These 
Topical Reports will be utilized by GPU Nuclear to perform core reload 
design analysis for the Three Mile Island, Unit 1 (TMI-1) Facility. TS 
6.9.5.2 is also being editorially revised to relocate the existing note 
that the current revision level shall be specified in the Core 
Operating Limits Report (COLR) such that it applies to the additional 
Topical Reports, as well as BAW-10179 P-A.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:
    GPU Nuclear has determined that this Technical Specification 
Change Request poses no significant hazards as defined by 10 CFR 
50.92.
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The proposed change to reference the 
analytical methodologies specified in GPU Nuclear Topical Reports 
TR-078, TR-087, TR-091,and TR-092 use[d] in TMI-1 core reload design 
analysis is considered administrative since these Topical Reports

[[Page 30634]]

have been reviewed and approved by the NRC for use at TMI-1.
    Therefore, the proposed change does not involve a significant 
increase in the probability of occurrence or the consequences of an 
accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. The proposed change 
to reference NRC-approved GPU Nuclear Topical Reports TR-078, TR-
087, TR-091, and TR-092P will continue to ensure that approved 
methods and criteria are used to establish core operating limits.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed change to reference NRC-approved GPU Nuclear 
Topical Reports TR-078, TR-087, TR-091, and TR-092P maintains 
existing margins of safety since approved methods and criteria are 
still used to establish core operating limits.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Patrick D. Milano, Acting

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 8, 1997
    Description of amendment request: The proposed amendment would 
modify the minimum accuracy stated in Technical Specification (TS) 
Table 3.3-8, Meteorological Monitoring Instrumentation,'' 
for the instruments used to measure wind speed and air temperature - 
delta T. TS Bases Section 3/4.3.3.4 would also be modified to reflect 
the proposed changes to TS Table 3.3-8.
    Regulatory Guide 1.23 (Safety Guide 23), ``Onsite Meteorological 
Programs,'' dated March 17, 1972, provides recommended instrument 
accuracies for meteorological instrumentation. The proposed minimum 
instrument accuracies for the air temperature - delta T and the wind 
speed (only when the wind speed is greater than 5 miles per hour) do 
not meet the recommended accuracies of Regulatory Guide 1.23. However, 
margin is included to account for uncertainties.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes modify the accuracy requirements for the 
instruments which are used to measure wind speed and air temperature 
- delta T. The data obtained from the meteorological instrumentation 
would be used to: a) estimate the public dose following routine or 
accidental releases of airborne radioactivity, b) make decisions 
regarding actions to protect the public in the event of an accident 
involving a release of airborne radioactivity, and c) establish 
radiological dispersion parameters to determine radiological doses 
in design basis accident calculations.
    The proposed minimum instrument accuracy requirements are more 
than sufficient to meet the purposes denoted above. The 
meteorological parameters measurement uncertainties insignificantly 
affect the results when compared to the accuracies of the source 
term estimates, meteorological dispersion models, dose models, and 
meteorological forecasting. Therefore, there is no impact on the 
consequences (offsite doses) associated with previously evaluated 
accidents.
    The proposed changes do not alter the way any structure, system, 
or component functions, do not alter the manner in which the plant 
is operated, and do not have any impact on the protective boundaries 
and safety limits for the protective boundaries. Therefore, the 
proposed changes do not impact the probability of any previously 
evaluated accidents.
    Thus, the license amendment request does not impact the 
probability of an accident previously evaluated nor does it involve 
a significant increase in the consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes modify the accuracy requirements for the 
instruments which are used to measure wind speed and air temperature 
- delta T. The data provided by these instruments assist in 
responding to a design basis accident which may involve a release of 
airborne radioactivity. The instruments are used for post accident 
monitoring and serve a passive role; they cannot initiate or 
mitigate any accident.
    The proposed changes do not alter the way any structure, system, 
or component functions and do not alter the manner in which the 
plant is operated. They do not introduce any new failure modes.
    Thus, the license amendment request 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.
    As discussed above, the proposed changes modify the accuracy 
requirements for the instruments which are used to measure wind 
speed and air temperature - delta T which could impact the 
radiological dispersion coefficient used to determine radiological 
doses in design basis accident calculations. However, the 
differences in the instrument accuracies and the Regulatory Guide 
1.23 requirements have been determined not to significantly affect 
the dispersion coefficients. Thus, there is no significant impact on 
offsite doses associated with previously analyzed accidents. 
Therefore, there is no significant reduction in the margin of safety 
for the design basis accident analysis.
    Thus, the license amendment request does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: April 14, 1997
    Description of amendment request: Technical Specification 3.4.9.3.a 
requires two relief valves be operable to protect the reactor coolant 
system from overpressurization when any reactor coolant system cold leg 
is less than 350F. The proposed amendment revises the setpoint of the 
residual heat removal suction relief valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 30635]]

consideration, which is presented below:
    NNECO has reviewed the proposed change in accordance with 10CFR 
50.92 and has concluded that the change does not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR 50.92(c) are not 
satisfied. The proposed change does not involve a SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to Technical Specification 3.4.9.3 to 
decrease the setpoint of the residual heat removal suction relief 
valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus] 
3% ([greater than or equal to] 426.8 psig and [less than or equal 
to] 453.2 psig) is consistent with the design capabilities and 
system requirements of the relief valves and the relief valves are 
not credited in previously evaluated accidents.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change to Technical Specification 3.4.9.3 to 
decrease the setpoint of the residual heat removal suction relief 
valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus] 
3% ([greater than or equal to] 426.8 psig and [less than or equal 
to] 453.2 psig) does not change the operation of the residual heat 
removal system, reactor coolant system or any system component 
during normal or accident evaluations. The proposed change to the 
setpoint of the residual heat removal suction relief valves from 450 
psig [plus or minus] 3% to 440 psig [plus or minus] 3% ([greater 
than or equal to] 426.8 psig and [less than or equal to] 453.2 psig) 
also ensures protection of the reactor coolant system against cold 
overpressurization transients in accordance with the requirements of 
10CFR50, Appendix G.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to Technical Specification 3.4.9.3 to 
decrease the setpoint of the residual heat removal suction relief 
valves from 450 psig [plus or minus] 3% to 440 psig [plus or minus] 
3% ([greater than or equal to] 426.8 psig and [less than or equal 
to] 453.2 psig) provides an acceptable allowance between the maximum 
relief valve setpoint ([less than or equal to] 453.2 psig) and 
10CFR50, Appendix G requirements. The proposed change to the 
setpoint provides sufficient allowance between the minimum relief 
valve setpoint ([greater than or equal to] 426.8 psig) and reactor 
coolant system pressure when residual heat removal system is 
unisolated from the reactor coolant system to minimize the 
probability of an inadvertent residual heat removal system relief 
valve opening.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed change does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: April 28, 1997
    Description of amendment request: Technical Specification 
Surveillances 4.1.2.3.1, 4.1.2.4.1, 4.5.2.f, and 4.5.2.h require the 
charging and safety injection pumps to be tested on a periodic basis 
and after modifications that alter subsystem flow characteristics. The 
proposed amendment would increase the required differential pressure at 
recirculation flow for the safety injection and centrifugal charging 
pumps; decrease the required individual safety injection and 
centrifugal charging pump injection line flow rate; increase the 
allowed individual safety injection pump total flow rate; and make 
editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed changes in accordance with 
10CFR50.92 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 
satisfied. The proposed changes do not involve [an] SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes to Technical Specification Surveillances 
4.1.2.3.1, 4.1.2.4.1, and 4.5.2.f to increase the required discharge 
pressure for the centrifugal charging pumps on recirculation flow 
during surveillance testing from [greater than or equal to] 2411 
psid to [greater than or equal to] 5676 ft (2464 psid) are 
consistent with centrifugal charging pump design requirements. The 
change in the referenced units from differential pressure measured 
in psid to total head measured in feet for the centrifugal charging 
pumps and safety injection pumps during surveillance testing is an 
administrative change.
    The proposed changes to Technical Specification Surveillance 
4.5.2.f to increase the required discharge pressure for the safety 
injection pumps on recirculation flow during surveillance testing 
from [greater than or equal to] 1348 psid to [greater than or equal 
to] 3240 ft (1406 psid) are consistent with safety injection pump 
design requirements.
    The proposed changes to Surveillance 4.5.2.h: to decrease the 
required individual centrifugal charging pump injection line flow 
rate sum from [greater than or equal to] 339 gpm to [greater than or 
equal to] 310.5 gpm, decrease the required individual safety 
injection pump injection line flow rate sum from [greater than or 
equal to] 442.5 gpm to [greater than or equal to] 423.4 gpm, 
increase the required individual safety injection Pump A total flow 
rate from [less than or equal to] 670 gpm to [less than or equal to] 
675 gpm, and increase the required individual safety injection Pump 
B total flow rate from [less than or equal to]
    650 gpm to [less than or equal to] 675 gpm are consistent with 
centrifugal charging pump and safety injection pump design 
requirements.
    The proposed changes are consistent with equipment design 
requirements and performing surveillance testing does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed changes to the surveillance testing of the 
centrifugal charging pumps and safety injection pumps provide the 
necessary assurance that the pumps will function consistent with the 
flows used in the accident analyses and does not involve a 
significant increase in the consequence of an accident previously 
evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the surveillance testing of the 
centrifugal charging pumps and safety injection pumps do not change 
the operation of the centrifugal charging or safety injection 
systems or any of its components during normal or accident 
evaluations. The increase in the allowed maximum safety injection 
pump flow does not impact the cold overpressure accident analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.

[[Page 30636]]

