[Federal Register Volume 62, Number 146 (Wednesday, July 30, 1997)]
[Notices]
[Pages 40843-40868]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-11910]


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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.

[[Page 40844]]

    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 3, 1997, through July 18, 1997. The 
last biweekly notice was published on July 16, 1997.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By August 29, 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 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

[[Page 40845]]

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. 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.

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

    Date of amendments request: May 23, 1997
    Description of amendments request: The proposed amendment would 
revise Technical Specification 3/4.4.4 to allow the installation of 
ABB/CE welded sleeves, in accordance with ABB/CE Topical Report CEN-
630-P, ``Repair of 3/4 Inch Outer Diameter Steam Generator Tubes Using 
Leak Tight Sleeves,'' Revision 1, in the Palo Verde Units 1, 2 and 3 
steam generators.
    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 amendment to permit the use of steam generator tube 
sleeves as an alternative to tube plugging is a safe and effective 
repair procedure that does not result in removing a tube from 
service. Mechanical strength, corrosion resistance, installation 
methods, and inservice inspection techniques of sleeves have been 
shown to meet NRC acceptance criteria.
    Analytical verifications were performed using design and 
operating transient parameters selected to envelope loads imposed 
during normal operating and accident conditions. Fatigue and stress 
analysis of sleeved tube assemblies were completed in accordance 
with the requirements of Section III of the ASME Code. The results 
of qualification testing, analysis and plant operating experience at 
other facilities demonstrates that the sleeving process is an 
acceptable means of maintaining steam generator tube integrity. The 
sleeve configuration has been designed and analyzed in accordance 
with the structural margins specified in Regulatory Guide 1.121 (RG 
1.121). Furthermore, the installed sleeve will be monitored through 
periodic inspections on a sample basis with eddy current techniques. 
A sleeve-specific plugging margin, per the recommendations of 
Regulatory Guide 1.121, has been specified with appropriate 
allowances for NDE uncertainty and defect growth rate. Therefore, 
since the sleeve provides the same protection against a tube rupture 
as the original tube, the use of sleeves does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    Recently, industry experience with forced shutdown events 
associated with tube failures at sleeve junctions was assessed by 
APS and ABB-CE. The root cause of these events has been attributed 
to the lack of proper post-installation stress relief and/or the 
imposition of high stresses due to tube growth restrictions at 
locked tube supports. The material and design of the PVNGS steam 
generator supports minimizes the potential for locked supports. The 
tube supports are of eggcrate design and are constructed of ferric 
stainless steel. The large flow area in the eggcrate design provides 
better irrigation and reduces the potential for steam blanketing, 
therefore, the tube-to-tube support crevices are less likely to be 
blocked by crud, boiler water deposits and corrosion products. Since 
the support material is type 409 ferric stainless steel, it is not 
susceptible to magnetite corrosion which has resulted in denting and 
lockup at plants with carbon steel supports. These conclusions have 
been substantiated via tube pull activities conducted in PVNGS Unit 
2. Although ABB/CE does not require post-weld heat treatment in all 
applications, APS will require that a post-weld stress relief be 
conducted for sleeve installations. Therefore, with proper sleeve 
installation the proposed change will not involve a significant 
increase in the probability of an accident previously evaluated.
    The consequences of accidents previously analyzed are not 
increased as a result of sleeving activities. The hypothetical 
failure of the sleeve would be bounded by the current steam 
generator tube rupture analysis contained in the PVNGS UFSAR. Due to 
the slight reduction in diameter caused by the sleeve wall 
thickness, it is expected that the primary release rates would be 
less than assumed for the steam generator tube rupture analysis, 
and, therefore, would result in lower primary fluid mass release to 
the secondary system. Additionally, further conservatism is 
introduced if the break were postulated to occur at a location on 
the tube higher than the location where a sleeve is installed. The 
overall effect would be reduced steam generator tube rupture release 
rates. The minimal reduction in flow area associated with a tube 
sleeve has no significant affect on steam generator performance with 
respect to heat transfer or system flow resistance and pressure 
drop. The installation of sleeves rather than plugging also 
maintains a greater heat transfer surface in the steam generator. In 
any case, the impacts are bounded by evaluations which demonstrate 
the acceptability of tube plugging, which totally removes the tube 
from service.
    Therefore, in comparison to plugging, tube sleeving is 
considered a significant improvement with respect to steam generator 
performance. Therefore, based on the above, the proposed amendment 
does not significantly increase the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    A sleeved steam generator tube performs the same function in the 
same passive manner as an unsleeved steam generator tube. Tube 
sleeves are designed and qualified to the stress and pressure limits 
of Section III of the ASME Code and Regulatory Guide 1.121.
    The installation of the sleeve, including weld and welder 
qualification and nondestructive examination (NDE), meets or exceeds 
the requirements of ASME Section XI. Three types of NDE are 
conducted. Ultrasonic Testing (UT) is performed to verify the 
adequacy of the tube to sleeve weld assuring proper fusion. Eddy 
Current testing (ECT) is performed following each installation to 
establish baseline data for each sleeve in order to monitor future 
degradation of the primary to secondary pressure boundary. Visual 
inspections will be performed to verify or ascertain the mechanical 
and structural condition of a weld. Critical conditions which are 
checked include weld width and completeness, and the absence of 
visibly noticeable indications such as cracks, pits, and burn 
through.
    ABB Combustion Engineering, Inc., Report CEN-630-P, Revision 01, 
``Repair of 3/4'' O.D. Steam Generator Tubes Using Leak Tight 
Sleeves'' dated November, 1996, demonstrates that the repair of 
degraded steam generator tubes using tube sleeves will result in 
tube bundle integrity consistent with the original design basis. 
Extensive analyses and testing have been performed on the sleeve and 
sleeve to tube joints to demonstrate that the design criteria are 
met. The proposed amendments have no significant effect on the 
configuration of the plant, and the change does not affect the way 
in which the plant is operated. Therefore, reactor operation with 
sleeves installed in the steam generator tubes 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.

[[Page 40846]]

    Evaluation of the sleeved tubes indicates no detrimental effects 
on the sleeve-tube assembly resulting from reactor coolant system 
flow, coolant chemistries, or thermal and pressure conditions. 
Structural analyses have been performed for sleeves which span the 
tube at the top of the tube sheet and which span the flow 
distribution plate or eggcrate support. Mechanical testing has been 
performed to support the analyses. Corrosion testing of typical 
sleeve-tube assemblies has been completed and reveals no evidence of 
sleeve or tube corrosion considered detrimental under anticipated 
service conditions.
    Steam generator tube integrity is maintained under the same 
limits for sleeved tubes as for unsleeved tubes, ie., Section III of 
the ASME Code and Regulatory Guide 1.121. The portions of the 
installed sleeve assembly which represents the reactor coolant 
pressure boundary can be monitored for the initiation and 
progression of sleeve/tube wall degradation, thus satisfying the 
requirements of Regulatory Guide 1.83. The degradation limit at 
which a sleeve/tube boundary is considered inoperable has been 
analyzed in accordance with Regulatory Guide 1.121 and is specified 
in the proposed amendment. Eddy current detectability of flaws has 
been verified by ABB Combustion Engineering. Additionally, the 
Technical Specifications continue to require monitoring and 
restriction of primary- to- secondary system leakage through the 
steam generators. The minimal reduction in RCS flow due to sleeving 
results in an insignificant impact on RCS operation during normal or 
accident conditions and is bounded by tube plugging evaluations.
    Based upon the testing and analyses performed, the installation 
of tube sleeves will not result in a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

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

    Date of amendments request: 3April 30, 1997
    Description of amendments request: The proposed amendments would 
revise Surveillance Requirements (SRs) 4.7.2.b.2 and 4.7.2.c in the 
Technical Specifications for the Brunswick Steam Electric Plant, Units 
1 and 2. These SRs require periodic testing of the control room 
emergency ventilation system charcoal filters. The proposed amendments 
would revise the temperature and relative humidity conditions under 
which the testing is performed. The revised conditions were selected to 
approximate operating or accident conditions. Testing at the revised 
conditions is more conservative than testing at the currently required 
conditions. Additionally, the proposed amendments would relax the 
acceptance criterion for filtration efficiency from 95% to a value 
corresponding to a filtration efficiency of 90%. The 90% value is the 
filtration efficiency assumed in the current bounding calculations for 
control room dose under accident conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendments revise Surveillance Requirements 
4.7.2.b.2 and 4.7.2.c to require testing of the control room
    emergency ventilation system (CREVS) charcoal in accordance with 
ASTM D3803-1989, ``Standard Test Method for Nuclear-Grade Activated 
Carbon.'' Currently, Surveillance Requirements 4.7.2.b.2 and 4.7.2.c 
to [sic] require testing in accordance with the criteria of 
Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, 
1976. The purpose of the CREVS is to mitigate an accident. It is not 
associated with any initiating events and, therefore, cannot affect 
the probability of any accident.
    ASTM D3803-1989 is an industry accepted standard for charcoal 
filter testing. The conditions employed by this standard were 
selected to approximate operating or accident conditions of a 
nuclear reactor which would severely reduce the performance of 
activated carbons. The ASTM D3803-1989 testing is more stringent 
than that required by the criteria of Regulatory Position C.6.a of 
Regulatory Guide 1.52, Revision 1, 1976. Specifically, the testing 
temperature of ASTM D3803-1989 is 30.0 [plus or minus] 0.2 deg.C 
versus 80 deg.C for the Regulatory Guide 1.52 testing. Also, ASTM 
D3803-1989 requires a relative humidity of 93 to 96% versus [greater 
than or equal to] 70% for the Regulatory Guide 1.52 testing. Both 
these parameters result in the ASTM D3803-1989 test being a more 
conservative test [than] that required by the criteria of Regulatory 
Position C.6.a of Regulatory Guide 1.52, Revision 1, 1976.
    The proposed changes to Surveillance Requirements 4.7.2.b.2 and 
4.7.2.c require that charcoal samples tested in accordance with the 
methodology of ASTM D3803-1989 meet the acceptance criteria of < 
5.0% penetration of methyl iodide. This corresponds to a 90% 
filtration efficiency which is the filtration efficiency assumed in 
the current bounding calculations of control room doses. As such, 
the proposed acceptance criteria of < 5.0% penetration of methyl 
iodide ensures that General Design Criterion 19 dose limits for 
control room operators are not exceeded.
    Therefore, the proposed amendments do not involve an increase in 
the consequences of an accident.
    2. The proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    As stated above, the proposed amendments revise the required 
testing methodology for the CREVS charcoal. The CREVS is not 
associated with any initiating events. The system design is not 
affected by the proposed change. Therefore, the proposed amendments 
cannot create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed amendments upgrade the CREVS charcoal testing 
requirements from the criteria of Regulatory Position C.6.a of 
Regulatory Guide 1.52, Revision 1, 1976 to ASTM D3803-1989. The 
conditions employed by ASTM D3803-1989 were selected to approximate 
operating or accident conditions of a nuclear reactor which would 
severely reduce the performance of activated carbons. The ASTM 
D3803-1989 testing is more stringent than that required by the 
criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, 
Revision 1, 1976. The testing temperature of ASTM D3803-1989 [is] 
lower than that of Regulatory Guide 1.52 and the relative humidity 
required by ASTM D3803-1989 is higher than that required by 
Regulatory Guide 1.52. This makes the ASTM D3803-1989 test being 
[sic] a more conservative test [than] that required by the criteria 
of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, 
1976. Additionally, the proposed acceptance criteria of < 5.0% 
penetration of methyl iodide ensures that General Design Criterion 
19 dose limits for control room operators are not exceeded. As such, 
the proposed license amendments 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road,

[[Page 40847]]

Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Gordon E. Edison, Acting

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

    Date of amendments request: May 23, 1997
    Description of amendments request: The proposed amendments to 
Technical Specification 3/4.4.5 for the Brunswick Steam Electric Plant, 
Units 1 and 2, reduce the short-term limit for Dose Equivalent I-131 
activity in the reactor coolant from 4.0 microcuries/gram to 3.0 
microcuries/gram. With coolant specific activity greater than 0.2 
microcuries/gram Dose Equivalent I-131 but less than or equal to the 
short-term limit, operation of the affected unit may continue for up to 
48 hours provided that operation under these conditions does not exceed 
10 percent of the unit's total yearly operating time. With coolant 
specific activity greater than 0.2 microcuries/gram I-131 Dose 
Equivalent for more than 48 hours during one continuous time interval 
or greater than the short-term limit, the affected unit must be placed 
in Hot Shutdown within 12 hours. The purpose of the reduction of the 
short-term limit is to ensure control room operator dose following a 
Main Steam Line Break event is within the guidelines contained in 10 
CFR Part 100 and the limits contained in Criterion 19 of Appendix A to 
10 CFR Part 50.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendments conservatively revise Action Statements 
a.1 and a.2 of Technical Specification 3/4.4.5 by reducing the 
maximum allowed reactor coolant specific activity from 4.0 to 3.0 
[microcuries]/gram dose equivalent I-131. The purpose of the maximum 
allowable iodine specific activity is to ensure that the thyroid 
dose from a main steam line break (MSLB )is within the 10 CFR 100 
dose guidelines and the General Design Criteria 19 dose limits for 
control room operators. The maximum allowable iodine specific 
activity is not associated with any initiating event and, therefore, 
cannot affect the probability of any accident. The proposed 
amendments result in a more conservative action limit and, 
therefore, do not increase the consequences of any accident.
    2. The proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendments conservatively reduce the maximum 
allowable reactor coolant iodine specific activity. The activity 
limit is not associated with any initiating event and the system 
design is not affected. Therefore, the proposed amendments cannot 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed amendments revise Action Statements a.1 and a.2 of 
Technical Specification 3/4.4.5 by reducing the maximum allowed 
reactor coolant specific activity from 4.0 to 3.0 [microcuries]/gram 
dose equivalent I-131. As stated above, the purpose of the maximum 
allowable iodine specific activity is to ensure that the thyroid 
dose from a MSLB is within the 10 CFR 100 dose guidelines and the 
General Design Criteria 19 dose limits for control room operators. 
The reduction in the activity limit is a conservative change and, 
therefore, the proposed license amendments do 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Gordon E. Edison, Acting

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

    Date of amendment request: June 12, 1997
    Description of amendment request: The amendment would make changes 
to the operations organization description.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    This change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment deals with changing position titles and 
clarification of the Harris Nuclear Plant (HNP) Operations 
management organization and responsibilities. The changes are 
considered to be admnistrative in nature and do not involve any 
modifications to any plant equipment or [affect] plant operation.
    Therefore, there would be no increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment deals with changing position titles and 
clarification of the HNP Operations management organization and
    responsibilities. The changes are considered to be 
administrative in nature and do not involve any modifications to any 
plant equipment or [affect] plant operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed amendment does not reduce the margin of safety as 
defined in the Safety Analysis Report or the bases contained in the 
Technical Specifications. The requirement to have a licensed SRO 
[Senior Reactor Operator] management position responsible for plant 
operations is maintained within the proposed amendment.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the 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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Gordon E. Edison, Acting

[[Page 40848]]

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

    Date of amendment request: May 27, 1997
    Description of amendment request: The proposed amendments would 
revise Technical Specification Section 6, ``Administrative Controls,'' 
to incorporate revised organizational titles and would modify License 
Condition 2.C.(30)(a) to reflect that the Shift Technical Advisor 
function may be filled by someone other than a designated Senior 
Reactor Operator (SRO). In addition, the proposed amendments would 
change the submittal frequency of the Radiological Effluent Release 
Report from semiannually to annually. The proposed amendments will also 
make several administrative and editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect any accident initiators or 
precursors and do not change or alter the design assumptions for 
systems or components used to mitigate the consequences of an 
accident. The proposed changes do not affect the design or operation 
of any system, structure, or component in the plant. There are no 
changes to parameters governing plant operation, and, no new or 
different type of equipment will be installed.
    The proposed changes provide clarification, consistency with 
station procedures, programs, the Code of Federal Regulations 
(10CFR), other Technical Specifications, and Improved Technical 
Specifications. These changes do not impact any accident previously 
evaluated in the UFSAR [Updated Final Safety Analysis Report]. There 
is no relaxation of applicable administrative controls. Those 
administrative requirements which have no effect on safe operation 
of the plant are eliminated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not affect the design or operation of 
any plant system, structure, or component. There are no changes to 
parameters governing plant operation, and, no new or different type 
of equipment will be installed. The organizational and 
administrative changes proposed have no effect on the design or 
operation of any system, structure, or component in the plant. There 
are no changes to parameters governing plant operation; no new or 
different type of equipment will be installed.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the margin of safety for any 
Technical Specification. The initial conditions and methodologies 
used in the accident analyses remain unchanged; therefore, accident 
analyses results are not impacted. Plant safety parameters or 
setpoints are not affected. All responsibilities described in the 
Technical Specifications for administrative controls will continue 
to be performed by individuals possessing the requisite 
qualifications. Clarifications, relocations, and nomenclature 
changes neither result in a reduction of personnel responsibilities, 
nor do they cause a relaxation of programmatic controls. There are 
no resulting effects on plant safety parameters or setpoints.
    Guidance has been provided in ``Final Procedures and Standards 
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744, 
for the application of standards to license change requests for 
determination of the existence of significant hazards 
considerations. This document provides examples of amendments which 
are and are not considered likely to involve significant hazards 
considerations. These proposed amendments most closely fit the 
example of a purely administrative change to the Technical 
Specifications to achieve consistency throughout the Technical 
Specifications, correction of an error, or a change in nomenclature.
    The proposed amendment does not involve a significant relaxation 
of the criteria used to establish safety limits, a significant 
relaxation of the bases for the limiting safety system settings, or 
a significant relaxation of the bases for the limiting conditions 
for operations. The proposed change does not reduce the margin of 
safety as defined in the basis for any Technical Specification.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: July 1, 1997
    Description of amendment request: The proposed amendments would 
change the definition of Channel Calibration in section 1.4 of the 
Technical Specifications to require an inplace qualitative assessment 
of thermocouple and resistance temperature detectors which cannot be 
calibrated. The proposed amendments will also correct typographical and 
miscellaneous errors in TS Table 3.3.2-1, Table 3.3.6-1, and Bases 
section 3/4.3.1.
    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 because:
    a. The change in the definition of a Channel Calibration is to 
make the wording more clear and to require an inplace qualitative 
assessment in place of the calibration of thermocouple and 
resistance temperature detector (RTD) sensors. The thermocouple and 
RTD sensors are not adjustable and are not subject to drift due to 
their design. The inplace qualitative assessments will assure proper 
functioning of the sensors, due to the nature of these sensors and 
the associated failure modes, and thus will verify that the sensors 
will be able to fulfill their intended function(s). Therefore the 
change to the definition will not change the probability or 
consequences of an accident previously evaluated.
    b. Manual initiation of isolation actuation instrumentation trip 
systems for inboard and outboard valves is required to be operable 
per TS Table 3.3.2-1, Trip Functions B.1 and B.2, respectively. Trip 
Function B.2, outboard valves, lists valve group 7, TIP system 
isolation valves. Valve group 7 consists of an automatic inboard 
isolation valve for each TIP guide tube penetrating the primary 
containment (correctly listed under B.1), and a manual outboard 
isolation valve on each guide tube, that is an explosive squib 
valve. Each explosive squib valve is manually actuated with a 
keylock switch from the main control room per design. Each is a 
positive control backup upon failure of an inboard valve in the open 
position. The squib valves are not actuated from isolation actuation 
channel logic. This configuration meets the current design and 
licensing basis. Therefore, deletion of valve group 7 from TS Table 
3.3.2-1 will not change the probability or consequences of an 
accident previously evaluated.
    c. The proposed change to TS Table 3.3.6-1, Control Rod 
Withdrawal Block Instrumentation, deletes Note (e) from Trip 
Function 4.a, IRM detector-not-full-in rod block. This rod 
withdrawal block functions during Operational Condition 2, Startup, 
and 5, Refuel, to assure that IRMs are operable during control rod 
withdrawal in these plant Operational Conditions. The rod block is 
not bypassed when the IRMs are on range 1. Thus Note (e) does not 
apply to this trip function and is being deleted. Therefore, the 
correction of this error will not change the probability or 
consequences of an accident previously evaluated.

[[Page 40849]]

    d. The change to TS Bases 3/4.3.1 to correct a typographical 
error referencing TS Table 3.3.1-2, Note , instead of Note 
 is an administrative change and thus will not 
change the probability or consequences of an accident.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The changes to the definition of Channel Calibration and 
correction of the other miscellaneous errors in the TS and TS Bases 
will not create the possibility of a new or different kind of 
accident, because the changes will not affect the design or 
operation of any structure, system, or component in the plant.
    3) Involve a significant reduction in the margin of safety 
because:
    a. The definition of Channel Calibration is being changed to be 
like the definition in NUREG 1434, Standard Technical Specifications 
General Electric Plants, BWR/6, Revision 1. The primary changes 
involve requiring only an inplace qualitative assessment of 
thermocouple and RTD sensors. These sensors are not adjustable and 
not susceptible to setpoint drift. Thus the appropriate check of the 
sensors is a qualitative assessment only. The inplace qualitative 
assessment assures operability of the sensors. Therefore there is no 
reduction in the margin of safety.
    b. The remaining miscellaneous changes are corrections due to 
errors in the TS. The corrections will make the associated TS 
consistent with the design and licensing basis of LaSalle or correct 
typographical errors. Therefore, there is no reduction in the margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: July 17, 1996, as supplemented by 
letters dated June 3, and July 7, 1997.
    Description of amendment request: The proposed change request 
modifies Waterford Steam Electric Station, Unit 3, Technical 
Specifications (TSs) 3/4.7.1.3, ``CONDENSATE STORAGE POOL,'' by 
increasing the minimum Condensate Storage Pool (CSP) level from 82 
percent to 91 percent in Modes 1, 2, and 3. The July 7, 1997, 
supplement proposes to expand the applicability of TS 3.7.1.3 to 
include Mode 4 operational requirements and maintains the 91 percent 
minimum CSP level previously requested for Modes 1, 2, and 3. The staff 
previously issued No Significant Hazard Considerations notice on March 
26, 1997 (62 FR 14461).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?Response: No.
    Increasing the minimum required Condensate Storage Pool (CSP) 
level to 91 percent will insure that the minimum required 170,000 
gallons of water is available to supply the Emergency Feedwater 
System and that 3,500 gallons of water is available for use by the 
Component Cooling Water Makeup System in Modes 1, 2, and 3. 
Maintaining a minimum required CSP level of 11 percent will insure 
that 3,500 gallons of water is available for use by the Component 
Cooling Water Makeup System in Mode 4. Maintaining the minimum 
required water volume will not increase the probability of any 
accident previously evaluated. Additionally, it will not affect the 
consequences of any accident. Maintaining a minimum required CSP 
level will ensure that the system remains within the bounds of the 
accident analysis. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    Increasing the minimum water volume of the CSP from 82 percent 
to 91 percent in Modes 1, 2, and 3 does not create a possibility for 
a new or different kind of accident. Maintaining a minimum water 
volume of the CSP at 11 percent in Mode 4 does not create a 
possibility for a new or different kind of accident. The CSP will be 
operated in the same manner as previously evaluated. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    Operation in accordance with this proposed change will ensure 
that the minimum contained water volume of the CSP will remain 
adequate under all conditions. This will improve the present margin 
of safety. Therefore, the proposed change will not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: James W. Clifford, Acting Director

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

    Date of amendment request: May 29, 1997
    Description of amendment request: The proposed amendments will 
improve consistency throughout the Technical Specifications and their 
related Bases by removing outdated material, incorporating minor 
changes in text, making editorial corrections, and resolving other 
inconsistencies identified by the licensee's plant operations staff.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments consist of administrative changes to the 
Technical Specifications (TS) for St. Lucie Units 1 and 2. The 
amendments will implement minor changes in text to rectify 
reference, typographic, spelling, and/or consistency-in-format 
errors; update the TS Bases; and/or otherwise improve consistency 
within the TS for each unit. The proposed amendments do not involve 
changes to the configuration or method of operation of 
plantequipment that is used to mitigate the consequences of an 
accident, nor do the changes otherwise affect the initial conditions 
or conservatisms assumed in any of the plant accident analyses. 
Therefore, operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.

[[Page 40850]]

    The proposed administrative revisions will not change the 
physical plant or the modes of plant operation defined in the 
Facility License for each unit. The changes do not involve the 
addition or modification of equipment nor do they alter the design 
or operation of plant systems. Therefore, operation of the facility 
in accordance with the proposed amendments would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendments are administrative in nature and do not 
change the basis for any technical specification that is related to 
the establishment of, or the preservation of, a nuclear safety 
margin. Therefore, operation of the facility in accordance with the 
proposed amendments would not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420
    NRC Project Director: Frederick J. Hebdon

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island 
Nuclear Station, Unit No. 2 (TMI-2), Dauphin County, Pennsylvania

    Date of amendment request: December 2, 1996
    Description of amendment request: The proposed amendment would 
relocate the audit frequency requirements from the plant Technical 
Specifications to the Quality Assurance Plan. In addition, the maximum 
interval between certain types of audits will be extended. This change 
would make the TMI-2 technical specifications consistent with the 
Technical Specifications for Three Mile Island, Unit 1.
    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:
    10 CFR 50.92 provides the criteria which the Commission uses to 
perform a No Significant Hazards Consideration. 10 CFR 50.92 states 
that an amendment to a facility license involves No Significant 
Hazards if operation of the facility in accordance with the proposed 
amendment would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated, or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the technical specifications is 
administrative and does not involve any physical changes to the 
facility. No changes are made to operating limits or parameters, nor 
to any surveillance activities. Based on this, GPU Nuclear has 
concluded that the proposed change does not:
    1. Involve a significant increase in the probability of 
occurrence of the consequences of an accident previously evaluated.
    The proposed amendment is administrative and does not affect the 
function of any system or component. Therefore this change does not 
increase the probability of occurrence or the 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 change is administrative and no new failure modes 
or potential accident scenarios are created.
    3. Involve a change in the margin of safety.
    This change is administrative in nature and does not affect any 
safety settings, equipment, or operational parameters.
    Based on the above analysis it is concluded that the proposed 
changes involve no significant safety hazards considerations as 
defined by 10 CFR 50.92.
    The NRC staff has reviewed the analysis of the licensee and, based 
on this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW, Washington, D.C. 20037 
NRC Project Acting Director: Marvin M. Mendonca

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

    Date of amendment request: July 16, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specification Table 2.2-1 and 3/4.2.5 to allow the 
reactor coolant system total flow to be determined using cold leg elbow 
tap differential pressure measurements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10[]CFR[]50.92 each application for amendment to an 
operating license must be reviewed to determine if the proposed 
change involves a Significant Hazards Consideration. The amendment, 
as defined below, describing the Technical Specification change 
associated with the change has been reviewed and determined to not 
involve Significant Hazards Considerations. The basis for this 
determination follows.
    Proposed Change: The current Technical Specification Table 2.2-1 
(page 2-4) ``Reactor Trip System Instrumentation Trip Setpoints,'' 
provides the Trip Setpoint and Allowable Value for the RCS [reactor 
coolant system] Flow-Low trip. The Allowable Value will be changed 
to reflect the increased uncertainty associated with the correlation 
of the elbow taps to a previous baseline calorimetric. In addition, 
Technical Specification 3.2.5 (page 3/4.2-11), ``Power Distribution 
Limits, DNB Parameters'', will be changed to allow the RCS total 
flow to be measured by the elbow tap [delta]p method. These changes 
will include the modification of surveillance requirement 4.2.5.3, 
which currently requires performance of a precision heat balance 
every 18 months, to allow use of the elbow tap [delta]p method for 
RCS flow measurement. Appropriate Technical Specification Bases 
sections will also be revised to reflect use of the elbow tap 
[delta]p method for flow measurement and to provide clarification. 
The revised Technical Specifications are in Appendix C.
    Background: The 18-month total RCS flow surveillance is 
typically satisfied by a secondary power calorimetric-based RCS flow 
measurement. In recent cycles, South Texas Project has experienced 
apparent decreases in flow rates which have been attributed to 
variations in hot leg streaming effects. These effects directly 
impact the hot leg temperatures used in the precision calorimetric, 
resulting in the calculation of low RCS flow rates. The apparent 
flow reduction has become more pronounced in fuel cycles which have 
implemented aggressive low leakage loading patterns. Evidence that 
the flow reduction was apparent, but not actual, was provided by 
elbow tap measurements. The results of this evaluation, including a 
detailed description of the hot leg streaming phenomenon, are 
documented in Westinghouse report SAE/FSE-TGX/THX-0152, ``RCS Flow 
Verification Using Elbow Taps.''
    South Texas Project intends to begin using an alternate method 
of measuring RCS flow using the elbow tap [delta]p measurements. For 
this alternate method, the RCS elbow tap measurements are correlated 
to precision

[[Page 40851]]

calorimetric measurements performed during earlier cycles which 
decreased the effects of hot leg streaming.
    The purpose of this evaluation is to assess the impact of using 
the elbow tap [delta]p measurements as an alternate method for 
performing the 18-month RCS flow surveillance on the licensing basis 
and demonstrate that it will not adversely affect the subsequent 
safe operation of the plant. This evaluation supports the conclusion 
that implementation of the elbow tap [delta]p measurement as an 
alternate method of determining RCS total flow rate does not 
represent a significant hazards consideration as defined in 
10[]CFR[]50.92.
    Evaluation: Use of the elbow tap [delta]p method to determine 
RCS total flow requires that the [delta]p measurements for the 
present cycle be correlated to the precision calorimetric flow 
measurement which was performed during the baseline cycle(s). A 
calculation has been performed to determine the uncertainty in the 
RCS total flow using this method. This calculation includes the 
uncertainty associated with the RCS flow baseline calorimetric 
measurement, as well as uncertainties associated with [delta]p 
transmitters and indication via QDPS [qualified display processing 
system] or the plant process computer. The uncertainty calculation 
performed for this method of flow measurement is consistent with the 
methodology recommended by the Nuclear Regulatory Commission (NUREG/
CR-3659, PNL-4973, 2/85). The only significant difference is the 
assumption of correlation to a previously performed RCS flow 
calorimetric. However, this has been accounted for by the addition 
of instrument uncertainties previously considered to be zeroed out 
by the assumption of normalization to a calorimetric performed each 
cycle. Based on these calculations, the uncertainty on the RCS flow 
measurement using the elbow tap method is 2.6% flow which results in 
a minimum RCS total flow of 391,500 gpm and must be measured via 
indication with QDPS or the plant process computer at approximately 
100% power.
    The specific calculations performed were for Precision RCS Flow 
Calorimetrics for the specified baseline cycles, Indicated RCS Flow 
(either QDPS or the plant process computer), and the Reactor Coolant 
Flow - Low reactor trip. The calculations for Indicated RCS Flow and 
Reactor Coolant Flow - Low reactor trip reflect correlation of the 
elbow taps to baseline precision RCS Flow Calorimetrics. As 
discussed above, additional instrument uncertainties were included 
for this correlation.
    The uncertainty associated with the RCS Flow - Low trip 
increased slightly. It was determined that due to the availability 
of margin in the uncertainty calculation, no change was necessary to 
either the Trip Setpoint (91.8% flow) or to the current Safety 
Analysis Limit (87% flow) to accommodate this increase. The 
Allowable Value is to be modified to allow for the increased 
instrument uncertainties associated with the [delta]p to flow 
correlation.
    Since the flow uncertainty did not increase over the currently 
analyzed value, no additional evaluations of the reactor core safety 
limits must be performed. In addition, it was determined that the 
current Minimum Measured Flow (MMF) assumed in the safety analyses 
(389,200 gpm) bounds the required MMF calculated for the elbow tap 
method (391,500 gpm).
    Based on these evaluations, the proposed change would not 
invalidate the conclusions presented in the UFSAR [Updated Final 
Safety Analysis Report].
    1. Does the proposed modification involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Sufficient margin exists to account for all reasonable 
instrument uncertainties; therefore, no changes to installed 
equipment or hardware in the plant are required, thus the 
probability of an accident occurring remains unchanged.
    The initial conditions for all accident scenarios modeled are 
the same and the conditions at the time of trip, as modeled in the 
various safety analyses, are the same. Therefore, the consequences 
of an accident will be the same as those previously analyzed.
    2. Does the proposed modification create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    The proposed change revises the method for RCS flow measurement, 
and therefore does not introduce any new accident indicators or 
failure mechanisms.
    No new accident scenarios have been identified. Operation of the 
plant will be consistent with that previously modeled, i.e., the 
time of reactor trip in the various safety analyses is the same, 
thus plant response will be the same and will not introduce any 
different accident scenarios that have not been evaluated.
    3. Does the proposed modification involve a significant 
reduction in a margin of safety[?]
    There are no changes to the Safety Analysis assumptions. 
Therefore, the margin of safety will remain the same.
    The proposed change does not impact the results from any 
accidents analyzed in the safety analysis.
    Conclusion: Based on the preceding information, it has been 
determined that this proposed change to allow an alternate RCS total 
flow measurement based on elbow tap [delta]p measurements does not 
involve a Significant Hazards Consideration as defined by 10 CFR 
50.92(c).
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: James W. Clifford, Acting

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 1, Oswego County, New York

    Date of amendment request: July 2, 1997
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.2.3 regarding reactor coolant 
chemistry in accordance with a report by Electrical Power Research 
Institute, Inc. (EPRI) TR-103515-R1, ``BWR Water Chemistry Guidelines, 
1996 Revision,'' also known as Boiling Water Reactor Vessel and 
Internals Project (BWRVIP)-29. Specifically, the amendment would define 
new conductivity limits in TS 3.2.3a (when reactor coolant is 200 
degrees F or more and reactor thermal power is no more that 10%), and 
in TS 3.2.3b (when reactor thermal power exceeds 10%). The new 
conductivity limits would be 1 micro-mho/cm, which is less than the 
existing limits of 2 micro-mho/cm and 5 micro-mho/cm. The chloride ion 
limit in TS 3.2.3a, 0.1 ppm, would remain at this value but would be 
designated as 100 ppb. The chloride ion limit in TS 3.2.3b would be 
changed from 0.2 ppm to 20 ppb. Sulfate ion limits would be added to TS 
3.2.3a and TS 3.2.3b at 100 ppb and 20 ppb, respectively. In TS 3.2.3c, 
the maximum conductivity limit would be changed from 10 micro-mho/cm to 
5 micro mho/cm when reactor coolant temperature is 200 degrees F or 
more; the maximum chloride ion concentration limit would be changed 
from 0.5 ppm to 100 ppb (when reactor thermal power exceeds 10%) and 
200 ppb (when reactor coolant temperature is 200 degrees F or more and 
reactor thermal power is no more than 10%); and the maximum sulfate ion 
concentration of 100 ppb (when reactor thermal power exceeds 10%) and 
200 ppb (when reactor coolant temperature is 200 degrees F or more and 
reactor thermal power is no more than 10%) would be added. The 
requirement to place the reactor in the cold shutdown condition as 
currently specified in TS 3.2.3d (when TSs 3.2.2a, b, and c are not 
met) and TS 3.2.3e (when the continuous conductivity monitor is 
inoperable for more than 7 days) would be changed to require that the 
reactor coolant temperature be reduced to below 200 degrees F. TS 4.2.3 
would be revised to add that the samples taken and analyzed for 
conductivity and chloride ion content are also to be analyzed for 
sulfate ion content. TS Bases 3/4.2.3 would also be changed to