    3. Involve a significant reduction in a margin of safety.
    The proposed changes to Technical Specification Surveillances 
4.1.2.3.1, 4.1.2.4.1 and 4.5.2.f to increase the required discharge 
pressure for the centrifugal charging pumps on recirculation flow 
during surveillance testing from [greater than or equal to] 2411 
psid to [greater than or equal to] 5676 ft (2464 psid) provides an 
acceptable margin between the required surveillance and design pump 
performance to provide assurance that the pumps will operate 
consistent with the assumptions of the accident analysis.
    The proposed changes to Technical Specification Surveillance 
4.5.2.f to increase the required discharge pressure for the safety 
injection pumps on recirculation flow during surveillance testing 
from [greater than or equal to] 1348 psid to [greater than or equal 
to] 3240 ft (1406 psid) provides an acceptable margin between the 
required surveillance and design pump performance to provide 
assurance that the safety injection pumps will operate consistent 
with the assumptions of the accident analysis.
    The proposed changes to Surveillance 4.5.2.h to decrease the 
required individual centrifugal charging pump injection line flow 
rate sum from [greater than or equal to] 339 gpm to [greater than or 
equal to] 310.5 gpm, decrease the required individual safety 
injection pump injection line flow rate sum from [greater than or 
equal to] 442.5 gpm to [greater than or equal to] 423.4 gpm, 
increase the required individual safety injection Pump A total flow 
rate from [less than or equal to] 670 gpm to [less than or equal to] 
675 gpm and increase the required individual safety injection Pump B 
total flow rate from [less than or equal to] 650 gpm to [less than 
or equal to] 675 gpm are consistent with the assumptions of the 
accident analysis. The maximum allowed safety injection flow is 
consistent with the vendor recommendation for maximum continuous 
runout flow. Also, the safety injection
    pumps are disabled during specific normal operating modes, 
consistent with the assumptions of the accident analysis, to ensure 
that they can not be an injection source when the cold overpressure 
system is required to be operable and thus the increase in maximum 
safety injection pump flow does not affect the cold overpressure 
accident analysis.
    The change in the referenced units in Technical Specification 
Surveillances 4.1.2.3.1, 4.1.2.4.1 and 4.5.2.f from differential 
pressure measured in psid to total head measured in feet for the 
centrifugal charging pumps and safety injection pumps during 
surveillance testing is an administrative change.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed changes do not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: April 28, 1997
    Description of amendment request: Technical Specification 3.7.6 
requires that flood protection be provided for the service water pump 
cubicles and components when the water level exceeds a specific value. 
The proposed amendment (1) adds the closing of the service water pump 
cubicle sump drain valves, (2) revises the wording of the action 
statement to be consistent with the limiting condition for operation, 
and (3) revises the associated Bases section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed change in accordance with 
10CFR50.92 and has concluded that the change does not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed change does not involve [an] SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed changes to Technical Specification 3.7.6 identify 
additional manual actions to be performed to provide external lood 
protection for the service water pump cubicles in the event of high 
water level (13 ft MSL) [mean sea level]. The cubicle sump drain 
valves which are to be closed are part of a modification which 
installed a drain line from the sump of each cubicle to the intake 
bay in order to provide a passive means of removing internal leakage 
from the cubicle. The cubicle sump drain valves are normally 
maintained in the open position.
    The drain valves meet the intent of RG [Regulatory Guide] 1.59 
for ``hardened protection'' and RG 1.102 for ``incorporated 
barriers'' in a manner similar to that of the cubicle watertight 
doors. RG 1.59 states that hardened protection ``must be passive and 
in place, as it is to be used for flood protection, during normal 
plant operation''. RG 1.102 states that ``the plant should be 
designed and operated to keep doors necessary for flood protection 
closed during normal operation''. The Response to FSAR [Final Safety 
Analysis Report] Question No. 240.9 established the acceptability of 
the practice of maintaining one service water pump cubicle 
watertight door open and the other door closed during normal 
operations.
    The proposed change in the action statement to initiate action 
when water level is exceeding 13 feet MSL rather than at 13 feet MSL 
is a clarification only which provides consistency between the 
limiting condition for operation and the action statements.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to Technical Specification 3.7.6 identify 
additional, simple to perform manual actions to provide external 
flood protection for the service water pump cubicles.
    The proposed change in the action statement to initiate action 
when water level is exceeding 13 feet MSL rather than at 13 feet MSL 
and the proposed changes to the bases are considered clarifications.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to Technical Specification 3.7.6 identify 
additional, simple to perform manual actions to provide external
    flood protection for the service water pump cubicles in the 
event of high water level (13 ft MSL). The plant modification which 
made these additional actions necessary was made to provide for 
improved internal flood protection.
    The proposed change in the action statement to initiate action 
when water level is exceeding 13 feet MSL rather than at 13 feet MSL 
and the proposed changes to the bases are considered clarifications.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed change does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike,

[[Page 30637]]

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: May 1, 1997
    Description of amendment request: Technical Specifications 3/
4.8.2.2 and 3/4.8.3.2 specify which electrical power systems are 
required to be operable in Modes 5 and 6. The proposed amendment would 
clarify the requirements by identifying the specific equipment required 
and their alignments in Modes 5 and 6.
    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 10CFR 
50.92 and has concluded that the change does not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR 50.92(c) are not 
satisfied. The proposed changes do not involve [an] SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed change to Technical Specification 3/4.8.2.2 to 
replace the wording ``As a minimum, one 125 volt battery bank and 
its associated full capacity charger'' to ``As a minimum, one Train
    (A or B) of batteries and their associated full capacity 
chargers'' will increase the required battery banks operable from 
one to two.[]
    This change is being proposed to resolve an inconsistency with 
Technical Specification 3/4.8.3.2 which currently requires two 
battery banks energized in modes 5 and 6.
    The proposed change to...Technical Specifications 3/4.8.2.2 and 
3/4.8.3.2 to identify the specific equipment required and its 
alignment during modes 5 and 6 is being proposed to reduce the 
vagueness in the present Technical Specifications. This proposed 
change will specify the equipment required operable for the 
electrical distribution systems during modes 5 and 6.
    These proposed changes are considered administrative and do not 
alter the manner in which any system or component is operated or 
expected to respond during an accident. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to Technical Specification 3/4.8.2.2 to 
increase the required battery banks operable from one to two and to 
reword Technical Specifications 3/4.8.2.2 and 3/4.8.3.2 to identify 
the specific equipment required operable during modes 5 and 6 do not 
alter the manner in which any system or component is operated or 
expected to respond during normal or accident conditions.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to Technical Specification 3/4.8.2.2 to 
increase the required battery banks operable from one to two is 
being proposed to resolve an inconsistency with Technical 
Specification 3/4.8.3.2 which currently requires two battery banks 
energized in modes 5 and 6. This is considered an administrative 
change.
    The proposed changes to...Technical Specifications 3/4.8.2.2 and 
3/4.8.3.2 are being proposed to reduce the vagueness in the present 
technical specifications by identifying the specific equipment 
required operable during modes 5 and 6. The change will provide a 
greater level of assurance that the electrical distribution systems 
will be correctly aligned and surveilled. This is also considered an 
administrative change.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed changes to not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: May 5, 1997
    Description of amendment request: Technical Specification 
Surveillance 4.8.4.1 requires periodic testing of lower voltage circuit 
breakers for all containment penetration conductor overcurrent 
protective devices. The proposed amendment would modify the 
requirements for determining the operability of lower voltage circuit 
breakers by using the manufacturer's curve of current versus time to 
test delay trip elements, clarify the use of two pole in series 
testing, and expand the Bases description of the testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed revision does not involve [an] SHC because 
the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed change to Technical Specification Surveillance 
4.8.4.1 to modify the requirements for determining the operability 
of lower voltage circuit breakers by using the manufacture's curve 
of current versus time to test long time and short-time delay trip 
elements will not change the requirement that periodic testing be 
performed to determine breaker operability. The circuit breaker 
testing is consistent with the design of the components and 
performing surveillance testing does not involve a significant 
increase in the probability of an accident previously evaluated. The 
proposed change will provide assurance that the breakers will 
perform consistent with accident analyses and does not involve a 
significant increase in the consequence of an accident previously 
evaluated.
    The proposed change to the surveillance to modify the wording 
associated with the use of two pole in series testing to determine 
Molded Case Circuit Breaker (MCCB) operability following the failure 
of [an] MCCB to pass a single pole test was previously approved in 
License Amendment No. 13. The modified wording clarifies the testing 
by specifically stating in the surveillance that the two pole in 
series test determines MCCB operability. This is considered an 
administrative change.
    The proposed change to expand the description of the long-time 
and short-time delay trip elements testing in the Bases Section is 
also considered an administrative change.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.