[[Page 40852]]

reflect that the purpose of TS 3/4.2.3 is to limit crack growth rates 
to values consistent with Unit 1 core shroud analyses in accordance 
with an NRC letter dated May 8, 1997.
    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 operation of Nine Mile Point Unit 1, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The changes to the conductivity and chloride ion action levels 
and the addition of sulfate ion levels as an action level in reactor 
water chemistry are being made to make the TS and its Bases 
consistent with the values used in the core shroud vertical weld 
cracking evaluations. These new values reflect the BWR water 
chemistry guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29) 
and are equal to or more restrictive than the present TS values. No 
physical modification of the plant is involved and no changes to the 
methods in which plant systems are operated are required. None of 
the precursors of previously evaluated accidents are affected and 
therefore, the probability of an accident previously evaluated is 
not increased. These changes to the coolant chemistry TS are more 
restrictive limits and no new failure modes are introduced. 
Therefore, these changes will not involve a significant increase in 
the consequences of an accident previously evaluated.
    2. The operation of Nine Mile Point Unit 1, in accordance with 
the proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The changes to the conductivity and chloride ion action levels 
and the addition of sulfate ion levels as an action level in reactor 
water chemistry are being made to make the TS and its Bases 
consistent with the values use in the core shroud vertical weld 
cracking evaluations. The new values reflect the BWR water chemistry 
guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29) and are 
equal to or more restrictive than the present TS values. No physical 
modification of the plant is involved and no changes to the methods 
in which plant systems are operated are required. The change does 
not introduce any new failure modes or conditions that may create a 
new or different accident. Therefore, this change does not create 
the possibility of a new or different kind of accident [from any 
accident] previously evaluated.
    3. The operation of Nine Mile Point Unit 1, in accordance with 
the proposed amendment, will not involve a significant reduction in 
a margin of safety.
    The changes to the conductivity and chloride ion action levels 
and the addition of sulfate ion levels as an action level in reactor 
water chemistry are being made to make the TS and its Bases 
consistent with the values used in the core shroud vertical weld 
cracking evaluations. These new values reflect the BWR water 
chemistry guidelines, 1996 revision (EPRI TR-103515-R1, BWRVIP-29) 
and are equal to or more restrictive than the present TS values. No 
physical modification of the plant is involved and no changes to the 
methods in which plant systems are operated are required. This 
change does not adversely affect any physical barrier to the release 
of radiation to plant personnel or the public. Therefore, the change 
does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Alex Dromerick, Acting

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: June 19, 1997
    Description of amendment request: Technical Specification Table 
2.2-1 Notes 1 and 3 define the values for the constants used in the 
Overtemperature Delta-T and Overpower Delta-T reactor trip system 
instrumentation setpoint calculators. The proposed amendment would make 
changes to the notes as well as 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 revision in accordance with 
10CFR50.92 and has concluded that the revision does 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 revision does not involve an SHC because the 
revision would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to Technical Specification Table 2.2-1 
Notes 1 and 3 for the addition of the inequalities ensure that the 
constants used for [Overtemperature Delta-T] and [Overpower Delta-T] 
will be set conservatively with respect to the assumptions in the 
accident analysis. The effect on the turbine
    runback function has been evaluated with respect to the Loss of 
External Electrical Load And/Or Turbine Trip analysis and it has 
been determined that this change does not increase the probability 
of this transient. The change was also reviewed to determine if it 
produced an increase in the probability of an unnecessary or 
spurious reactor trip and it was determined that it did not. This 
change does not increase the probability of any previously evaluated 
accident.
    The consequences of previously evaluated accidents, including 
Uncontrolled Rod Cluster Assembly Bank Withdrawal At Power, Rod 
Cluster Control Assembly Misalignment, Uncontrolled Boron Dilution, 
Loss of External Electrical Load And/Or Turbine Trip, Excessive Heat 
Removal Due To Feedwater System Malfunctions, Excessive Load 
Increase Incident, Accidental Depressurization Of The Reactor 
Coolant System, Accidental Depressurization Of The Main Steam 
System, Loss of Reactor Coolant From Small Ruptured Pipes Or From 
Cracks In Large Pipes Which Actuate ECCS [emergency core cooling 
system], or Major Secondary System Pipe Ruptures have not changed.
    The administrative changes have no impact on the design or 
operation of Millstone Unit 3.
    Therefore, the proposed revision 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 changes to Technical Specification Table 2.2-1 
Notes 1 and 3 do not alter the design, construction, operation, 
maintenance or method of testing of equipment. The proposed changes 
alter the Technical Specification description of [an] 
[Overtemperature Delta-T] and [Overpower Delta-T] setpoint functions 
and requires only slight changes to the actual setpoints in the 
field. The [Overtemperature Delta-T] and [Overpower Delta-T] 
functions serve to mitigate the effects of accidents by opening the 
Reactor Trip breakers or reduce power by ``running back'' turbine 
electrical load. The change does not create any new interfaces to 
plant control or protection systems and therefore, no new mechanism 
for accident initiation has been introduced. The proposed change 
does not introduce the possibility of an accident of a different 
type than previously evaluated.
    Therefore, the proposed revision 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 Table 2.2-1 
Notes 1 and 3 do not affect the integrity of any physical fission 
protective boundaries, increase the delays in actuation of safety 
systems beyond that assumed in the safety analysis or reduce the

[[Page 40853]]

margin of safety of any system. These changes ensure that actuation 
of Overtemperature [Delta-T] and Overpower [Delta-T] reactor trips 
will occur conservatively with respect to the assumptions of the 
accident analysis.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision 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: June 19, 1997
    Description of amendment request: Technical Specification 3/4.7.1.3 
requires sufficient water to be available for the auxiliary feedwater 
(AFW) system to maintain the reactor coolant system at hot standby for 
10 hours before cooling down to hot shutdown in the next 6 hours. The 
proposed amendment would increase the required volume of water when the 
condensate storage tank is used, make editorial changes, and expand the 
description in the appropriate 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 basis 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 
4.7.1.3.2 will account for the unusable Condensate Storage Tank 
(CST) inventory by increasing the required combined CST and 
Demineralized Water Storage Tank (DWST) inventory to 384,000 
gallons. The increased required water volume is consistent with the 
design of the CST and will provide assurance that sufficient water 
is available to maintain the reactor coolant system at Hot Standby 
for 10 hours before cooling down to Hot Shutdown in the next 6 
hours.
    The proposed changes to reword Technical Specification 3/
4.7.1.3, expand the description in Bases Section B3/4.7.1.3 and 
modify the description in Bases Section B3/4.7.1.2 are to update and 
clarify the requirements.
    Therefore, the proposed revision 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 changes to Technical Specification 3/4.7.1.3 do not 
change the use of DWST or CST during normal or accident evaluations.
    The proposed changes to reword Technical Specification 3/
4.7.1.3, Bases Section B3/4.7.1.3 and Bases Section B3/4.7.1.2 are 
to update and clarify the requirements.
    Therefore, the proposed revision 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 Surveillance 
4.7.1.3.2 will increase the required inventory for the combined CST 
and DWST to account for an additional 50,000 gallons of unusable 
inventory due to the CST discharge line location, other physical 
characteristics, and measurement uncertainty. The proposed change to 
the surveillance requirement will increase the required volume of 
the combined CST and DWST inventory to 384,000 gallons. The proposed 
change ensures that sufficient water is available to maintain the 
Reactor Coolant System at Hot Standby conditions for 10 hours with 
steam discharge to the atmosphere, concurrent with a total loss-of-
offsite power, and with an additional 6-hour cool down period to 
reduce reactor coolant temperature to 350 [degrees] F.
    The proposed changes to Technical Specification 3/4.7.1.3 and 
Bases Section 3/4.7.1.3 are to clarify the requirements. The 
proposed changes to the Bases Section 3/4.7.1.2 update and expands 
the description of the design bases accidents for which AFW System 
is credited for accident mitigation. This additional information is 
consistent with the current AFW System design bases.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision 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: June 30, 1997
    Description of amendment request: Technical Specification 
Surveillance Requirements 4.7.1.5.1 and 4.7.1.5.2 require the periodic 
testing of the main steam isolation valves (MSIVs) to demonstrate 
operability. The proposed amendment would (1) clarify when the MSIVs 
are partial stroked or full closure tested, (2) add a note to the Mode 
4 applicability of Technical Specification 3.7.1.5 to require that the 
MSIVs be closed and deactivated at less than 320 degrees F, (3) make 
editorial changes, and (4) make changes to the associated Bases 
sections.
    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 basis 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 changes to Technical Specifications Surveillances 
4.7.1.5.1 and 4.7.1.5.2 are to clarify the testing of the MSIVs by 
rewording and separating the requirements into three surveillances.

[[Page 40854]]

Currently, Technical Specifications Surveillance 4.7.1.5.1 requires 
``verifying full closure within 10 seconds ... in MODES 1, 2, and 3 
when tested pursuant to Specification 4.0.5.'' The current 
surveillance requirement to full stroke test the MSIVs is not 
performed during power operation as the Millstone Unit 3 Inservice 
Pump and Valve Test Program pursuant to Specification 4.0.5, has 
received relief from the quarterly full stroke surveillance testing 
requirement. The basis for the relief is that full stroking the 
MSIVs to the closed position during power operation would result in 
an unbalanced steam flow condition producing an abnormal power 
distribution in the reactor core, possibly causing a reactor trip. 
The MSIVs are equipped with provisions for inservice testing by 
partial stroking. The partial stroking is accomplished by opening a 
solenoid valve to admit steam pressure into the lower piston 
chamber. After a time delay the solenoid valve for the upper piston 
chamber opens. After 10 percent travel the position indicating 
device vents both piston chambers and the valve fully opens to the 
back seat due to pressure acting on the valve plug. The accepted 
alternate testing method is to partially stroke test the MSIVs 
during power operation and full stroke test the valves during 
shutdowns.
    The proposed changes to Technical Specifications Surveillance 
4.7.1.5.2 will identify a Mode 3 requirement to perform a 10 second 
full closure test of the MSIVs in Mode 3 or 4. Surveillance 
4.7.1.5.3 will identify a Mode 4 requirement to perform a 120 second 
full closure test of the MSIVs in Mode 4 when the RCS [reactor 
coolant system] temperature is greater than or equal to 320 degrees 
F. The 320 degrees F restriction on testing the valves is consistent 
with recommendations from the valve manufacturer. Additionally, a 
footnote is added to the LCO [limiting condition for operation] and 
the surveillance to identify that the MSIVs are required to be 
closed and deactivated when the RCS temperature is less than 320 
degrees F.
    The proposed changes are consistent with equipment design and 
the surveillance testing of the MSIVs provides the necessary 
assurance that the valves will function consistent with accident 
analyses.
    The other proposed changes to reword the Applicability and 
Action statements of Technical Specification 3.7.1.5 and Bases 
Section B3/4.7.1.5 are considered administrative changes.
    Therefore, the proposed revision 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 changes to the surveillance testing of the MSIVs 
does not change the operation of the valves as assumed for accident 
analyses. The MSIVs are currently equipped with provisions for 
partial stroking.
    Therefore, the proposed revision 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 Specifications Surveillances 
4.7.1.5.1 and 4.7.1.5.2 are to clarify the testing of the MSIVs by 
rewording and separating the requirements into three surveillances. 
Surveillance 4.7.1.5.1 will identify a Mode 1 and 2 requirement to 
partial stroke test the MSIVs in Mode 1 and 2 unless a successful 10 
second full stroke test was performed during the surveillance 
period. Surveillance 4.7.1.5.2 will identify a Mode 3 requirement to 
perform a 10 second full closure test of the MSIVs in Mode 3 or 4. 
Surveillance 4.7.1.5.3 will identify a Mode 4 requirement to perform 
a 120 second full closure test of the MSIVs in Mode 4 when the RCS 
temperature is greater than or equal to 320 degrees F. The 320 
degrees F restriction on testing the valves is consistent with 
recommendations from the valve manufacturer. Additionally, a 
footnote is added to the LCO and the surveillance to identify that 
the MSIVs are required to be closed and deactivated when the RCS 
temperature is less than 320 degrees F. The footnote will eliminate 
the potential to declare the MSIVs operable in the upper range of 
Mode 4 and then allow the MSIVs to remain open during a cooldown 
into the lower range of Mode 4 where they may not be able to meet 
their required stroke time. The full closure test times are 
consistent with the current MSIV surveillances and the partial 
stroke testing is consistent with the Millstone Unit 3 Inservice 
Pump and Valve Test Program.
    The other proposed changes to reword the Applicability and 
Action statements of Technical Specification 3.7.1.5 and Bases 
Section B3/4.7.1.5 are considered administrative changes.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision 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: June 30, 1997
    Description of amendment request: Technical Specifications 4.6.1.1, 
3/4.6.1.2, and 3/4.6.1.3 require the testing of the containment to 
verify leakage limits at a specified test pressure. The proposed 
amendment would (1) modify the list of valves that can be opened in 
Modes 1 through 4, (2) add a footnote on procedure controls, (3) remove 
a footnote on Type A testing, and (4) make editorial changes to the 
Technical Specifications and associated Bases sections.
    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 basis 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 changes to Technical Specification Surveillance 
4.6.1.1.a include the adding ``or procedure control***'' and adding 
footnote ``***''. The changes are requested since the Residual Heat 
Removal System (RHR) valves, 3RHS*MV8701A/B and 3RHS*MV8702A/B, are 
opened during cooldown and heatup in Mode 4. Allowing these 
containment isolation valves to be opened is consistent with 
Technical Specification 3.4.1.3, Reactor Coolant System - Hot 
Shutdown, which allows the RHR system to be used in Mode 4. The 
proposed changes to open the RHR system containment isolation 
valves, under procedure control in Mode 4, do not change the way the 
RHR system is operated or change the operator's response to an 
accident in Mode 4.
    The proposed changes to Technical Specification Surveillance 
4.6.1.1.a Footnote **, include the modification of the valves listed 
in the footnote. Valves 3FPW-V661, 3FPW-V666, 3SAS-V875, 3SAS-V50, 
3CCP-V886, 3CCP-V887 and 3CVS-V13 are being deleted and are local 
manual containment isolation valves. Deleting these valves from the 
list of valves that are allowed to be opened under administrative 
control does not modify plant response to or mitigation strategy for 
any accident. The valves being added, 3MSS*V885, 3MSS*V886, and 
3MSS*V887, are in the steam lines to the steam-driven auxiliary 
feedwater pump. These valves are opened to warm the steam lines 
prior to testing the steam-driven auxiliary feedwater pump. These 
valves were recently reclassified as containment isolation valves, 
which resulted in the need to add them to the list of valves allowed 
to be opened under administrative control. The administrative 
controls include the appropriate considerations that when

[[Page 40855]]

required, containment integrity will be established consistent with 
the assumptions in the design basis analyses.
    The proposed change to Technical Specification Surveillance 
4.6.1.2.a will delete footnote ``*'' which referred to an exemption 
granted by the NRC to permit the Type A test to be delayed until 
RFO6 [refueling outage 6]. However, the current extended shutdown 
has significantly delayed RFO6 and NNECO intends to perform the Type 
A test during this midcycle shutdown. The deletion of the footnote 
does not alter the operation of any system or the containment or 
containment airlocks, as assumed for accident analyses.
    Additionally, Technical Specifications 4.6.1.1, 3/4.6.1.2 and 3/
4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/4.6.1.3 are 
reworded to provide clarity and consistency. These proposed changes 
do not alter the operation of any system or the containment or 
containment airlocks during accident analyses. Therefore, the 
proposed revision does not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    1. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to Technical Specifications 4.6.1.1, 3/
4.6.1.2 and 3/4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/
4.6.1.3 do not alter the operation of any system or the containment 
or containment airlocks, during normal operation or as assumed in 
accident analyses.
    Therefore, the proposed revision 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 Specifications 4.6.1.1, 3/
4.6.1.2 and 3/4.6.1.3 and Bases Sections 3/4.6.1.1, 3/4.6.1.2 and 3/
4.6.1.3 do not alter the design, maintenance or function of any 
system or the containment or the containment airlocks. Additionally, 
the proposed changes do not alter the testing of any system or the 
containment or containment airlocks, or alter any assumption used in 
the accident analyses.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision 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

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of amendment requests: May 14, 1997
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 by revising Technical Specification (TS) 
6.9.1.8.b.5 to replace reference WCAP-10266-P-A with WCAP-12945-P for 
best estimate loss-of coolant accident (LOCA) analysis. The amendment 
would also revise TS Bases 3/4.2.2 and 3/4.2.3 to change the emergency 
core cooling system (ECCS) acceptance criteria limit to state that 
there is a high level of probability that the ECCS acceptance criteria 
limits are not exceeded. This is consistent with the best estimate LOCA 
methodology.
    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 to use of the Best Estimate Loss of Coolant 
Accident (LOCA) analysis methodology does not involve physical 
alteration of any plant equipment or change in operating practice at 
Diablo Canyon Power Plant (DCPP). Therefore, there will be no 
increase in the probability of a LOCA. The consequences of a LOCA 
are not being increased.
    The plant conditions assumed in the analysis are bounded by the 
design conditions for all equipment in the plant. 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, and it meets the five criteria listed in 
Section D. of this evaluation. No other accident is potentially 
affected by this change.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change would not result in any physical alteration 
to any plant system, and there would not be a change in the method 
by which any safety related system performs its function. The 
parameters assumed in the analysis are within the design limits of 
existing plant equipment.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    It has been shown that the analytic technique used in the 
analysis realistically describes the expected behavior of the DCPP 
Units 1 and 2 reactor system during a postulated LOCA. Uncertainties 
have been accounted for as required by 10 CFR 50.46. A sufficient 
number of LOCAs with different break sizes, different locations, and 
other variations in properties have been analyzed to provide 
assurance that the most severe postulated LOCAs were calculated. It 
has been shown by the analysis that there is a high level of 
probability that all criteria contained in 10 CFR 50.46, paragraph 
b, are met.
    Therefore the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of amendment requests: May 15, 1997
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant (DCPP), Unit Nos. 1 and 2 to revise the surveillance 
frequencies from at least once every 18 months to at least once per 
refueling interval (nominally 24 months) including (1) reactor coolant 
system total flow rate, (2) instrumentation for radiation monitoring, 
(3) instrumentation and controls for remote shutdown, (4) 
instrumentation for

[[Page 40856]]

accident monitoring, and (5) several miscellaneous TS.
    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 TS surveillance interval increases do not alter the 
intent or method by which the inspections, tests, or verifications 
are conducted, do not alter the way any structure, system, or 
component functions, and do not change the manner in which the plant 
is operated. The surveillance, maintenance, and operating histories 
indicate that the equipment will continue to perform satisfactorily 
with longer surveillance intervals. Few surveillance and maintenance 
problems were identified. No problems have recurred, or are expected 
to recur, following identification of root causes and implementation 
of corrective actions.
    There was one time-related degradation mechanism identified that 
could significantly degrade the performance of the evaluated 
equipment during normal plant operation. Accumulation of corrosion 
products and debris in the containment fan cooler unit (CFCU) 
monitoring system drain lines could affect the use of the CFCU 
drains as a backup to the containment gaseous monitor for RCS leak 
detection. Primarily because CFCU drain line cleaning has been 
instituted to reduce deposit buildup, and also because the CFCU 
monitoring systems are used as backup and they are redundant by a 
factor of five, it was evaluated that this time-related mechanism 
will not significantly degrade the leak detection performance of the 
CFCUs.
    All other potential time-related degradation mechanisms have 
insignificant effects in the period of interest (24 months plus 25 
percent allowance, or a maximum of 30 months). Instrument drift and 
uncertainty analyses show that, while slight increases in instrument 
drift can occur over a longer period, such increases are minimal and 
remain within specified instrument accuracy and calibration 
allowable values. In cases (pressurizer water level and RVLIS) where 
greater than expected instrument drift has been found, design and 
procedural changes have been implemented to improve the calibration 
process and instrument performance. Based on the past performance of 
the equipment, the probability or consequences of accidents 
previously evaluated would not be significantly affected by the 
proposed surveillance interval increases.
    The changes to commitments related to Bulletin 90-01 are 
supported by the conclusions above, and otherwise do not alter the 
intent or method by which the associated functions are tested, do 
not alter the way any structure, system, or component functions, and 
do not change the manner in which the plant is operated.
    The administrative changes to the Bases sections and to remove a 
duplicate line do not alter the frequency, intent, or method by 
which the associated functions are tested, do not alter the way any 
structure, system, or component functions, and do not change the 
manner in which the plant is operated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The surveillance and maintenance histories indicate that the 
equipment will continue to effectively perform its design function 
over the longer operating cycles. Additionally, the increased 
surveillance intervals do not result in any physical modifications, 
affect safety function performance or the manner in which the plant 
is operated, or alter the intent or method by which surveillance 
tests are performed. No problems have reoccurred following 
identification of root causes and implementation of corrective 
actions. Almost all identified potential time-related degradations, 
including instrument drift, have insignificant effects in the period 
of interest.
    The deposit buildup in the CFCU drain lines is time-related. 
This was evaluated to not to be significant to the leak detection 
function because the CFCUs have a redundancy factor of five (any one 
of the five CFCUs can be used for the leak detection function) and 
because the CFCU drain lines will be cleaned each refueling outage. 
The proposed surveillance interval increases would not affect the 
type or possibility of accidents.
    The changes to commitments related to Bulletin 90-01 are 
supported by the conclusions above, and otherwise do not result in 
any physical modifications, affect safety function performance or 
the manner in which the plant is operated, or alter the intent or 
method by which surveillance tests are performed.
    The administrative change to the Bases sections and to remove a 
duplicate line do not result in any physical modifications, affect 
safety function performance, or alter the frequency, intent, or 
method by which surveillance tests are performed.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Evaluation of historical surveillance and maintenance data 
indicates that there have been few problems experienced with the 
evaluated equipment. There are no indications that potential 
problems would be cycle-length dependent, with the exception of the 
CFCU leak detection function, or that potential degradation would be 
significant for the period of interest and, therefore, increasing 
the surveillance interval will have negligible impact on safety. The 
accumulation of corrosion products and debris in the CFCU drain 
lines is cycle-length dependent, but has been evaluated to have 
insignificant effect on its leak detection function. There is no 
safety analysis impact since these changes will have no effect on 
any safety limit, protection system setpoint, or limiting condition 
for operation, and there are no hardware changes that would impact 
existing safety analysis acceptance criteria. Safety margins are not 
significantly impacted by surveillance intervals or by the slight 
increases in instrument drift that may occur during the extended 
interval.
    The changes to commitments related to Bulletin 90-01 are 
supported by the conclusions above, and otherwise will have no 
effect on any safety limit, protection system setpoint, or limiting 
condition for operation, and there are no hardware changes that 
would impact existing safety analysis acceptance criteria.
    The administrative change to the Bases sections and to remove a 
duplicate line will have no effect on any safety limit, protection 
system setpoint, or limiting condition for operation, and there are 
no hardware changes that would impact existing safety analysis 
acceptance criteria.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay 
Power Plant, Unit 3, Humboldt County, California

    Date of amendment request: December 9, 1996, as supplemented on 
June 12, 1997.
    Description of amendment request: The proposed amendment would 
revise the Humboldt Bay Power Plant (HBPP), Unit 3 Technical 
Specifications (TSs) to incorporate the requirements of 10 CFR Part 50, 
Appendix I, into the Radiological Effluent Technical Specifications 
(RETS) and to relocate the controls and limitations on RETS and 
radiological monitoring from the technical specifications to the 
Offsite Dose Calculation Manual (ODCM) and the Process Control Program 
(PCP). Additional minor administrative changes are proposed to make the 
TSs on High Radiation Areas consistent with