[[Page 30638]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change to use a curve of current versus time 
instead of the description in Technical Specification Surveillance 
4.8.4.1 of the [] long-time and short-time delay trip element 
testing does not alter the design, operation, or maintenance of the 
lower voltage circuit breakers.
    The proposed change to the surveillance to modify the wording 
associated with the use of two pole in series testing to determine 
MCCB operability and the expanded description of the long-time and 
short-time delay elements testing in the Bases Section are 
considered administrative changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The current wording of Technical Specification Surveillance 
4.8.4.1 requires testing of long-time delay trip elements with a 
current value of exactly 300% of the pickup setting and short-time 
delay trip elements with a current value of exactly 150% of the 
pickup setting. The testing [cannot] be performed at exact values. 
Circuit breaker manufactures develop a curve of current versus time 
for each breaker type that specifies the allowable time to trip for 
a specified current. Using the curve for a given breaker type, the 
operability of a circuit breaker can be verified by inserting a 
given current and verifying that the breaker trips within the 
allowable time delay band width for that current. Testing by the 
industry is typically performed at approximately 300% of the pickup 
setting for long-time delay trip elements and approximately 150% of 
the pickup setting for short-time delay trip elements. The proposed 
change to the surveillance to modify the requirements for 
determining the operability of circuit breakers by using the 
manufacturer's curve of current versus time to test delay trip 
elements will continue to provide assurance that lower voltage 
circuit breakers for all containment penetration conductor 
overcurrent protective devices will operate consistent with the 
assumptions of the accident analysis.
    The proposed change to the surveillance to modify the wording 
associated with the use of two pole in series testing to determine 
MCCB operability and the expanded description of the long-time and 
short-time delay trip elements testing in the Bases Section are 
considered administrative changes.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed changes do not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: May 5, 1997
    Description of amendment request: Technical Specification 
Surveillance 4.5.2.b.1 requires that the emergency core cooling system 
(ECCS) piping be verified full of water at least once per 31 days. The 
proposed amendment would revise the surveillance to exempt the 
operating charging pump(s) and associated piping from the requirement 
to be verified full of water and move the description of the 
verification method from the surveillance to the Bases section.
    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 revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed revision does not involve [an] SHC because 
the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed change to Technical Specification Surveillance 
4.5.2.b.1 to exempt the operating centrifugal charging pump(s) and 
associated piping from the requirement to be vented will not effect 
the requirement the ECCS piping be full of water. An operating 
centrifugal charging pump and the associated piping is self venting 
and cannot develop voids and pockets of entrained gases. This change 
is consistent with the design of the charging system and ensuring 
that ECCS piping is full of water does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    The proposed change Technical Specification Surveillance 
4.5.2.b.1 to move and expand the description of the venting method 
from the surveillance to the Bases Section are considered 
administrative changes.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change to exempt the operating centrifugal charging 
pump(s) and associated piping from the requirement to be 
periodically vented by crediting its self venting capabilities does 
not change the operation of the charging system or any of its 
components during normal or accident evaluations.
    The proposed changes to move and expand the description of the 
venting method from the surveillance to the Bases Section are 
considered administrative changes.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to Technical Specification Surveillance 
4.5.2.b.1 to exempt the operating centrifugal charging pump(s) and 
associated piping from the requirement to be manually vented by 
crediting its self venting capabilities, is consistent with the 
design of the charging system. This proposed change continues to 
ensure that ECCS piping is full of water and thus, does not involve 
a significant reduction in a margin of safety.
    The proposed change to Technical Specification Surveillance 
4.5.2.b.1 to move the description of the venting method from 
thesurveillance to the Bases Section is considered an administrative 
change. Currently the surveillance identifies that ECCS piping is to 
be verified full of water by venting ECCS pump casings and 
accessible discharge piping high points except for the RSS 
[recirculation spray system] pump, RSS heat exchanger and associated 
RSS piping that are not maintained filled with water during plant 
operation. The venting description will be expanded when moved to 
the bases to include an exclusion for the above described operating 
centrifugal charging pump(s) and associated piping and the venting 
method used for nonoperating centrifugal charging pumps. The 
centrifugal charging pumps have top mounted suction and discharge 
nozzles and do not have casing vents. The pump manufacturer has 
indicated that venting the pump suction pipe will assure that the 
pump is full of water. This venting of the nonoperating centrifugal 
charging pumps is accomplished by using a pump suction line test 
connection.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed change does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 30639]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: April 17, 1997
    Description of amendment request: This license amendment request 
revises Technical Specification (TS) 2.12, ``Control Room System,'' to 
delete the Limiting Conditions of Operation (LCO) and associated 
surveillance for the control room temperature and replace it with an 
LCO and surveillance on the control room air conditioning (A/C) system. 
The remainder of TS 2.12 is being rewritten consistent with the 
requirements of the Combustion Engineering Standard TS (NUREG-1432, 
Rev. 1). In reviewing requirements for refueling and shutdown 
operations, additional TS improvement were identified. Therefore, the 
definition section, TS 2.1 ``Reactor Coolant System,'' 2.6 
``Containment System,'' 2.8 ``Refueling Operations,'' and associated 
surveillance requirements are proposed for revision to incorporate the 
design basis requirements for refueling operations and to correspond to 
NUREG-1432.
    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 will incorporate new requirements for the 
control room air conditioning system, control room filtration 
system, and refueling operations. In addition, the proposed change 
will ensure that the Limiting Condition for Operations and 
surveillance requirements are consistent with the design basis of a 
fuel handling accident as documented in the FCS Updated Safety 
Analysis Report (USAR).
    CONTROL ROOM SYSTEMS
    The control room air conditioning (A/C) system consists of two 
redundant A/C units, VA-46A and VA-46B. Each unit has sufficient 
capacity to meet the cooling requirements for personnel and 
equipment inside the control room envelope. Each A/C unit is 
equipped with an air-cooled condenser located inside a protective 
enclosure outdoors on the roof of the Auxiliary Building. Each A/C 
unit's refrigerant compressor, air cooling coils, fans, and dampers 
are located inside of the control room envelope. Each unit has a 
waterside economizer coil that allows air cooling with Component 
Cooling Water (CCW). When cooling water temperature is sufficiently 
low, a temperature-activated valve at each A/C unit allows cooling 
water flow through the waterside economizer. This valve also diverts 
flow away from the waterside economizer if cooling water temperature 
is too high. The air-operated CCW isolation valves to the A/C units 
fail closed and are automatically closed on a Ventilation Isolation 
Actuation Signal (VIAS) to prevent CCW flow through the waterside 
economizers in a post-accident situation.
    Technical Specification (TS) 2.12(1) requires that the 
temperature within the control room and control cabinets be 
maintained below 120 deg.F does not meet any of the four criteria 
contained in 10 CFR 50.36 for inclusion in TS. However, the 
equipment required to maintain this temperature, the control room 
air conditioning system, meets Criterion 3 of 10 CFR 50.36 in that 
the system functions to mitigate a design basis accident by 
maintaining the control room in a habitable environment.
    Therefore, it is proposed that this TS be revised to delete the 
control room temperature as a LCO and require that two control room 
air conditioning trains be operable when the reactor coolant 
temperature is above 210 deg.F. The design temperature limits of 
instrumentation and controls inside of the control room will be 
maintained in the Basis Section of TS 2.12.
    The allowed outage time for one train of control room air 
conditioning is proposed as 30 days. This is consistent with 
Combustion Engineering Standard TS 3.7.12 (NUREG-1432 Rev. 1). In 
addition, the FCS Probabilistic Risk Assessment model was reviewed 
and validated a 30 day outage time as being non-risk significant. 
The impact on Core Damage Frequency (CDF) from a 30 day LCO was 
based on the assumption that one cooling unit was always inoperable 
and thus under the LCO for an entire year. This allows the analysis 
to consider unlimited entries into the LCO and a full LCO duration 
for each entry. Using this assumption, the baseline (annually) CDF 
of 1.53E-5 would increase by 21.6% to a frequency of 1.86E-5. In 
accordance with EPRI's ``PSA Applications Guide,'' this small 
increase in CDF can be classified as ``non-risk significant.''
    Specification 2.12(2)
    Specification 2.12(2) requires that a thermometer be in the 
control room at all times. This instrumentation does not meet any of 
the four criteria contained in 10 CFR 50.36 for inclusion in the FCS 
TS. Therefore, the requirement is proposed for relocation to the FCS 
USAR.
    Specification 2.12(3)
    Specification 2.12(3) requires that all areas of the plant 
containing safety related instrumentation be observed during hot 
functional testing to determine local temperatures and monitored 
during operation if normal plant ventilation is not available. It is 
proposed to delete this TS. The requirement to monitor and determine 
local temperatures during hot functional testing was met during the 
initial startup phase of FCS and is no longer applicable. The 
requirement to monitor temperatures within the plant during normal 
operation does not meet any of the four criteria contained in 10 CFR 
50.36 for inclusion in TS and therefore is being deleted.
    The requirement to control temperatures for safety related 
instrumentation and controls, and initiate supplementary cooling if 
required, is currently described in USAR Section 9.10. These USAR 
requirements are controlled by plant procedures. Any changes to 
these requirements would require an evaluation be conducted in 
accordance with 10 CFR 50.59.
    Specification 2.12(4)
    Specification 2.12(4) allows one control room air filtration 
system to be inoperable for 7 days or a plant shutdown be commenced. 
This specification does not state which modes of operation it 
applies to.
    Therefore, it is proposed to revise this specification to 
require two trains of control room air filtration systems to be 
operable when the reactor coolant temperature is above 210 deg.F. 
The allowed outage time will be maintained at 7 days and a total of 
42 hours will be allowed to take the plant to cold shutdown. The 42 
hour time period is consistent with TS 2.0.1 which addresses 
equipment outages in excess of what is specifically allowed by 
individual specifications.
    The proposed changes for the control room systems consist of 
providing additional restrictions on operation of the control room 
air filtration systems and control room air conditioning system. 
These changes ensure that equipment required to mitigate the 
consequences of an accident are operable. Therefore, the proposed 
changes do not increase the probability or consequences of an 
accident previously evaluated.
    REFUELING OPERATIONS
    The design bases of the fuel handling accident and refueling 
operations were reviewed and several inadequacies were identified 
related to refueling operations. Therefore, revisions are proposed 
for the TS Definition section, TS 2.6 on containment integrity, and 
TS 2.8 on refueling operations to reflect NUREG-1432.
    Definitions
    Cold Shutdown Condition & Refueling Shutdown Condition
    The changes proposed for the definitions of Cold Shutdown 
Condition, and Refueling Shutdown Condition clarify these 
definitions. The plant is in Cold Shutdown when Tcold is 
less than 210 deg.F, and the reactor coolant is at least Shutdown 
Boron Concentration but less than Refueling Boron Concentration. 
Similarly, the definition for Refueling Shutdown is clarified to 
apply when Tcold is less than 210 deg.F and the reactor

[[Page 30640]]

coolant is at least Refueling Boron Concentration. This change does 
not propose any new operating modes but merely clarifies when the 
definitions are applicable.
    Core Alterations
    The definition for Core Alterations is being revised to reflect 
the requirements of NUREG-1432. This revision deletes ``any 
component'' from the definition and clarifies that the components 
considered by this definition are those that could affect 
reactivity. In addition, the revision adds nuclear fuel to the 
definition such that movement of fuel within the reactor vessel will 
be defined as a core alteration and not a refueling operation.
    Refueling Operations
    The definition of Refueling Operations is being revised to 
delete control element assemblies (CEA) or startup sources from the 
definition since these are items that are included in the definition 
of Core Alterations. Additionally, it is being revised to specify 
that the definition is limited to movement of irradiated fuel 
outside of the reactor pressure vessel since fuel movement inside 
the reactor vessel is included in the definition of Core Alteration. 
Finally, a clarification is being added to state that suspension of 
refueling operations shall not preclude completion of movement of 
irradiated fuel to a safe, conservative position.
    In Operation
    The definition of In Operation is being revised to include the 
definition of operable. This is a more conservative interpretation 
than currently exists.
    Specification 2.1 ``Reactor Coolant System''
    It is proposed to revise TS 2.1.1(3) to include shutdown cooling 
requirements when the reactor coolant system (RCS) temperature is 
below 210 deg.F with fuel in the reactor and the reactor vessel head 
fully tensioned. The definitions of Cold Shutdown (Mode 4) and 
Refueling Shutdown (Mode 5) contained in the TS make no distinction 
as to the status of the reactor vessel head or RCS temperature. The 
only difference between the two defined modes is boron 
concentration. Higher or lower boron concentration affects shutdown 
margin but does not affect decay heat load, which is the basis for 
this specification.
    Technical Specification 2.1.1(4) was intended to address 
shutdown cooling requirements during refueling operations. However, 
this is already addressed in TS 2.8. Therefore, it is proposed to 
delete TS 2.1.1(4) and the exception since new specifications 
addressing shutdown cooling loop requirements during Mode 5 with 
fuel in the reactor and with one or more reactor vessel head closure 
bolts less than fully tensioned are proposed for inclusion in TS 2.8 
(Refueling Operations).
    The associated statements supporting these items in the Basis 
section are also proposed for deletion. Prior to any reactor vessel 
head closure bolts being loosened, TS 2.1.1 will be applicable which 
will require two shutdown cooling loops. As soon as a closure bolt 
is loosened, the new proposed TS 2.8 would be applicable which also 
requires two shutdown cooling loops whenever there is less than 23 
feet of water above the core. The requirements of TS 2.1.1(3) are 
similar to NUREG-1432, Specifications 3.4.7 and 3.4.8.
    Specification 2.6 ``Containment System''
    Currently, TS 2.6(1)c states that containment integrity shall 
not be violated when the reactor vessel head is removed if the boron 
concentration is less than refueling concentration. However, 
Specification 2.6(1)c has no required actions and therefore, TS 
2.0.1 must be entered when the LCO is not met. In this situation, 
(reactor vessel head removed), TS 2.0.1 is ineffective because the 
plant would already be in Refueling Shutdown. Thus, TS 2.6(1)c is 
proposed for deletion.
    Currently, Specification 2.6(1)d requires that except for 
testing one control element drive mechanism at a time, positive 
reactivity changes shall not be made by CEA motion or boron dilution 
unless containment integrity is intact. Specification 2.6(1)d is 
proposed for deletion as it is unnecessarily restrictive.
    Specification 2.8.1(1) as proposed eliminates the need for 
containment integrity when the reactor is in Refueling Shutdown. 
Specification 2.8.1(1) requires sufficient shutdown margin to 
preclude a criticality event and also prescribes actions to restore 
the shutdown margin if necessary. Small positive reactivity 
increases whether by CEA motion or boron dilution will not cause a 
criticality event due to the need to maintain at least a 5% shutdown 
margin. Therefore, the requirement to maintain containment integrity 
is unnecessarily restrictive since a criticality event cannot occur 
when a shutdown margin of at least 5% exists. Specification 2.8.1(1) 
is consistent with the requirements of NUREG-1432, Specification 
3.9.1.
    A new specification (TS 2.8.2(1)) is proposed that provides 
requirements for containment closure during core alterations and 
refueling operations inside of containment. The design basis of the 
Fort Calhoun Station does not require full containment integrity 
during a fuel handling accident. As stated in USAR Section 14.18, 
the fuel handling accident does not take credit for containment 
isolation. Therefore, requiring full containment integrity is 
inappropriate and requirements for containment closure are proposed 
for addition to TS 2.8 consistent with NUREG-1432 Specification 
3.9.2.
    Specification 2.10.2 governs operation of CEAs and monitoring of 
selected core parameters. Specification 2.10.2 ensures (1) adequate 
shutdown margin following a reactor trip, (2) that the moderator 
temperature coefficient (MTC) is within the limits of the safety 
analysis, and (3) CEA operation is within the limits of the setpoint 
and safety analysis. Specification 2.10.2 ensures that the reactor 
will be maintained sufficiently subcritical to preclude inadvertent 
criticality and provides actions (i.e., boration) to be taken to 
ensure that the required shutdown margin is available. Thus, TS 
2.10.2 precludes the need for containment integrity when the plant 
is in cold shutdown.
    Specification 2.8 ``Refueling Operations''
    It is proposed that TS 2.8 be rewritten to reflect NUREG-1432. 
Currently, this specification applies to any refueling operation. 
However, no distinction is made between refueling operations within 
containment and refueling operations within the spent fuel pool. In 
addition, several initial assumptions of a fuel handling accident 
are not addressed by the current TS 2.8.
    Specification 2.8(1)
    The current TS 2.8(1) is inadequate. This specification requires 
that the equipment hatch and one door in the Personnel Air Lock be 
properly closed, and all automatic containment isolation valves be 
operable or at least one valve closed. The specification does not 
define what is meant by a properly closed equipment hatch; that 
information is currently contained in the Basis of TS 2.1.1. In 
addition, inclusion of all automatic containment isolation valves 
instead of those providing direct access to the outside atmosphere 
is incorrect.
    The containment isolation system is defined in USAR Section 
5.9.5 as those devices actuated by a Containment Isolation Actuation 
Signal (CIAS) or a Steam Generator Isolation Signal (SGIS). This 
includes many valves that have no design basis function during a 
fuel handling accident. A CIAS is initiated by a Containment 
Pressure High Signal or a Pressurizer Pressure Low Signal. Neither 
of these signals are required to be operable during refueling 
operations as these signals would/could not respond to a fuel 
handling accident.
    The correct requirements are specified in TS 2.8(2) which only 
requires that closure be initiated by the Ventilation Isolation 
Actuation Signal (VIAS) for the containment pressure relief, air 
sample, and purge system valves. Due to these inadequacies, it is 
proposed to delete TS 2.8(1) and replace it with a new Specification 
2.8.2(1) which is consistent with NUREG-1432 Specification 3.9.3.
    Specification 2.8(2)
    It is proposed that TS 2.8(2) be deleted and replaced by new 
Specifications 2.8.2(3) and 2.8.3(5). The requirement to maintain an 
operable Ventilation Isolation Actuation Signal with input from the 
containment atmosphere gaseous and auxiliary building exhaust stack 
gaseous radiation monitors is consistent with current requirements 
and required actions are consistent with NUREG-1432, Specification 
3.3.8. Radiation Monitor RM-052 functions as a ``swing'' monitor, 
i.e., it can be aligned to monitor either containment or the 
auxiliary building exhaust ventilation stack. Radiation Monitor RM-
052 is powered by either MCC-3B1/AI-40C (like RM-051) or MCC-4C2/AI-
40D (like RM-062).
    Technical Specification 2.7, Electrical System is not required 
to be applied when the RCS is below 300 deg.F. Above 300 deg.F, TS 
2.7 requires both 4160-VAC buses to be operable. Thus, above 
300 deg.F the required radiation monitors must be powered from 
independent 480-VAC buses supplied by independent 4160-VAC buses. 
During refueling outages, bus alignments other than those used 
during power operation are used to permit electrical system 
maintenance and modifications.
    In the loss of offsite power event, the radiation monitor sample 
pumps and control room HVAC units stop and will not restart

[[Page 30641]]

until the emergency diesel generators (EDGs) reenergize the system. 
The fuel handling equipment also stops and does not restart when the 
EDGs reenergize the system, thus minimizing the potential of a fuel 
handling accident. When the EDGs reenergize the buses, VIAS will 
operate as designed. Therefore, when the RCS is below 300 deg.F, the 
required monitors need only be powered from independent 480-VAC 
buses supplied by a single 4160-VAC bus.
    There is no need to assume that a fuel handling accident occurs 
immediately followed by a loss of offsite power. However, in the 
unlikely event that this should occur, there would be no effect on 
the site boundary dose since VIAS is not credited in USAR Section 
14.18 (Fuel Handling Accident). In this situation, when the EDGs 
reenergize the buses, the control room HVAC units will restart in 
the filtered air makeup mode and the stack radiation monitor sample 
pump will restart. However, the containment radiation monitor sample 
lines remain isolated preventing the restart of the monitor sample 
pump after receipt of a VIAS.
    Specification 2.8(3)
    It is proposed that TS 2.8(3) be deleted. This requirement does 
not meet any of the four criteria contained in 10 CFR 50.36 for 
inclusion in the TS. The requirement that radiation levels in 
containment and the spent fuel pool shall be monitored during 
refueling operations will be incorporated into the FCS USAR.
    Specification 2.8(6)
    It is proposed that TS 2.8(6) be deleted. This requirement does 
not meet any of the four criteria contained in 10 CFR 50.36 for 
inclusion in the TS. The requirements that direct communication 
between personnel in the control room and at the refueling machine 
shall be available whenever core alterations are taking place will 
be incorporated into the FCS USAR.
    Specification 2.8(7)
    It is proposed that TS 2.8(7) be deleted and replaced with a new 
Specification 2.8.3(4). The requirement to place the spent fuel pool 
ventilation system in operation prior to refueling operations is 
consistent with the current TS. It is being clarified that this 
specification only applies to refueling operations in the spent fuel 
pool, and not when conducting refueling operations inside of 
containment. Additionally, it is being clarified that TS 2.0.1 is 
not applicable to this activity, as reactor operation is independent 
of fuel movements in the spent fuel pool.
    Specification 2.8(9)
    The current Specification 2.8(9) is inadequate. This 
specification requires a minimum of 23 feet of water above the top 
of the core. This does not meet the initial conditions assumed in 
the fuel handling accident as documented in USAR Section 14.18. USAR 
Section 14.18 assumes 23 feet of water above where the fuel could 
land if dropped. In order to meet this initial condition, a minimum 
of 23 feet of water above the reactor vessel flange is required, as 
this is the highest point where a fuel bundle could land if dropped. 
Procedures reflect the requirement to maintain 23 feet of water 
above the reactor vessel flange during refueling operations. The 
proposed revision is consistent with NUREG-1432, Specification 
3.7.16.
    Specification 2.8(11)
    The current specification is inadequate. The specification 
provides restrictions on storage of fuel in the spent fuel pool; 
however, there are no required actions to address situations when 
the specification is not met. It is proposed that TS 2.8(11) be 
deleted and replaced with a new Specification 2.8.3(1) that requires 
that a misloaded fuel assembly be moved immediately. Additionally, 
it is being clarified that TS 2.0.1 is not applicable to this 
activity, as reactor operation is independent of fuel movements in 
the spent fuel pool.
    Specification 2.8(12)
    It is proposed that TS 2.8(12) be deleted and replaced with a 
new Specification 2.8.3(3). The requirement to maintain 500 ppm 
boron concentration in the spent fuel pool whenever unirradiated 
fuel is stored there is consistent with the current TS and the 
required actions are consistent with NUREG-1432, Specification 
3.7.17.
    Restriction on Movement of Irradiated Fuel from the Reactor Core
    The restriction on irradiated fuel movement unless the core has 
been subcritical for at least 72 hours if the reactor has been 
operated at power levels above 2% is proposed for relocation to the 
Bases of TS 2.8.2(2). This requirement does not meet any of the four 
criteria contained in 10 CFR 50.36 for inclusion in the TS. This is 
consistent with NUREG-1432, B 3.9.6.
    Reactor Coolant System Boron Concentration
    Currently, there is no specification for boron concentration. 
Refueling boron concentration is included in the definition of Mode 
5. However, there are no required actions to be taken if the boron 
concentration should be below refueling concentration. Therefore, it 
is proposed that a new Specification 2.8.1(1) be incorporated 
consistent with NUREG-1432, Specification 3.9.1.
    Spent Fuel Pool Water Level
    Currently, there is no specification for spent fuel pool water 
level. The water level of the spent fuel pool is an initial 
condition assumed in USAR Section 14.18. It is proposed that a new 
Specification 2.8.3(2) be incorporated into TS 2.8, which is 
consistent with NUREG-1432, Specification 3.7.16.
    The proposed changes for the RCS and containment during 
shutdown, and requirements for refueling operations, consist of 
providing additional restrictions on operation, and changes to make 
the requirements of the TS Limiting Conditions for Operation 
consistent with the initial conditions and assumptions of the fuel 
handling accident as documented in USAR Section 14.18. Therefore, 
the proposed changes do not increase the probability or consequences 
of an accident previously evaluated.
    SURVEILLANCE REQUIREMENTS
    CONTROL ROOM
    Specification 3.1, Table 3-3, Item 13.
    Specification 3.1, Table 3-3, Item 13 requires that the 
thermometer in the control room be compared with a calibrated 
thermometer and replaced if out of tolerance on a refueling
    frequency. It is proposed that this surveillance be deleted to 
be consistent with deletion of the LCO requirement to maintain a 
thermometer in the control room.
    A new surveillance is proposed to verify that the control room 
air conditioning system has the capability to remove the assumed 
heat load. This surveillance will ensure the operability 
requirements for TS 2.12 are met. The test and frequency is 
consistent with NUREG-1432.
    The air-operated CCW isolation valves to the A/C units fail 
closed and are automatically closed on a VIAS to prevent CCW flow 
through the waterside economizers in a post-accident situation. 
These valves are currently tested in accordance with TS 3.3 (FCS 
Inservice Testing Program). Prior to the modification, the valves 
were tested as fail-open valves. No TS changes are necessary.
    The control room air filtration system is currently tested on a 
refueling frequency in accordance with TS 3.2, Table 3-5, Item 10a. 
No TS changes are necessary.
    REFUELING OPERATIONS
    Reactor Coolant Boron Concentration During Refueling Operations
    The Reactor Coolant System boron concentration is currently 
sampled in accordance with TS 3.2, Table 3-4, Item 1(e). It is 
proposed to revise the frequency from once per shift during 
refueling operations to once per 3 days which is consistent with 
NUREG-1432. As stated in the basis of TS 2.8 and USAR Section 14.18, 
the reactor cavity is filled with over 200,000 gallons of borated 
water prior to the start of refueling operations. The requirements 
for sampling the reactor coolant during the remainder of Mode 5 is 
performed once per 3 days in accordance with Table 3-4, Item 1(d). 
This proposed change will make the sampling consistent with the 
requirements of Item 1(d) and NUREG-1432.
    Spent Fuel Pool Boron Concentration
    The spent fuel pool boron concentration is currently sampled in 
accordance with TS 3.2, Table 3-4, Item 5. It is proposed to revise 
the frequency of the sampling to prior to movement of unirradiated 
fuel in the spent fuel pool and once per week whenever unirradiated 
fuel is stored there to be consistent with the requirements of the 
LCO.
    Source Range Neutron Monitors
    Currently, a channel check and calibration of the wide range 
neutron monitors is performed in accordance with TS 3.1, Table 3-1, 
Item 2.
    Containment Penetrations
    Currently, there is no surveillance to determine the status of 
containment penetrations during refueling operations. Therefore, a 
new surveillance is proposed for TS 3.2, Table 3-5 to verify the 
status of required containment penetrations once per 7 days 
consistent with NUREG-1432.
    The requirement of NUREG-1432 to verify that the containment 
purge and exhaust valves actuate to the isolation position on a 
refueling frequency is currently tested as part of the Containment 
Radiation High Signal test required by TS 3.1, Table 3-2. Item 4.
    Shutdown Cooling Loops
    Currently, there is no surveillance requirement to verify that 
the required

[[Page 30642]]

shutdown cooling loops are operable and in operation or to verify 
correct breaker lineup for the shutdown cooling loop that is not in 
operation. Therefore a new surveillance is proposed to be 
incorporated into TS 3.2, Table 3-5 consistent with NUREG-1432.
    Refueling Water Level
    Currently, there is no surveillance requirement to verify the 
refueling water level during refueling operations. Therefore, a new 
surveillance is proposed for incorporation into TS 3.2, Table 3-5 
consistent with NUREG-1432.
    Spent Fuel Pool Water Level
    Currently, there is no surveillance requirement to verify the 
spent fuel pool water level during refueling operations. Therefore, 
a new surveillance is proposed for incorporation into TS 3.2, Table 
3-5 consistent with NUREG-1432.
    Spent Fuel Initial Enrichment/Burnup Verification
    Currently, the requirement to conduct a verification of initial 
enrichment and burnup of spent fuel that will be stored in Region 2 
is included as a general requirement of TS 2.8. It is proposed to 
relocate this requirement into a surveillance in TS 3.2, Table 3-5, 
consistent with NUREG-1432.
    The proposed changes for the surveillance requirements consist 
of providing additional testing requirements to ensure that the 
Limiting Condition for Operations will be met. One surveillance 
frequency related to the sampling of the reactor coolant system 
boron concentration during refueling operations is being reduced 
from a frequency of once per shift to once every 3 days. However, 
this frequency is consistent with the frequency of sampling during 
the remainder of Mode 5 when fuel is in the
    reactor and is more than adequate due to the large volume (over 
200,000 gallons) of borated water required during refueling 
operations. Therefore, the proposed changes do not increase the 
probability or consequences of an accident previously evaluated.
    ADMINISTRATIVE CHANGES
    The remainder of TS 2.8 requirements of refueling operations are 
proposed to be reformatted into individual TS LCOs. It is also 
proposed that sampling frequencies of items contained in TS 3.2, 
Table 3-4, (page 3-19), be revised to incorporate frequencies 
defined in TS 3.0.2. Therefore, frequencies stated as once per 31 
days will be noted as ``M,'' and frequencies stated as once per 7 
days will be noted as ``W.'' These proposed changes have no effect 
on the probability or consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There will be no physical alterations to the plant 
configuration. No changes in operating modes are proposed although 
minor changes to the definitions of Cold Shutdown Condition and 
Refueling Shutdown Condition are proposed for clarification 
purposes. The proposed changes incorporate additional restrictions 
on the operation and testing of equipment required to mitigate an 
accident and to ensure the initial conditions and assumptions of the 
design basis accidents are maintained and controlled by the 
Technical Specifications.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes ensure that assumptions of the fuel 
handling accident are maintained by Technical Specification Limiting 
Condition for Operation and surveillance requirements. The 
assumptions of the fuel handling accident that may affect a margin 
of safety are not being changed. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