[[Page 40857]]

the revised requirements in the new 10 CFR Part 20.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Operation of the facility in accordance with the proposed 
amendment would not involve any increase in the probability or 
consequences of an accident previously evaluated. This change places 
new requirements in the Administrative Controls section of the 
Technical Specifications to establish programs for the control of 
radiological effluents and the conduct of radiological environmental 
monitoring in the ODCM. The new Administrative Control requirements 
for radiological effluents to be placed in the ODCM incorporate 10 
CFR 50, Appendix I, limitations on dose to individual members of the 
public that are much more restrictive than the current Technical 
Specification limitations. The proposed changes do not involve 
modifications to existing plant equipment, the addition of new 
equipment, or operation of the plant in a different manner than 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability of consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Operation on the facility in accordance with the proposed 
amendment will not create any new or different kind of accident from 
any accident previously evaluated. As stated above, new programmatic 
controls on radiological effluents and radiological environmental 
monitoring are established in the Administrative Controls section of 
the Technical Specifications. Additionally, this change is 
administrative in nature; procedural details for radiological 
effluents and radiological environmental monitoring are being 
relocated to the ODCM and PCP consistent with the guidance provided 
[by the NRC] in Generic Letter 89-01. The proposed changes do not 
involve alterations to plant operating philosophy or methods, or in 
changes to installed plant systems, structures, or components.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Operation of the facility in accordance with the proposed 
amendment would not involve any reduction in the margin of safety. 
These changes do not involve a significant reduction in the margin 
of safety. These changes do not involve a significant reduction in 
the margin of safety. The changes will provide control over 
radiological effluent releases, solid waste management, and 
radiological environmental monitoring activities. Also, these 
changes will increase the margin of safety for members of the public 
by imposing additional controls to ensure that dose to members of 
the public resulting from radioactive effluent releases will be 
maintained ALARA.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the analysis of the licensee and, based 
on this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Humboldt County Library, 636 F 
Street, Eureka, California 95501
    Attorney for licensee: Christopher J. Warner, Esquire, Pacific Gas 
and Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: Seymour H. Weiss

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch 
Nuclear Plant, Unit 1, Appling County, Georgia

    Date of amendment request: May 9, 1997
    Description of amendment request: The proposed amendment would 
revise the Safety Limit Minimum Critical Power Ratio (SLMCPR) in 
Technical Specification (TS) 2.1.1.2 to reflect results of a cycle-
specific calculation performed for Unit 1 Operating Cycle 18 (expected 
to commence November 1997).
    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 specification changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The derivation of the revised SLMCPR for Plant Hatch Unit 1 
Cycle 18 for incorporation into the TS, and its use to determine 
cycle-specific thermal limits, have been performed using NRC 
approved methods. Additionally, interim implementing procedures that 
incorporate cycle-specific parameters have been used which result in 
a more restrictive value for SLMCPR. These calculations do not 
change the method of operating the plantand have no effect on the 
probability of an accident initiating event or transient.
    The basis of the MCPR Safety Limit is to ensure no mechanistic 
fuel damage is calculated to occur if the limit is not violated. The 
new SLMCPR preserves the existing margin to transition boiling and 
the probability of fuel damage is not increased. Therefore, the 
proposed changes do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes result only from a revised method of 
analysis for the Unit 1 Cycle 18 core reload. These changes do not 
involve any new method for operating the facility and do not involve 
any facility modifications. No new initiating events or transients 
result from these changes. Therefore, the proposed TS 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 margin of safety as defined in the TS bases will remain the 
same. The new SLMCPR is calculated using NRC approved methods which 
are in accordance with the current fuel design and licensing 
criteria. Additionally, interim implementing procedures, which 
incorporate cycle-specific parameters, have been used. The SLMCPR 
remains high enough to ensure that greater than 99.9% of all fuel 
rods in the core are expected to avoid transition boiling if the 
limit is not violated, thereby preserving the fuel cladding 
integrity.
    Therefore, the proposed TS changes do 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 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: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, 
Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, 
Georgia

    Date of amendment request: May 9, 1997
    Description of amendment request: The proposed amendments would 
revise the operability requirements for the Rod Block Monitor system of 
Technical Specification (TS) Table 3.3.2.1-1. The amendments would also

[[Page 40858]]

delete the requirements of TS Section 5.6.5 to report Rod Block Monitor 
operability requirements in the cycle-specific Core Operating Limits 
Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

Southern Nuclear Operating Company has evaluated the proposed 
changes to the Plant Hatch Units 1 and 2 Technical Specifications 
in accordance with the criteria specified in 10 CFR 50.92 and has 
determined that they do not involve a significant hazards 
consideration because:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
since they are more restrictive than the existing requirements for 
operation of the plant. These changes provide assurance that the Rod 
Block Monitor system will remain operable when necessary to prevent 
or mitigate the consequences of an anticipated operational 
occurrence that could threaten the integrity of the fuel cladding 
integrity. Since changes in RBM [Rod Block Monitor] operability 
requirements do not involve any physical or functional modifications 
in any plant system, structure or component, there will be no 
increase in the probability or consequences of any accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated because 
they do not involve any changes in the plant configuration or in the 
operation of any system, structure or component.
    3. The proposed changes do not reduce a margin of safety in the 
plant because they impose more restrictive operability requirements 
on the Rod Block Monitor system than those imposed by the existing 
specifications. The changes are more restrictive in that they delete 
the conditions under which the RBM is allowed to be bypassed at core 
thermal power equal to or greater than 29% of rated power. These 
more restrictive requirements ensure the RBM will not only prevent 
fuel rods from under going transition boiling, they also prevent 
fuel rods from exceeding 1% plastic strain (thereby avoiding fuel 
cladding damage) during an RWE [rod withdrawal error] event.
    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: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns 
Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of amendment request: June 19, 1997 (TS 391T)
    Description of amendment request: The proposed amendment extends 
the allowed outage time for emergency diesel generators from 7 to 14 
days on a one-time basis. This extension should permit completion of 
extensive recommended maintenance within a single outage interval.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the ssue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The EDGs [emergency diesel generators] are designed as backup AC 
[alternating current] power sources in the event of loss of off-site 
power. The proposed AOT [allowed outage time] does not change the 
conditions, operating configurations, or minimum amount of operating 
equipment assumed in the safety analysis for accident mitigation. No 
changes are proposed in the manner in which the EDGs provide plant 
protection or which create new modes of plant operation. Also, the 
TS [technical specification] change will improve the overall EDG 
availability by allowing the consolidation of planned maintenance 
outages and, hence, reducing the time period that each EDG will be 
in an outage. Therefore, the proposed amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not introduce any new modes of plant 
operation or make physical changes to plant systems. Therefore, the 
proposed one-time extension of the allowable AOT for EDGs does not 
create the possibility of a new or different accident.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    BFN's [Browns Ferry Nuclear Plant's] emergency AC system is 
designed with sufficient redundancy such that an EDG may be removed 
from service for maintenance or testing. The remaining EDGs are 
capable of carrying sufficient electrical loads to satisfy the UFSAR 
[updated final safety analysis report] requirements for accident 
mitigation or unit safe shutdown.
    Since the 12-year EDG PM [preventive maintenance] work activity 
and vendor recommended PMs are required tasks which must be 
performed, the proposed TS would reduce EDG unavailability since 
multiple outages with resultant longer EDG outage times would not be 
necessary to accomplish the planned maintenance activities.
    The proposed change does not impact the redundancy or 
availability requirements of off-site power supplies or change the 
ability of the plant to cope with station blackout events. The TS 
change improves overall EDG availability. 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: Athens Public Library, South 
Street, Athens, Alabama 35611
    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 1, Ottawa County, Ohio

    Date of amendment request: June 24, 1997
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 3/4.3.2.1, ``Safety 
Features Actuation System Instrumentation,'' TS Section 3/4.6.1.7, 
``Containment Ventilation System,'' TS Section 3/4.6.3.1, ``Containment 
Isolation Valves,'' and TS Section 3/4.9.4, ``Refueling Operations - 
Containment Penetrations,'' and the associated TS Bases. Valve position 
requirements would be added, and certain containment radiation monitor 
requirements, valve isolation verification requirements, and 
containment radiation monitor optional uses would be deleted.
    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:
    Toledo Edison has reviewed the proposed changes and determined 
that a significant hazards consideration does not exist because 
operation ofthe Davis-Besse Nuclear Power

[[Page 40859]]

Station (DBNPS), Unit No. 1, in accordance with this change would:
    1a Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions, or assumptions are affected by the proposed changes.
    The proposed changes to the Technical Specifications and their 
Bases ensure that during Modes 1 through 4 the Containment (CTMT) 
purge and exhaust isolation valves are closed with control power 
removed. Having these valves closed will not increase the 
probability of an accident because these valves are not accident 
initiators. They are used to mitigate the consequences of an 
accident. The proposed changes require these valves to be maintained 
in a closed position as required by design basis accident analysis.
    The removal of the Safety Features Actuation System (SFAS) 
Radiation Monitors (RE's) and their associated SFAS Level 1 
actuations does not affect any accident initiator, condition, or 
assumption.
    During Modes 1 and 2 and partially in Mode 3, for design basis 
accidents which require CTMT isolation, the high/high-high CTMT 
pressure or low/low-low Reactor Coolant System (RCS) signals provide 
CTMT isolation and isolation and actuation of those components 
presently actuated by an SFAS Level 1 High Radiation signal. During 
Mode 3, when the RCS pressure is below 1800 psig, the low RCS 
pressure trip may be manually bypassed, and when the RCS pressure is 
below 600 psig, the low-low pressure trip may be manually bypassed. 
During the short period of time that these bypasses are activated in 
Mode 3, CTMT isolation is only automatically initiated by the CTMT 
high/high-high pressure trips. Manual SFAS actuation is also 
available, including Modes 1 through 4. Removing the SFAS RE's does 
not affect the operation of the SFAS Levels 2-4 actuation since 
these are based only on containment pressure and RCS pressure. 
Therefore, the assumption of CTMT isolation following design basis 
accidents is maintained.
    The SFAS is not required in Mode 5. During Mode 6, the SFAS RE's 
and their associated SFAS Level 1 actuation are not credited during 
a fuel handling accident inside CTMT. The analysis for a fuel 
handling accident inside CTMT assumes that there is no isolation of 
CTMT. The probability of a fuel handling accident is not affected by 
these changes.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
change the source term, CTMT isolation, or allowable releases.
    The proposed changes to the Technical Specifications and their 
Bases ensure that during Modes 1 through 4, the CTMT purge and 
exhaust isolation valves are closed with control power removed.
    Having these valves closed and their control power removed 
ensures that the valves are in and will remain in, the proper 
position for CTMT isolation during and following design basis 
accidents. Also, during Modes 1 and 2 and partially in Mode 3, SFAS 
actuation on high/high-high CTMT pressure or low/low-low RCS 
pressure provides for diverse CTMT isolation. As noted above, during 
Mode 3, when the RCS pressure is below 1800 psig, the low RCS 
pressure trip may be manually bypassed, and when the RCS pressure is 
below 600 psig, the low-low pressure trip may be manually bypassed. 
During the short period of time that these bypasses are activated in 
Mode 3, CTMT isolation is only automatically initiated by the CTMT 
high/high-high pressure trips. In addition, manual SFAS actuation is 
also available, including during Modes 1 through 4. Therefore, 
removal of the SFAS RE's and their actuation signal does not prevent 
CTMT isolation.
    The SFAS RE's and automatic isolation of the CTMT purge and 
exhaust isolation valves during a fuel handling accident is not 
required because the CTMT purge and exhaust isolation system, 
including the associated noble gas monitor, with operator action, 
can provide the necessary actions to mitigate a fuel handling 
accident inside CTMT, assuming the purge and exhaust valves are 
open. Therefore, removing the SFAS RE's and their actuation signal 
will not increase the consequences of an accident because CTMT 
closure is ensured. Further, it is noted that CTMT isolation is not 
assumed in the accident analysis for the fuel handling accident.
    The Containment Radiation-High trip feature is not credited for 
any DBNPS Updated Safety Analysis Report (USAR) accident analysis, 
therefore the proposed removal of this feature will not impact 
radiological consequences of such accidents.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes.
    As stated above, the CTMT purge and exhaust isolation valves, 
the SFAS RE's, and SFAS actuation are not accident initiators. 
Maintaining the CTMT purge and exhaust isolation valves closed and 
control power removed ensures that the design basis assumption of 
CTMT isolation is maintained. Also, since SFAS Levels 2-4 actuation, 
as applicable, on high/high-high CTMT pressure or low/low-low RCS 
pressure or by manual actuation provides the required diversity of 
sensing parameters and isolation of CTMT, the SFAS RE's and their 
associated automatic isolation of the CTMT purge and exhaust 
isolation valves is not required during Modes 1 through 4. 
Therefore, no new or different kind of accident will be introduced.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes maintain a redundant and diverse CTMT 
isolation capability following design basis accidents. Under TS 3/
4.3.2, diversity in achieving CTMT isolation by means of a high/
high-high CTMT pressure or low/low-low RCS pressure SFAS actuation 
will be maintained during Modes 1 through 3 (except during brief 
periods of bypass in Mode 3), and the redundancy of the SFAS sensor 
instrumentation channels and actuation channels themselves will be 
maintained. During Modes 1 through 4 the manual actuation capability 
of SFAS will be maintained. During Modes 1 through 4, control room 
indication of normal and accident range radiation monitoring will be 
maintained in accordance with TS 3/4.3.3.1 and 3/4.4.6.1.
    Under TS 3/4.6.1.7, requiring the CTMT purge and exhaust 
isolation valves to be closed with control power removed, and 
requiring an open CTMT purge and exhaust isolation valve to be 
closed with control power removed within 24 hours is more 
restrictive than the current Technical Specifications or ``The 
Improved Standard Technical Specifications for Babcock and Wilcox 
Plants,'' NUREG-1430, Revision 1. Under TS 3/4.9.4, the existing 
requirements already allow for the SFAS-initiated closure of the 
CTMT purge and exhaust isolation valves to be unavailable and the 
CTMT purge and exhaust system noble gas monitor used as an 
alternative means of achieving CTMT isolation. Further, it is noted 
that CTMT isolation is not credited in the accident analysis for the 
fuel handling accident. Therefore, these proposed changes do not 
significantly reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