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

    Date of amendment request: February 25, 1997
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would amend the Limerick Generating Station 
(LGS) Unit 1 and Unit 2 Facility Operating Licenses (FOLs), and 
Appendix B of the licenses (i.e., Environmental Protection Plan (EPP)), 
reflecting a corporate name change from Philadelphia Electric Company 
to PECO Energy Company. In addition, the application would make changes 
to the LGS Units 1 and 2, FOL, and Appendix A (i.e., TS) of the 
licenses, which would remove obsolete information and correct 
typographical errors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The company name change and typographical corrections are 
editorial and will not alter the operation of equipment assumed to 
be an initiator of any analyzed event or transients previously 
evaluated. The license provisions were satisfactorily completed, and 
as such, have no effect on any previously evaluated accident 
scenario. The changes will not alter the operation of equipment 
assumed to be available for the mitigation of accidents or 
transients, nor will they alter the operation of equipment important 
to safety previously evaluated.
    Therefore, the changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The company name change and typographical corrections are 
editorial and will not involve any physical changes to the plant 
systems, structures, or components. The license provisions were 
satisfactorily completed, and as such, have no effect on any 
previously evaluated accident scenario. The proposed changes do not 
allow plant operation in any mode that is not already evaluated. The 
changes will not alter the operation of equipment important to 
safety previously evaluated.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The company name change and typographical corrections are 
editorial and will not affect the manner in which the facility is 
operated, or change equipment or features which affect the 
operational characteristics of the facility. There is no margin of 
safety as defined in the bases of any TS regarding the name of the 
company, or affected by the corrections or deletion of obsolete 
license provisions.
    Therefore, these proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: John F. Stolz

[[Page 30643]]

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

    Date of amendment request: March 24, 1997
    Description of amendment request: The proposed Technical 
specifications (TS) changes would delete the Drywell and Suppression 
Chamber Purge System operational time limit, and add a surveillance 
requirement to ensure the purge system large supply and exhaust valves 
are closed as required.
    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 changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    This activity does not increase the probability of occurrence of 
an accident previously evaluated in the SAR [Safety Analysis 
Report]. This activity involves deleting the allowable operating 
limit (180 hours each 365 days) for the Drywell and Suppression 
Chamber Purge system, while maintaining specific criteria for when 
the valves are allowed to be open. These changes do not increase the 
probability that this system will be in service should a LOCA [loss-
of-coolant-accident] occur and does not increase the probability 
that a LOCA will occur. These changes also do not impact the 
probability of occurrence of any anticipated operational occurrence, 
other postulated design basis accident, or other event in which the 
plant was designed to respond.
    This activity does not increase the consequences of an accident 
previously evaluated in the SAR. UFSAR [Updated Final Safety 
Analysis Report] Section 9.4.5.1.2.2 for high volume purging, 
although limiting the operating time the vent and purge system is to 
be in service, evaluates the consequences of a LOCA should the vent 
and purge valves be open. System operating procedures for venting 
and purging assure the availability of SGTS [standby gas treatment 
system] should a LOCA occur.
    This activity will not increase the probability of a LOCA 
occurring during the time the Drywell and Suppression Chamber Purge 
system is in operation as previously evaluated. The Improved TS do 
not identify a specific time limit value as long as the valves are 
operated under the stated conditions (inerting, de inerting, 
pressure control, ALARA [as low as reasonably achievable] or air 
quality considerations for personnel entry or Surveillances that 
require that the valves be open). These proposed changes will 
incorporate the ITS [Improved Technical Specifications] operational 
controls which will result in the same order of magnitude of 
equipment malfunction probability as that provided by limiting 
purging to 180 hours per 365 days. A LGS [Limerick Generating 
Station] Level 2 PSA [Probability Risk Assessment] Analysis was 
performed to determine the additional risk associated with changing 
the operating limit from 90 hours to a nominal 500 hours each 365 
days. This analysis concluded that the increase in risk of 
containment failure is well within the bounds of the EPRI [Electric 
Power Research Institute] PSA Applications Guideline for permanent 
changes and the NRC [Nuclear Regulatory Commission] Staff's safety 
goal value of 1.0 E-6 per year of reactor operation. Industry and 
LGS historical operating experience confirms that the purging lines 
are opened only for the specified reasons stated in ITS and for 
periods which do not exceed the current magnitude of equipment 
malfunction probability. Therefore, earlier engineering judgment is 
being replaced by operating experience.
    Failure of the operating SGTS filter bank following a LOCA has 
been found to be acceptable due to the limited benefit derived from 
SGTS for accident sequences important to plant risk and the 
possibility that the backup filter bank would be available. 
Additionally, as discussed in UFSAR Section 9.4.5.1.2.2, the failure 
of SGTS during a LOCA does not contribute to any significant 
releases and is bounded by the analysis performed to address 
containment overpressure rupture.
    Deleting the time limit restriction that the vent and purge line 
isolation valves may be open does not increase the probability that 
these valves will not perform as designed (close upon isolation 
signal) in response to a LOCA. Removing the 180 hour requirement 
will not increase the likelihood that the vent and purge valves will 
be called upon to close from that previously evaluated. UFSAR 
Section 6.2 states that the containment purge valves have undergone 
extensive testing and analyses to demonstrate the operability of 
these valves following a LOCA.
    These changes do not directly or indirectly degrade the 
performance of any other safety system (assumed to function in the 
accident analysis) design basis. The potential for other equipment 
failures in the reactor enclosure due to duct impact, impingement, 
and the resulting environmental conditions was previously evaluated 
in the LGS SAR. It was concluded that the environmental 
qualifications for the LGS equipment are sufficient to ensure 
operability under the predicted environmental condition, and, the 
potential does not exist for impact or impingement - related damage 
to essential equipment. Maintaining the existing SAR analysis and 
retaining operating criteria for opening the containment purge 
valves, demonstrates that the risk of equipment failure and 
resulting radiological consequences will not increase.
    Therefore, deleting the TS operating limit for the Drywell and 
Suppression Chamber Purge system from 180 hours each 365 days and 
the addition of a TS Surveillance Requirement verifying that the 
purge valves are closed under certain conditions does not increase 
the probability or consequences of an accident previously evaluated.
    2. The proposed Technical Specifications changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    This activity does not change the function of the Drywell and 
Suppression Chamber Purge system, the containment isolation system, 
or SGTS as previously evaluated. Deleting the operational time limit 
that the vent and purge system is in service and the addition of a 
surveillance requirement does not create an accident initiator not 
already considered.
    In addition, this activity does not create a failure mode not 
considered. All evaluated equipment failures that could occur as a 
result of a LOCA during high volume purging have previously been 
identified and evaluated. Therefore, these changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed Technical Specifications changes do not involve 
a significant reduction in a margin of safety.
    The bases of TS 3.6.1.8 state that the 180 hour each 365 day 
operating limit for the Drywell and Suppression Chamber Purge system 
is imposed to protect the integrity of the SGTS filters. The LGS 
Offsite Dose Calculation Manual assures the availability of the 
backup SGTS filter train during operation of the vent and purge 
system. Furthermore, deleting the operating limit (180 hours each 
365 days) does not reduce the margin of safety since specific 
criteria for opening the purge valves is being maintained and does 
not involve an increase in risk. Therefore, the proposed changes do 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: 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: April 9, 1997
    Description of amendment request: The proposed Technical 
Specifications (TS) changes would clarify existing battery specific 
gravity requirements, delete the requirement to correct specific 
gravity values based on electrolyte level, and allow the use of 
charging current measurements to verify the batterys state of charge.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 30644]]

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 changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    Changes to Technical Specifications surveillance requirements 
for specific gravity and Technical Specifications Bases commitments 
do not change the frequency or consequences of any accident 
previously evaluated. The proposed changes which commit to IEEE 
Standard 450-1995 for specific gravity testing, providing battery 
charging current as an alternate method to specific gravity 
measurements, and eliminating the commitment to perform electrolyte 
level correction do not prevent the DC system from performing its 
intended safety function. The proposed changes to the Technical 
Specification battery surveillance requirements and commitment to 
IEEE Standard 450-1995 for specific gravity are in accordance with 
current industry practices. These changes do not reduce the 
readiness and performance of the 1E DC power system to perform its 
intended function during a design basis event.
    The proposed changes do not affect seismic specifications, 
separation criteria or environmental qualifications. The proposed 
changes do not impose an increase in or more severe test 
requirements, an increase in the frequency of operation, reduce 
independence or redundancy, modify the system or equipment 
protective features, introduce new equipment failures or impose 
additional loads than any previously evaluated. The Class 1E battery 
system will continue to meet all of the design standards applicable 
to the system and will not cause the system to operate outside of 
its design or testing limits.
    Batteries or battery chargers and their failure are not 
initiators of the accidents previously evaluated. The proposed 
changes do not affect, degrade or prevent the response of active or 
passive systems described or assumed in the LGS accidents previously 
evaluated. In addition, the proposed TS changes will improve the 
availability of the station batteries.
    Therefore, the changes will not increase the probability or 
consequences of an accident previously evaluated.
    2. The proposed Technical Specifications changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The proposed Technical Specifications changes which will revise 
the surveillance requirements and the TS Bases, do not increase the 
failure rate of the battery. The proposed changes clarify and 
enhance Operation's focus on the key battery parameters which will 
improve the availability of the station batteries. The station 
batteries are not accident initiators. The single failure of an 
electrical component was previously evaluated in the LGS accident 
analysis. Unexpected failures beyond the postulated single failure 
are no more likely to occur under the clarified surveillance 
requirements.
    Therefore, these changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed Technical Specifications changes do not involve 
a significant reduction in a margin of safety.
    The revision clarifies and reduces the battery surveillance 
requirements for specific gravity. The revision eliminates the 
possibility for misinterpretation and provides consistency of the 
surveillance requirements. The specific gravity value for each 
connected cell is being revised to reflect a discrete number which 
meets the existing manufacturer's recommendations and does not 
differ from the value described in the present bases. LGS is 
currently committed to earlier revisions of IEEE Standard 450 (i.e., 
1975 and 1980), and the incorporation of IEEE Standard 450-1995 for 
specific gravity will reflect current industry practices regarding 
specific gravity.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, PA 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: April 30, 1997 (TS 97-01)
    Description of amendment request: The proposed amendment would 
change the design features section of the Technical Specifications to 
provide for insertion of Lead Test Assemblies (LTAs) containing Tritium 
Producing Burnable Absorber Rods (TPBARs) in the Watts Bar Nuclear 
Plant (WBN) reactor core during Cycle 2. The purpose of the change is 
to provide irradiation services to support U.S. Department of Energy 
(DOE) investigations into the feasibility of using commercial light 
water reactors to maintain the DOE inventory of tritium.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    LTAs do not adversely affect reactor neutronic or thermal-
hydraulic performance; therefore, they do not significantly increase 
the probability of accidents or equipment malfunctions while in the 
reactor. The neutronic behavior of the LTAs mimics that of standard 
burnable absorbers with only slight differences which are 
accommodated in the core design. The reload safety analysis 
performed for WBN Unit 1, Cycle 2 will confirm that any minor 
effects of LTAs on the reload core will be within established fuel 
design limits.
    As described in DOE Technical Report PNNL-11419, Revision 1, the 
LTA design is robust to all accident conditions except the large 
loss of coolant accident where the rods are susceptible to failure. 
However, the failure of the small number of TPBARS rods has been 
determined to have an insignificant effect on the thermal hydraulic 
response of the core to this event.
    The impacts of LTAs on the radiological consequences for certain 
postulated events [as shown in Table 6-1 of the licensee's 
submittal, including Large Break LOCAs{time}  are very small, and 
they remain within 10 CFR 100 regulatory limits. The additional 
offsite doses due to tritium leakage from the containment are small 
with respect to loss of coolant accident source terms and are well 
within regulatory limits.
    The LTAs will not result in an increase in combustible gas 
released to the containment. Therefore, the LTAs do not result in a 
significant increase in the consequences of those previously 
considered.
    Analysis has shown that TPBARs will not fail during Condition I 
through IV events, with the exception of a Large Break LOCA. The 
radiological consequences of the non-Large-Break LOCA events are 
essentially unchanged by the expected TPBAR tritium leakage to 
reactor coolant, and doses remain within a small fraction of 10 CFR 
100 regulatory limits. Therefore, there is no significant increase 
in the consequences of these previously evaluated accidents.
    The expected occupational and offsite doses, as reported in 
Technical report PNNL-11419, Revision 1, resulting from release of 
tritium from TPBARs over the plant operating cycle, including 
refueling, are not significantly increased and are within applicable 
regulatory limits.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    LTAs have been designed to be compatible with existing 
Westinghouse 17x17 fuel assemblies and conventional Burnable Poison 
Rod Assembly (BPRA) handling tools, equipment, and procedures, and 
therefore no new accidents or equipment malfunctions are created by 
the handling of LTAs.
    LTAs use materials with known and predictable performance 
characteristics and are compatible with PWR coolant. The LTA design 
has specifically included material similar to those used in standard 
burnable absorber rods with the exception of internal

[[Page 30645]]

assemblies used in the production and retention of tritium. As 
described in the technical report, these materials are compatible 
with the reactor coolant system and the core design. For the 
irradiation proposed, the quantities of these materials is small. 
Therefore, no new accidents or equipment malfunctions are created by 
the presence of the LTAs in the reactor coolant system.
    Thermal-hydraulic criteria have been established to ensure that 
TPBARs will not fail during Condition I or II events. Analysis has 
shown that TPBARs, appropriately positioned in the core, operate 
within the established thermal-hydraulic criteria. Therefore, no new 
accidents or equipment malfunctions are created by the presence of 
the LTAs in the reactor.
    Analysis has shown that TPBARs will not fail during Condition 
III and IV events, with the exception of a large-break loss-of-
coolant-accident. The radiological consequences of these events are 
small, with doses that are a small fraction of the 10 CFR 100 
limits. Therefore there is no significant increase in consequences 
of these previously evaluated accidents.
    LTAs do not adversely affect reactor neutronic or thermal-
hydraulic performance; therefore, they do not create the possibility 
of accidents or equipment malfunctions of a different type than 
previously evaluated while in the reactor.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    LTAs do not adversely affect reactor neutronic or thermal-
hydraulic performance. Analysis indicates that reactor core behavior 
and offsite doses remain relatively unchanged. TPBAR performance 
under Condition I, II, III, and IV events are very similar to 
standard burnable absorber rods previously evaluated. For these 
reasons, the proposed amendment does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: April 18, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3/4.3.2, ``Safety System 
Instrumentation,'' and TS Section 3/4.5.2, ``Emergency Core Cooling 
Systems - ECCS Subsystems - Tavg (greater than or equal to) 
280 deg.F.'' Certain surveillance intervals would be changed from 18 
months to once each refueling interval, and certain setpoints would be 
changed. The associated bases would also be changed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS) has 
reviewed the proposed changes and determined that a significant 
hazards consideration does not exist because operation of the DBNPS, 
in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because the initiation of such 
accidents are not affected by the proposed revisions to increase the 
surveillance test intervals from 18 to 24 months for TS 3/4.3.2.1, 
``Safety Features Actuation System Instrumentation,'' and TS 3/
4.5.2, ``Emergency Core Cooling Systems - ECCS Subsystems - Tavg 
(greater than or equal to) 280F.'' Initiating conditions and 
assumptions remain as previously analyzed for accidents in the DBNPS 
Updated Safety Analysis Report.
    Results of the instrument drift study analysis and review of 
historical 18-month surveillance data and applicable maintenance 
records support an increase in the surveillance test intervals from 
18 to 24 months (and up to 30 months on a non-routine basis) 
because: the projected instrument errors caused by drift are bounded 
by the existing setpoint analysis or a new analysis has been 
performed incorporating a more conservative setpoint; and no 
potential for a significant increase in a failure rate of a system 
or component was identified during surveillance data and applicable 
maintenance records reviews.
    These proposed revisions are consistent with the NRC guidance on 
evaluating and proposing such revisions as provided in Generic 
Letter 91-04, ``Changes in Technical Specification Surveillance 
Intervals to Accommodate a 24-Month Fuel Cycle,'' dated April 2, 
1991.
    The proposed revisions to Allowable Values for Safety Features 
Actuation System (SFAS) Reactor Coolant System (RCS) Pressure - Low, 
RCS Pressure - Low-Low, RCS Pressure - Low-Low bypass permissive, 
and Decay Heat Isolation Valve and Pressurizer Heater Interlocks 
have no bearing on the probability of the initiation of an accident 
previously evaluated.
    The application of the Allowable Value to only the Channel 
Functional Test and not the Channel Calibration, the proposed 
deletion of the Trip Setpoints, the proposed revision of the TS 
3.3.2.1 Limiting Condition for Operation (LCO) and Action Statement 
3.3.2.1.a, and the proposed revisions to Actions 13 and 14 of TS 
Table 3.3-3, are associated with the proposed revision of the 
Allowable Values for SFAS RCS Pressure - Low, RCS Pressure - Low-
Low, and Decay Heat Isolation Valve and Pressurizer Heater 
Interlocks, and are consistent with NUREG-1430, Revision 1, 
``Standard Technical Specifications, Babcock and Wilcox Plants,'' 
dated April 1995. The proposed revisions have no bearing on the 
probability of the initiation of an accident previously evaluated.
    The proposed changes to TS Bases 3/4.3.1 and 3/4.3.2, ``Reactor 
Protection System and Safety System Instrumentation,'' and TS Bases 
3/4.5.2 and 3/4.5.3, ``ECCS Subsystems,'' are administrative changes 
associated with the other proposed changes, and do not affect 
previously analyzed accidents.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the slight increase in doses 
due to a letdown line break event as a result of the proposed change 
to the SFAS RCS Pressure - Low Allowable Value still satisfy the NRC 
Standard Review Plan Section 15.6.2 acceptance criteria that doses 
do not exceed a small fraction (10%) of the 10 CFR 100 guideline 
values. The remaining proposed changes to Allowable Values, and the 
other changes proposed by this License Amendment Request do not 
increase the radiological consequences of previously analyzed 
accidents because the source term, containment isolation, or 
radiological releases are not being changed by the proposed 
revisions.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated, for the reasons 
discussed below.
    No changes are being proposed to the type of testing currently 
being performed, only to the length of the surveillance test 
interval.
    Results of the instrument drift study analysis and review of 
historical 18-month surveillance data and maintenance records 
support an increase in the surveillance test intervals from 18 to 24 
months (and up to 30 months on a non-routine basis) because: the 
projected instrument errors caused by drift are bounded by the 
existing setpoint analysis or a new analysis has been performed 
incorporating a more conservative setpoint; and no potential for a 
significant increase in a failure rate of a system or component was 
identified during surveillance data and applicable maintenance 
records reviews.
    The proposed revisions to Allowable Values for SFAS RCS Pressure 
- Low, RCS Pressure - Low-Low, RCS Pressure Low-Low bypass 
permissive, and Decay Heat Isolation Valve and Pressurizer Heater 
Interlocks, do not alter the type of any testing currently being 
performed.
    The application of the Allowable Value to only the Channel 
Functional Test and not the Channel Calibration, the proposed 
deletion of the Trip Setpoints, revision of the TS 3.3.2.1 LCO and 
Action Statement 3.3.2.1.a, and the proposed revisions to Actions 13 
and 14 of

[[Page 30646]]

TS Table 3.3-3, are associated with the proposed revision to the 
Allowable Values for SFAS RCS Pressure - Low, RCS Pressure - Low-
Low, RCS Pressure Low-Low bypass permissive, and Decay Heat 
Isolation Valve and Pressurizer Heater Interlocks, and are 
consistent with NUREG-1430, Revision 1, ``Standard Technical 
Specifications, Babcock and Wilcox Plants,'' dated April 1995. The 
proposed revisions do not alter the type of testing currently being 
performed.
    The proposed changes to TS Bases 3/4.3.1 and 3/4.3.2, ``Reactor 
Protection System and Safety System Instrumentation,'' and TS Bases 
3/4.5.2 and 3/4.5.3, ``ECCS Subsystems,'' are administrative changes 
associated with the other proposed changes, and do not alter any 
testing currently being performed.
    3. Not involve a significant reduction in a margin of safety. 
The results of the instrument drift study analysis and review of 
historical 18-month surveillance data and applicable maintenance 
records support an increase in the surveillance test intervals from 
18 to 24 months (and up to 30 months on a non-routine basis) 
because: the projected instrument errors caused by drift are bounded 
by the existing setpoint analysis or a new analysis has been 
performed incorporating a more conservative setpoint; and no 
potential for a significant increase in a failure rate of a system 
or component was identified during surveillance data and applicable 
maintenance records reviews. Existing system and component 
redundancy is not affected by these proposed changes.
    There are no new or significant changes to the initial 
conditions contributing to accident severity or consequences, 
consequently there are no significant reductions in a margin of 
safety.
    The NRC staff has reviewed the licensees' analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.
    Attorney for licensees: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: April 18, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 3/4.7.6, ``Plant Systems - 
Control Room Emergency Ventilation System.'' Additional Limiting 
Conditions for Operation (LCO) would be added related to the 
availability of the station vent normal range radiation monitoring 
instrumentation. The associated TS bases would also be modified 
consistent with these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power Station 
(DBNPS), Unit No. 1, in accordance with this change would not:
    1a. Involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions, or assumptions are affected by the proposed changes.
    The proposed change to LCO 3.7.6.1 would include new required 
Action statements in the event that one or both channels of station 
vent normal range radiation monitoring instrumentation become 
inoperable. In the event that one channel is inoperable for greater 
than 7 days, or in the event that both channels are inoperable, the 
proposed Action statement would require that the control room normal 
ventilation system be isolated and at least one Control Room 
Emergency Ventilation System (CREVS) train be placed in operation.
    Under the proposed actions, the ventilation systems would be 
placed in a state equivalent to that which occurs were a high 
radiation isolation to occur. These proposed changes have no bearing 
on the probability of an accident.
    The proposed change to the terminology utilized in Surveillance 
Requirement (SR) 4.7.6.1.e is an administrative change made to make 
the terminology consistent with the proposed new Action statements. 
The proposed changes to Bases 3/4.7.6 are administrative changes 
consistent with the proposed changes to LCO 3.7.6.1. These changes 
have no bearing on the probability of an accident.
    1b. Involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
change the source term, containment isolation, or allowable 
releases.
    As described above, under the proposed new LCO 3.7.6.1 Actions, 
in the event that one station vent normal range radiation monitoring 
instrumentation channel is inoperable for greater than 7 days, or in 
the event that both channels are inoperable, the ventilation systems 
would be placed in a state equivalent to that which occurs were a 
high radiation isolation to occur. Therefore, in the unlikely event 
of an accident requiring control room isolation while in this 
condition, the dose consequences to control room operators would be 
unchanged.
    The proposed change to the terminology utilized in SR 4.7.6.1.e 
is an administrative change made to make the terminology consistent 
with the proposed new Action statements. The proposed changes to 
Bases 3/4.7.6 are administrative changes consistent with the 
proposed changes to LCO 3.7.6.1. These changes have no bearing on 
the consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because no new accident 
initiators or assumptions are introduced by the proposed changes.
    As described above, under the proposed new LCO 3.7.6.1 Actions, 
in the event that one station vent normal range radiation
    monitoring instrumentation channel is inoperable for greater 
than 7 days, or in the event that both channels are inoperable, 
theventilation systems would be placed in a state equivalent to that 
which occurs were a high radiation isolation to occur. Operation of 
the equipment and components in this manner would not introduce the 
possibility of any new or different kinds of accidents.
    The proposed change to the terminology utilized in SR 4.7.6.1.e 
is an administrative change made to make the terminology consistent 
with the proposed new Action statements. The proposed changes to 
Bases 3/4.7.6 are administrative changes consistent with the 
proposed changes to LCO 3.7.6.1. These changes would not introduce 
the possibility of any new or different kinds of accidents.
    3. Involve a significant reduction in a margin of safety because 
the proposed changes to the Action under LCO 3.7.6.1 ensure that 
control room isolation capability is maintained in the event a 
station vent radiation monitor is inoperable. The proposed allowable 
outage time of 7 days for one inoperable channel is consistent with 
the presently allowable outage time for one inoperable CREVS. The 
proposed Action to place at least one CREVS train in operation 
within 1 hour, in the event both channels of radiation monitoring 
become inoperable, is more conservative than the present Action 
which requires that a plant shutdown commence within 1 hour, but 
does not require the CREVS be placed in operation.
    The proposed change to the terminology utilized in SR 4.7.6.1.e 
is an administrative change made to make the terminology consistent 
with the proposed new Action statements. The proposed changes to 
Bases 3/4.7.6 are administrative changes consistent with the 
proposed changes to LCO 3.7.6.1. These changes would not affect the 
margin of safety.The NRC staff has reviewed the licensees' analysis 
and, based on this review, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 30647]]