Date of application request: April 24, 1997, as supplemented by 
letters dated June 6, 1997, and June 27, 1997.

    Description of amendment request: The amendment would revise 
Section 6.0 of the Technical Specifications to change the title 
``Senior Vice President, Nuclear'' to ``Vice President and Chief 
Nuclear Officer.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This proposed change does not involve any hardware or design 
changes, plant procedures, or administrative changes, other than a 
revision of title designation in documentation. Within the Union 
Electric

[[Page 40860]]

organizational structure, the departments reporting to the former 
Senior Vice-President, Nuclear now report to the Vice President and 
Chief Nuclear Officer. The position of Vice-President and Chief 
Nuclear Officer now reports to the President & Chief Executive 
Officer of Union Electric, which is the same management level of 
reporting as the previous title, Senior Vice-President, Nuclear. 
This change has no impact on the probability or consequences of 
accidents previously evaluated in the Final Safety Report (FSAR).
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change does not involve any hardware or design 
changes, plant procedures, or administrative changes, other than a 
revision of title designation in documentation. Within the Union 
Electric organizational structure, the departments reporting to the 
former Senior Vice-President, Nuclear now report to the Vice 
President and Chief Nuclear Officer. The position of Vice-President 
and Chief Nuclear Officer now reports to the President & Chief 
Executive Officer of Union Electric, which is the same management 
level of reporting as the previous title, Senior Vice-President, 
Nuclear. No new or different kind of accident is introduced by this 
purely administrative change to revise documentation to reflect 
current organizational titles.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The safety margins of the Technical Specifications are based on 
the actual plant design and are unaffected by this purely 
administrative change. This change merely updates the Technical 
Specifications to reflect the current organizational title for 
senior management of the Callaway Plant, and within the 
organizational structure of Union Electric. This change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: May 14, 1997
    Description of amendment request: The proposed change will provide 
clarification to the testing and inspection requirements that each of 
the turbine control valves be cycled and movement verified through at 
least one complete cycle from the running position and revise the 
current wording in Surveillance Requirement 4.7.1.7.2.a for both units 
to clarify the testing and inspection methodology of the turbine 
control valves. Additionally, Technical Specification Bases Section 3/
4.7.1.7 will be revised to clarify the testing requirements for the 
turbine governor control valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Specifically, operation of the North Anna Power Station in 
accordance with the proposed Technical Specifications changes will 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No new or unique accident precursors are introduced by these 
changes in surveillance requirements. The clarification for the 
turbine control valve testing and inspections do not change
    the design, operation, or failure modes of the valves and other 
components in the turbine overspeed protection system.
    The verification of the operability of the turbine control 
valves will continue to provide adequate assurance that the turbine 
overspeed protection system will operate as designed, if needed. 
Therefore, these 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 previous[ly] evaluated.
    Since the implementation of the proposed change to the 
surveillance requirements is to clarify the wording only, operation 
of the facilities with these proposed Technical Specifications does 
not create the possibility for any new or different kind of accident 
which has not already been evaluated in the Updated Final Safety 
Analysis Report (UFSAR).
    The proposed wording changes to the Technical Specifications 
will not result in any physical alteration to any plant system, nor 
would there be a change in the method by which any safety-related 
system performs its function. The design and operation of the 
turbine overspeed protection and turbine control systems are not 
being changed. The proposed change merely represents a clarification 
to more specifically state current test requirements and test 
practice.
    These changes do not change the design, operation, or failure 
modes of the valves and other components of the turbine overspeed 
protection system. 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 would not reduce the margin of safety as 
defined in the basis for any Technical Specifications. The design 
and operation of the turbine overspeed protection and turbine 
control systems are not being changed and the operability of the 
turbine control valves are being demonstrated in the same manner. In 
addition, the results of the accident analyses which are documented 
in the UFSAR continue to bound operation under the proposed changes, 
so that there is no safety margin reduction. Therefore, the proposed 
change does not involve a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: Gordon E. Edison, Acting

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

    Date of amendment request: July 3, 1997
    Description of amendment request: This license amendment request 
revises Technical Specification Section 5.3.1, Fuel Assemblies, to 
allow the use of an alternate zirconium based fuel cladding material, 
ZIRLO. Wolf Creek Nuclear Operating Corporation (WCNOC) is planning to 
insert Westinghouse fuel assemblies containing ZIRLO fuel rod cladding 
during the ninth refueling outage, which is currently scheduled to 
begin in October 1997. This request proposes to incorporate additional 
information, associated with the requested change, into Technical 
Specification 6.9.1.9, ``CORE OPERATING LIMITS REPORT (COLR).'' This 
revised submittal supersedes the staff's proposed no significant 
hazards consideration determination evaluation for the requested 
changes that were published on April 23, 1997 (62 FR 19839).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 40861]]

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 methodologies used in the accident analysis remain 
unchanged. The proposed changes do not change or alter the design 
assumptions for the systems or components used to mitigate the 
consequences of an accident. Use of ZIRLO fuel cladding does not 
adversely affect fuel performance or impact nuclear design 
methodology. Therefore accident analyses are not impacted.
    The operating limits will not be changed and the analysis 
methods to demonstrate operation within the limits will remain in 
accordance with NRC approved methodologies. Other than the changes 
to the fuel assemblies, there are no physical changes to the plant 
associated with this technical specification change. A safety 
analysis will continue to be performed for each cycle to demonstrate 
compliance with all fuel safety design basis.
    VANTAGE 5H with IFMs fuel assemblies with ZIRLO clad fuel rods 
meet the same fuel assembly and fuel rod design bases as other 
VANTAGE 5H with IFMs fuel assemblies. In addition, the 10 CFR 50.46 
criteria are applied to the ZIRLO clad rods. The use of these fuel 
assemblies will not result in a change to the reload design and 
safety analysis limits. The clad material is similar in chemical 
composition and has similar physical and mechanical properties as 
Zircaloy-4. Thus, the cladding integrity is maintained and the 
structural integrity of the fuel assembly is not affected. ZIRLO 
cladding improves corrosion performance and dimensional stability. 
No concerns have been identified with respect to the use of an 
assembly containing a combination of Zircaloy-4 and ZIRLO clad fuel 
rods. Since the dose predictions in the safety analyses are not 
sensitive to fuel rod cladding material, the radiological 
consequences of accidents previously evaluated in the safety 
analysis remain valid.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident or 
malfunction of equipment important to safety previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    VANTAGE 5H with IFMs fuel assemblies with ZIRLO clad fuel rods 
satisfy the same design bases as those used for other VANTAGE 5H 
with IFMs fuel assemblies. All design and performance criteria 
continue to be met and no new failure mechanisms have been 
identified. Since the original design criteria are met, the ZIRLO 
clad fuel rods will not be an initiator for any new
    accident or malfunction of equipment important to safety. The 
ZIRLO cladding material offers improved corrosion resistance and 
structural integrity.
    The proposed changes do not affect the design or operation of 
any system or component in the plant. The safety functions of the 
related structures, systems or components are not changed in any 
manner, nor is the reliability of any structure, system or component 
reduced. The changes do not affect the manner by which the facility 
is operated and do not change any facility design feature, structure 
or system. No new or different type of equipment will be installed. 
Since there is no change to the facility or operating procedures, 
and the safety functions and reliability of structures, systems and 
components are not affected, the proposed changes do not create the 
possibility of a new or different kind of accident or malfunction of 
equipment important to safety from any accident or malfunction of 
equipment important to safety previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Use of ZIRLO cladding material does not change the VANTAGE 5H 
with IFMs reload design and safety limits. The use of these fuel 
assemblies will take into consideration the normal core operating 
conditions allowed in the Technical Specifications. For each cycle 
reload core, the fuel assemblies will be evaluated using NRC 
approved reload design methods, including consideration of the core 
physics analysis peaking factors and core average linear heat rate 
effects.
    The use of Zircaloy-4, ZIRLO or stainless steel filler rods in 
fuel assemblies will not involve a significant reduction in the 
margin of safety because analyses using NRC approved methodologies 
will be performed for each configuration to demonstrate continued 
operation within the limits that assure acceptable plant response to 
accidents and transients. These analyses will be performed using NRC 
approved methods that have been approved for application to the fuel 
configuration.
    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

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

    Date of amendment request: July 3, 1997
    Description of amendment request: This license amendment request 
revises Definition 1.9, ``CORE ALTERATION.'' This change will more 
clearly define the types of components that constitute a core 
alteration when moved.
    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 probability of occurrence of a previously evaluated accident 
is not increased because this change to the definition of core 
alteration does not introduce any new potential accident initiating 
conditions. The proposed change will not affect any previously 
evaluated accident scenario. This proposed change will not affect 
any currently approved refueling-related operating activities. The 
consequences of an accident previously evaluated is not increased 
because the ability of containment to restrict the release of any 
fission product radioactivity to the environment will not be 
degraded by this change.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect any previously evaluated 
accident scenarios, nor does it create any new accident scenarios. 
The proposed change does not alter any of the currently-approved 
refueling operation activities, nor does it create any new refueling 
operating activities.
    Therefore, this proposed change will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    WCGS Technical Specification 3/4.9.1, Boron Concentration, 
specifies that Keff will be maintained equal to or less 
than 0.95 during Operating Mode 6 with fuel in the vessel and the 
vessel head removed. The proposed change in the definition of core 
alteration will allow ``non-core'' components, such as cameras, 
lights, fuel inspection tools, etc., to be moved or manipulated in 
the vessel, with fuel in the vessel and the vessel head removed, 
without constituting a core alteration. This is acceptable because 
these types of components will have no effect on core reactivity, 
and will not affect reactor coolant system boron concentrations. 
Therefore, operations using these types of components will not 
adversely affect Keff or the shutdown margin. Reactor 
subcriticality status is continuously monitored in the control room 
during Operating Mode 6, as specified in WCGS Technical 
Specification 3/4.9.2, Instrumentation.
    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