    Attorney for licensees: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts, and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: August 22, 1996
    Description of amendment request: The proposed change would remove 
the action statement of Technical Specification (TS) Section 3.2.G, 
Table 3.2.6, Note 7, requiring reactor shutdown after 30 days of 
inoperability of the high range stack gas monitor and substitute an 
action statement consistent with the guidance provided in NRC Generic 
Letter 83-36.
    The high range stack monitor provides an estimate of gross stack 
activity that has exceeded the upper limit of the normal range 
instrumentation. The high range monitor reading serves as input to dose 
projection systems for initial estimation of off-site conditions. The 
monitor reading would be used prior to the acquisition of stack 
isotopic sample data which would provide a more accurate indication of 
stack activity.
    The licensee stated, among other things, that due to the 
passivefunction of the instruments and the ability to monitor this 
parameter utilizing alternate methods, it is not appropriate to impose 
stringentrequirements on the operation of the unit. This monitor is 
identified in the Vermont Yankee Regulatory Guide 1.97 submittal as 
Category 2, Type E. This monitor provides post-accident information for 
use in determining the magnitude of the release of radioactive 
materials and for monitoring such release. However, the high range 
stack monitor does not have any safety function associated with the 
prevention or automatic mitigation of design-basis accidents, neither 
does it provide primary information needed to permit the control room 
operating personnel to take required manually controlled actions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91 (a),the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below.
    [(1) The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.]
    The High Range Stack Monitor is a RG [Regulatory Guide] 1.97, 
Category 2, Type E instrument with no specified safety function 
associated with the prevention or automatic mitigation of design 
basis accidents, neither does it provide primary information needed 
to permit the control room operating personnel to take required 
manually controlled actions. The proposed change to the action 
statement associated with this monitor will not change the function 
of this monitor, and since the monitor is not assumed to initiate 
any accidents, nor function to mitigate any accidents, this change 
will not significantly increase the probability or consequences of 
any previously analyzed accident.
    [(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 does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will still ensure effective methods 
are available to assess post accident conditions. Thus, this 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 proposed change to the action statement associated with 
this monitor will not change the function of this monitor, and since 
the monitor is not assumed to function for the prevention or 
mitigation of any previously evaluated accidents, this change will 
not significantly reduce a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301
    Attorney for licensee: R. K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624
    NRC Project Director: Patrick D. Milano, Acting

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two 
Creeks, Manitowoc County, Wisconsin

    Date of amendment request: January 24, 1997, as supplemented on May 
15, 1997 (TSCR 193)
    Description of amendment request: The proposed amendments (Point 
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request 
(TSCR) 193) would revise TS 15.5.4, ``Fuel Storage,'' to increase fuel 
assembly enrichment limits to 5.0 w/o U-235 while maintaining 
Keff in the storage pools (spent fuel pool and new fuel 
storage racks) less than 0.95. The May 15, 1997, supplement provided a 
revised no significant hazards consideration determination that 
superseded the licensee's determination noticed in the Federal Register 
on April 23, 1997 (62 FR 19837).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a change to structures, 
systems, or components that would affect the probability or 
consequences of an accident previously evaluated in the PBNP Final 
Safety Analysis Report (FSAR). The only relevant concern with 
respect to increasing enrichment limits in the spent fuel pool and 
new fuel storage racks is one of criticality. The proposed changes 
use the same criticality limit used in the current Technical 
Specifications. Therefore, margin to safe operation of Units 1 and 2 
is maintained. The probability and consequences of an accident 
previously evaluated are dependent on this criticality limit. 
Because the limit will not change, the probability and consequences 
of those accidents previously evaluated will not change.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve a change to the physical 
structure of the spent fuel pool or of the plant. The proposed 
increase in spent fuel pool and new fuel storage racks fuel assembly 
enrichment limits maintains the margin to safe operation of Units 1 
and 2 because the criticality limit for the spent fuel pool and new 
fuel storage racks will not change. The enrichment increase does not 
affect any of the parameters or conditions that contribute to the 
initiation of any accidents. Because the criticality limit remains 
the same, these changes have no effect on plant operation or on the 
initiation of any accidents. Therefore, the proposed changes will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The proposed changes maintain the margin to safe operation of 
Units 1 and 2. The margin of safety is based on the criticality 
limit of the spent fuel pool and the new fuel storage racks. Because 
this limit will not change, the margin of safety will not be 
affected. Therefore, the proposed changes will not create a 
significant reduction in a margin of safety.

[[Page 30648]]

    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: April 23, 1997
    Description of amendment request: This request proposes to revise 
Technical Specification 3/4.9.4, Containment Building Penetrations, and 
its associated Bases section, to allow selected containment isolation 
valves to be opened under administrative controls during periods of 
core alterations or movement of irradiated fuel inside containment.
    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 involves changes to the Technical 
Specification requirements for containment closure which is an 
accident mitigating feature. The changes would not affect the 
likelihood of occurrence of any accidents previously evaluated. The 
proposed change does not involve any hardware or plant design 
changes. The containment leakage value is not assumed to be an 
initiator of any analyzed event. Containment isolation valves and 
temporary closure devices serve to limit the radiological 
consequences of accidents. The proposed change would ensure the 
service air and breathing air manual isolation valves will perform 
their required containment closure function and will serve to limit 
the consequences of a fuel handling accident as described in the 
USAR, such that the results of the analyses in the USAR remain 
bounding. In considering the consequences of a design basis fuel 
handling accident inside containment, the assumptions in the 
analysis take no credit for the containment as a barrier to prevent 
the postulated release of radioactivity. For events that could occur 
during CORE ALTERATIONS or movement of irradiated fuel assemblies, 
containment closure is considered a defense-in-depth boundary to 
prevent uncontrolled release of radioactivity. Additionally, the 
proposed change does not impose any new safety analyses limits or 
alter the plant's ability to detect and mitigate events. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change involves reliance on manual actuation of 
containment penetration valves (Service Air valves KA V-039 and KA 
V-118 and Breathing Air valves KB V-001 and KB V-002 are manual 
valves) to block the unimpeded flow of the containment atmosphere to 
the environment under certain conditions. The proposed change would 
not necessitate a physical alteration of the plant features that 
provide core cooling or subcriticality (no new or different type of 
equipment will be installed) or changes in parameters governing 
plant operation during CORE ALTERATIONS or movement of irradiated 
fuel in containment. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is similar to the use of administrative 
controls to isolate an open containment airlock door. The use of 
administrative controls in this manner has been approved by the NRC 
(WCGS Technical Specification Amendment 95) for plant operations 
that would not require the containment to maintain a pressure 
boundary. This scenario is applicable during plant shutdown for 
refueling when CORE ALTERATIONS and movement of irradiated fuel 
assemblies in the containment occur. Accidental damage to spent fuel 
during these operations is classified as a fuel handling accident. 
The proposed change has been developed considering the importance of 
the containment boundary in limiting the consequences of a design 
basis fuel handling accident. The proposed change allows for 
protection equivalent to that provided by previously approved 
methods of containment closure. Considering the probability of an 
event that would challenge the containment boundary, the alternative 
protection provided by this change, and the operational requirements 
to occasionally open these penetrations, the proposed change is 
acceptable and any reduction in the margin of safety is 
insignificant.
    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: William H. Bateman

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of application for amendment: January 24, 1997, as 
supplemented March 27, 1997
    Brief description of amendment: The proposed amendment will update 
the

[[Page 30649]]

Safety Limit Minimum Critical Power Ratio (SLMCPR) in Technical 
Specification 2.1.2 and the associated Bases section to reflect the 
results of the latest cycle-specific calculation performed for the 
Pilgrim Nuclear Power Station Operating Cycle 12. In addition, the 
values provided in Note 5 of Table 3.2.C.1, which are based on the 
SLMCPR values, have been revised as a result of the changes to the 
SLMCPR value.
    Date of issuance: April 7, 1997
    Effective date: April 7, 1997
    Amendment No.: 171
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6568) The March 27, 1997, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 7, 1997 No 
significant hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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

    Date of application for amendments:  August 19, 1996, as 
supplemented on February 5, March 13, April 29 and April 30, 1997.
    Brief description of amendments: The amendment would revise 
Technical Specification (TS) Section 4.4.5.2 to extend, for one 
additional operating cycle (i.e., Cycle 7), the 1.0 volt and 3.0 volt 
interim plugging criteria (IPC) which were added to the Braidwood, Unit 
1, TSs by License Amendment No. 69, issued on November 9, 1995. 
Additionally, this amendment to the Braidwood, Unit 1, license added 
some definitions and reporting requirements to TS Section 4.4.5.2 and 
modified the designations for the IPC models in TS Bases Section 3/
4.4.4.5. Braidwood, Unit 1, Cycle 7, will end in fall 1998. While there 
are no revisions to the TS for Braidwood, Unit 2, both units are being 
amended to maintain the continuity of the amendment numbers.
    Date of issuance: May 14, 1997.
    Date of effective: Immediately, to be implemented within 30 days.
    Amendment Nos.: 82
    Facility Operating License Nos. NPF-72 and NPF-77: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6570). The February 5, March 13, April 29 and April 30, 1997, 
submittals provided clarifying technical information that did not 
affect the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 14, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments:  June 20, 1996, as supplemented 
December 30, 1996, and March 5, 1997.
    Brief description of amendments: The amendments would change the 
TSs by incorporating an NRC-approved thermal limit licensing 
methodology in the list of approved methodologies used in establishing 
the fuel cycle-specific thermal limits. In addition, the proposed 
amendments would change the TSs to reflect the use of Siemens Power 
Corporation (SPC) ATRIUM-9B fuel for the first time at Dresden, Units 2 
or 3. The proposed amendments would also correct minor editorial items 
in the TSs.
    In March 1997, the NRC staff performed an audit of the application 
of Advanced Nuclear Fuel for Boiling Water Reactors (ANFB) to ATRIUM-9 
fuel. The staff raised concerns associated with the ATRIUM-9B fuel 
additive constant uncertainty used as input to the NRC-approved 
methodology for the calculation of minimum critical power ratio (MCPR). 
In response to the audit findings, by letter dated April 18, 1997, SPC 
submitted a generic topical report (ANF-1125(P) Supplement 1 Appendix 
D), which is currently under staff review, for the future reload 
analysis in the safety limit MCPR calculation. The staff schedule for 
the review of the topical report will not be timely enough for the 
resolution of the ATRIUM-9B MCPR issue to support reload and restart of 
Dresden, Unit 3. Therefore, by letters dated May 2 and May 6, 1997, 
ComEd provided additional information concerning the MCPR issues and 
how it will affect the Dresden, Unit 3, D3R15 fuel cycle and provided 
additional information concerning the ATRIUM-9B fuel design and 
shutdown margin that are applicable during refueling and shutdown.
    The staff is currently reviewing the licensee's May 2 and May 6, 
1997, letters. To be more timely and support the reload schedule for 
Dresden, Unit 3 (currently scheduled for May 20, 1997), the staff has 
chosen to split its consideration of the proposed amendments into two 
parts. The first part of the amendment package now being evaluated 
would modify Section 5.3.A, ``Design Features'' of the TSs to reflect 
use of the ATRIUM-9B fuel design and would include two SPC topical 
reports in TS Section 6.9.A.6, ``Core Operating Limits Report,'' to 
reflect mechanical design criteria for this fuel. This change would 
allow this fuel to be loaded into the core only under Operational Modes 
3 (Hot Shutdown), 4 (Cold Shutdown), and 5 (Refueling) and does not 
permit startup or power operation using the ATRIUM-9B fuel.
    Date of issuance: May 16, 1997
    Date of effective: Immediately, to be implemented within 30 days.
    Amendment Nos.: 159 and 154
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17227). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 16, 1997 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket No. 50-265, Quad Cities Nuclear 
Power Station, Unit 2, Rock Island County, Illinois

    Date of application for amendment: April 21, 1997
    Brief description of amendment: The amendment increases the minimum 
critical power ratio safety limit for Unit 2 and adds a Siemens Power 
Corporation reference to the Technical Specifications (TS) to allow 
plant operation in Operational Modes 1 and 2.
    Date of issuance: May 22, 1997
    Date of effective: Immediately, to be implemented within 30 days.
    Amendment No.: 174
    Facility Operating License No. DPR-30: The amendment revised the 
TSs. Public comments requested as to proposed no significant hazards 
consideration: Yes (62 FR 23499 dated April 30, 1997). This notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant

[[Page 30650]]

hazards consideration determination. No comments have been received. 
The notice also provided for an opportunity to request a hearing by May 
30, 1997, but indicated that if the Commission makes a final no 
significant hazards consideration determination any such hearing would 
take place after issuance of the amendment. The Commission's related 
evaluation of the amendment, finding of exigent circumstances, and 
final no significant hazards consideration determination are contained 
in a Safety Evaluation dated May 22, 1997.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan Date of application for amendment: December 2, 1996 (NRC-
96-0134)

    Brief description of amendment: The amendment revises TS 3.1.4.3, 
TS Table 3.3.6-1, and TS Table 4.3.6-1 to change the operability 
requirements for the Rod Block Monitor (RBM). Specifically, the 
revision requires the RBM to be operable when reactor thermal power is 
greater than or equal to 30 percent of rated thermal power.
    Date of issuance: May 15, 1997
    Date of effective: May 15, 1997, with full implementation within 60 
days
    Amendment No.: 112
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
124) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 15, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

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: April 29, 1997
    Brief description of amendments: The amendments incorporate a 
license condition that will allow revisions to the Oconee Updated Final 
Safety Analysis Report (UFSAR) that clarifies the main turbine-
generated missile protection criteria.
    Date of issuance: May 16, 1997
    Date of effective: As of the date of issuance and implementation is 
the incorporation in the UFSAR the changes described in Duke Power 
Company's application dated April 29, 1997
    Amendment Nos.:  224, 224, and 221
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the UFSAR and added a new License Condition. Public 
comments requested as to proposed no significant hazards consideration: 
Yes. (62 FR 24512 dated May 5, 1997). The notice provided an 
opportunity to submit comments on the Commission's proposed no 
significant hazards consideration determination. No comments have been 
received as of the date of issuance. The notice also provided for an 
opportunity to request a hearing by June 9, 1997, but indicated that if 
the Commission makes a final no significant hazards consideration 
determination, any such hearing would take place after issuance of the 
amendments.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated May 
16, 1997.
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of application for amendment: November 26, 1996, as 
supplemented February 12, 1997.
    Brief description of amendment: The amendment changes the allowable 
primary-to-secondary leak rate and in the Surveillance Requirements 
section of the TSs it changes the acceptance criteria for steam 
generator tubes. The amendment changes the reference that is included 
in the tube acceptance criteria from Combustion Engineering topical 
report CEN-601-P Revision 01-P to CEN-630-P, Revision 01.
    Date of issuance: May 20, 1997
    Date of effective: May 20, 1997, to be implemented within 30 days.
    Amendment No.: 184
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64376) The February 12, 1997, submittal provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated May 20, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of application for amendment: April 4, 1995, as supplemented 
by letters dated August 25, 1995, and April 18, 1997.
    Brief description of amendment: The amendment changes the required 
frequency for inspecting reactor coolant pump flywheels.
    Date of issuance: May 20, 1997
    Date of effective: May 20, 1997, to be implemented within 30 days.
    Amendment No.: 185
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995, (60 FR 
35069) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 20, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear 
One,Unit No. 2, Pope County, Arkansas

    Date of application for amendment: October 7, 1996, as supplemented 
February 10, and May 8, 1997
    Brief description of amendment: The amendment changes the channel 
functional testing frequency for most of the Reactor Protection System 
(RPS) and Engineered Safety Feature Actuation System (ESFAS) 
instrumentation from monthly to every four months. In addition, the 
amendment allows the use of Cycle Independent Shape Annealing Matrix 
(CISAM) methodology in the Core Protection Calculators (CPCs). Finally, 
the amendment makes a number of administrative changes to the Technical 
Specifications (TS) to clarify the existing TS or correct previous 
errors in the TS.
    Date of issuance: May 21, 1997
    Date of effective: May 21, 1997
    Amendment No.: 186
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4346) The Commission's related evaluation of the amendment is contained 
in a Safety

[[Page 30651]]

Evaluation dated May 21, 1997 No significant hazards consideration 
comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

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

    Date of amendment request: March 27, 1997
    Brief description of amendment: The amendment changes TSs 
surveillance requirements 4.5.2.d.3 and 4.5.2.d.4 by increasing the 
required amount of trisodium phosphate dodecahydrate (TSP) stored in 
the containment sump from 97.5 cubic feet to 380 cubic feet, and 
adjusts the TSP sampling requirement accordingly.
    Date of issuance: May 15, 1997
    Date of effective: May 15, 1997, to be implemented within 60 days.
    Amendment No.: 127
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17234) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 15, 1997 No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of amendment request: August 21, 1996, as supplemented by 
letter dated March 17, 1997
    Brief description of amendment: The amendment approves revision of 
Attachment 1 to the operating license concerning design and testing 
modifications in the Containment Vacuum Relief System (CVR) that 
penetrate the primary containment at Waterford Steam Electric Station, 
Unit 3. The penetrations affected are commonly referred to as 
Penetrations 53 and 65.
    Date of issuance: May 20, 1997
    Date of effective: May 20, 1997, to be implemented within 90 days.
    Amendment No.: 128
    Facility Operating License No. NPF-38: Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57484) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 20, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of application for amendment: September 13, 1996, as 
supplemented by letter dated January 15, 1997.
    Brief description of amendment: The amendment revised the Technical 
Specifications to permit the use of 10 CFR Part 50, Appendix J, Option 
B, performance-based containment leakage rate testing.
    Date of issuance: May 19, 1997
    Date of effective: May 19, 1997, to be implemented within 60 days 
of the date of issuance.
    Amendment No.: 158
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57487) The January 15, 1997, supplemental letter provided additional 
clarifying information and did not change the initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
May 19, 1997. No significant hazards consideration comments received: 
Yes. Comments were submitted by Patrick J. Dostie on behalf of the 
State of Maine by letter dated April 15, 1997. The staff responded by 
letter dated May 19, 1997.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

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

    Date of amendment request: February 18, 1997, as supplemented by 
letter dated February 26, 1997.
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications relating to the reactor core fuel 
assembly design features requirements contained in Technical 
Specification 5.3.1, Fuel Assemblies. The changes made by this 
amendment allow for the limited replacement of failed or damaged fuel 
rods in fuel assemblies with solid stainless steel or zirconium alloy 
filler rods.
    Date of issuance: May 13, 1997
    Date of effective: May 13, 1997
    Amendment No.: 51
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11496) The licensee's letter dated February 26, 1997, provided a 
correction to a typographical error in the original application but 
does not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated May 13, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833

Portland General Electric Company, et al., Docket No. 50-344, 
Trojan Nuclear Plant, Columbia County, Oregon

    Date of application for amendment: November 2, 1995.
    Brief description of amendment: This amendment changes the TS to 
reflect changes in the organization as they apply to oversite and 
management of the Trojan Nuclear Plant.
    Date of issuance: October 31, 1996
    Date of effective: October 31, 1996
    Amendment No.: 195
    Facility Operating License No. NPF-1: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58404) No significant hazards consideration comments received: No.
    Local Public Document Room location: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207

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

    Date of application for amendments: January 7, 1997
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 3/4.2.5 to incorporate an exception to the 
provisions of TS 4.0.4 and to clarify the time at which the 
surveillance can be performed by adding that the surveillance is to be 
performed within 24 hours after attaining steady state conditions at or 
above 90% rated thermal power. The revised surveillance contains 
editorial enhancements that clarify the

[[Page 30652]]

surveillance requirement. Salem Unit 1 TS Table 3.2-1 is also being 
revised to delete reference to three loop operation.
    Date of issuance: May 8, 1997
    Date of effective: Both units, as of date of issuance, to be 
implemented prior to entry into Mode 1 from the current outage. 
Amendment Nos. 193 and 176
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4353) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 8, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph 
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama

    Date of amendment request: March 25, 1997
    Brief description of amendments: The amendment changes Technical 
Specification 3/4.4.9, ``Specific Activity,'' and the associated Bases 
to reduce the limit associated with dose equivalent iodine-131. The 
steady-state dose equivalent iodine-131 limit would be reduced by 40 
percent from 0.5 [micro]Ci/gram to 0.3 [micro]Ci/gram and the maximum 
instantaneous value would be reduced by 40 percent from 30 [micro]Ci/
gram to 18 [micro]Ci/gram.
    Date of issuance: May 19, 1997
    Date of effective: As of the date of issuance to be implemented 
within 30 days
    Amendment Nos.: 128
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: April 4, 1997 (62 FR 
16201) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 19, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Dated at Rockville, Maryland, this 28th day of May, 1997.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 97-14395 Filed 6-3-97; 8:45 am]
BILLING CODE 7590-01-F