[[Page 40862]]

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

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

    Date of amendment request: July 3, 1997
    Description of amendment request: This license amendment request 
revises Surveillance Requirements 4.3.1.2 and 4.3.2.2 of Technical 
Specification (TS) 3/4.3.1, ``Reactor Trip System Instrumentation'' and 
TS 3/4.3.2, ``Engineered Safety Features Actuation System 
Instrumentation'' and associated Bases to indicate that the total 
response time will be determined based on the results of WCAP-13632-P-A 
Revision 2, ``Elimination of Pressure Sensor Response Time Testing 
Requirements.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The same RTS [Reactor Trip System] and ESFAS [Engineered Safety 
Features Actuation System] instrumentation is being used. The time 
response allocations/modeling assumptions in the Updated Safety 
Analysis Report Chapter 15 analyses are still the same, only the 
method of verifying time response is changed. The proposed change 
will not modify any system interface and could not increase the 
likelihood of an accident since these events are independent of this 
change. The proposed activity will not change, degrade or prevent 
actions or alter any assumptions previously made in evaluating the 
radiological consequences of an accident described in the USAR. The 
proposed change will not affect the probability of any event 
initiators, nor will the proposed change affect the ability of any 
safety-related equipment to perform its intended function. There 
will be no degradation in the performance of, nor an increase in the 
number of challenges imposed on safety-related equipment assumed to 
function during an accident situation. Therefore, the proposed 
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.
    There are no hardware changes, nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The change will not alter the normal method of plant 
operation. No transmitter performance requirements will be affected. 
This change does not alter the performance of the pressure and 
differential pressure transmitters used in the plant protection 
systems. All sensors will still have response times verified by test 
before placing the sensors in operational service, and after any 
maintenance that could affect response time. Changing the method of 
periodically verifying instrument response for certain sensors 
(assuring equipment operability) from time response testing to 
calibration and channel checks will not create any new accident 
initiators or scenarios. Periodic surveillance of these instruments 
will detect significant degradation in the sensor response 
characteristic. No new transient precursors, failure mechanisms, or 
limiting single failures are introduced as a result of this change. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the acceptance criteria for 
any analyzed event. This change does not affect the total system 
response time assumed in the safety analysis. The periodic system 
response time verification method for selected pressure and 
differential pressure sensors is modified to allow use of actual 
test data or engineering data. The method of verification still 
provides assurance that the total system response is within
    that defined in the safety analysis, since calibration tests 
will detect any degradation which might significantly affect sensor 
response time. There will be no effect on the manner in which safety 
limits or limiting safety system settings are determined, nor will 
there be any effect on those plant systems necessary to assure the 
accomplishment of protection functions. There will be no impact on 
any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

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

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

    Date of application for amendment: January 27, 1997, as 
supplemented May 16, 1997.
    Brief description of amendment: The amendment revised the Technical 
Specifications to permit control rod misalignment of plus or minus 18 
steps when the core power is less than or equal to 85% of rated thermal 
power (RTP) and plus or minus 12 steps above 85% RTP.
    Date of publication of individual notice in Federal Register: June 
19, 1997 (62 FR 33445)
    Expiration date of individual notice: July 21, 1997
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601

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

[[Page 40863]]

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.

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

Date of application for amendments: January 20, 1997

    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.6.3, ``Containment Isolation Valves,'' to reflect 
modifications associated with steam generator replacement for Unit 1 of 
each station. TS Table 3.6-1, ``Containment Isolation Valves,'' will be 
modified to reflect the deletion of feedwater bypass valves and 
reassignment of certain isolation valves to different containment 
penetrations. TS pages for Unit 2 of each station are affected because 
Units 1 and 2 share common TS pages.
    Date of issuance: : July 10, 1997Effective date: Immediately, to be 
implemented within 30 days.
    Amendment Nos.: 91, 90, 84, and 83
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11489). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 10, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481

Commonwealth Edison Company, Docket No. 50-010, Dresden Nuclear 
Generating Station, Unit 1, Grundy County, Illinois

    Date of application for amendment: October 23, 1996, as 
supplemented November 25, 1996, and June 5, 1997.
    Brief description of amendment: The amendment replaces the Appendix 
A Technical Specifications of License DPR-2 in their entirety. The 
amendment revises the Dresden 1 Technical Specifications (TS) to the 
same format as the Dresden Nuclear Power Station, Units 2 and 3 
(Dresden 2/3) Technical Specification Upgrade Program (TSUP).
    Date of issuance: July 8, 1997
    Effective date: July 8, 1997
    Amendment No.: 39
    Facility Operating License No. DPR-2: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4343). The November 25, 1996, and June 5, 1997, submittals provided 
additional 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 July 8, 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 Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: January 20, 1997
    Brief description of amendments: The amendments revise the 
Technical Specifications for various instruments which have alarm or 
indication functions. The amendments relocate surveillance requirements 
for selected instrumentation from Technical Specifications to licensee 
controlled documents or replace selected surveillance requirements with 
those more appropriate to the associated LCOs. In addition, the 
amendments add an action statement related to the automatic 
depressurization system accumulator backup compressed gas system and 
delete action statements related to suppression chamber water level 
instrumentation.
    Date of issuance: July 16, 1997
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 118 and 103
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8795) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 16, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348

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

    Date of amendment request: August 15, 1996, as supplemented by 
letters dated October 31, 1996, and May 29, 1997.
    Brief description of amendments: The amendments removed a 
requirement for performance of a surveillance incorporating a high 
toxic gas test signal.
    Date of issuance: July 17, 1997
    Effective date: July 17, 1997, to be implemented within 30 days.
    Amendment Nos.: Unit 1 - Amendment No. 88; Unit 2 - Amendment No. 
75
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 25, 1996 (61 
FR 50344) The additional information contained in the supplemental 
letters dated October 31, 1996, and May 29, 1997, were clarifying in 
nature and thus, within the scope of the initial notice and did not 
affect the staff's proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 17, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior

[[Page 40864]]

College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 
77488

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

    Date of application for amendment: April 14, 1997
    Brief description of amendment: Technical Specification 3.4.9.3 
requires, in part, that two residual heat removal suction relief valves 
be operable to protect the reactor coolant system from 
overpressurization when any reactor coolant system cold leg is less 
than 350 degrees. The amendment revises the setpoint of the residual 
heat removal suction relief valves.
    Date of issuance: July 10, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 143
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30634) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 10, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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

    Date of application for amendments: June 10, 1996, as supplemented 
July 25, 1996
    Brief description of amendments: These amendments change the 
differential temperature Technical Specifications allowable values and 
trip setpoints for the reactor water cleanup system penetration room 
steam leak detection function.
    Date of issuance: June 26, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 166 and 140
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64389) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 26, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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

    Date of application for amendment: December 23, 1996, as 
supplemented February 26, 1997, May 12, 1997, June 16, 1997, and July 
2, 1997 and July 11, 1997.
    Brief description of amendment: The amendment changes the Technical 
Specifications to allow the use of VANTAGE+ fuel for cycle 10.
    Date of issuance: July 15, 1997
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 175
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6578). The February 26, 1997, May 12, 1997, and June 16, 1997, July 
2, 1997 and July 11, 1997, letters provided information that did not 
change the initial no proposed significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 15, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610

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

    Date of application for amendment: March 31, 1997
    Brief description of amendment: This amendment changes Hope Creek 
Technical Specification Section 3.6.5.3.2, ``Filtration, Recirculation 
and Ventilation System (FRVS),'' to provide an appropriate Limiting 
Condition for Operation and ACTION Statement that reflects the design 
basis for the FRVS.
    Date of issuance: July 9, 1997
    Effective date: July 9, 1997, to be implemented within 60 days
    Amendment No.: 99
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27798) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 9, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Pennsville Public Library, 
190 S. Broadway, Pennsville, NJ 08070

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: June 18, 1996, as supplemented 
August 19, 1996, April 28, 1997, and June 11, 1997
    Brief description of amendments: The amendments change Technical 
Specification (TS) 5.2.2, ``Design Pressure and Temperature,'' by 
adding design parameters for Main Steam Line Break (MSLB). The MSLB 
analysis results in a higher containment air temperature than the value 
that was in TS 5.2.2 prior to the issuance of these amendments.
    Date of issuance: July 17, 1997
    Effective date: July 17, 1997
    Amendment Nos.: 198 and 181
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37302) The supplemental letters did not change the original no 
significant hazards consideration determination nor the Federal 
Register notice. The Commission's related evaluation of the amendments 
is contained in a Safety Evaluation dated July 17, 1997. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: August 30, 1996 (TS 380)
    Brief description of amendment: The amendments remove License 
Condition 2.C.(3) regarding thermal water quality limits.
    Date of issuance: July 8, 1997
    Effective Date: Effective as of the date of issuance.
    Amendment Nos.: 232, 248 and 208
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revise the license.
    Date of initial notice in Federal Register: September 25, 1996 (61 
FR

[[Page 40865]]

50347) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 8, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Athens Public library, South 
Street, Athens, Alabama 35611

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

    Date of application for amendments: August 22, 1996 (TS 96-08)
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) by eliminating the emergency diesel 
generator accelerated testing and special reporting requirements TS 
4.8.1.1.2.a in accordance with NRC Generic Letter 94-01.
    Date of issuance: : July 14, 1997
    Effective date: July 14, 1997
    Amendment Nos.: 226 and 217
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52969) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 14, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendments: December 17, 1996
    Brief description of amendments: The proposed changes will allow 
one of the two service water loops to be isolated from the component 
cooling water heat exchangers (CCHXs) during power operation in order 
to refurbish sections of the isolated service water headers. The 
proposed temporary changes will be valid for two periods of up to 35 
days each for implementation of the service water upgrades associated 
with the repair of the sections of the 24-inch service water supply and 
return piping to/from the CCHXs.
    Date of issuance: July 17, 1997
    Effective date: July 17, 1997
    Amendment Nos.: 205 and 186
    Facility Operating License Nos. NPF-4 and NPF-7:. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6580) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 17, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498Virginia Electric and Power Company, et al., Docket 
Nos. 50-280 and 50-281, Surry Power Station, Units 1 and 2, Surry 
County, Virginia
    Date of application for amendments: November 26, 1997
    Brief description of amendments: These amendments revise the 
Technical Specifications (TSs) to eliminate the records retention 
requirements from Section 6.5 of the TSs. The relocation of those 
requirements to the Operational Quality Assurance program, contained in 
the Final Safety Analysis Report, has been completed.
    Date of issuance: July 15, 1997
    Effective date: July 15, 1997
    Amendment Nos.: 211 and 211
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14472) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 15, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

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

    Date of application for amendments: February 3, 1997, and March 18, 
1997
    Brief description of amendments: These amendments revise the 
Technical Specifications to eliminate the inconsistency between the 
current approved Inservice Inspection Program and ASME Code (1989 
Edition) and the Surry Technical Specifications (TS) as required by 10 
CFR 50.55a(g)95)(ii).
    Date of issuance: July 15, 1997
    Effective date: July 15, 1997
    Amendment Nos.: 212 and 212
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17242) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 15, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

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

    Date of application for amendment: May 20, 1997, as supplemented by 
letters dated June 6, 1997, and July 3, 1997. Additional information 
was also received by letters dated June 12, June 20, and June 25, 1997.
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TS) for the minimum critical power ratio 
(MCPR) safety limit in TS 2.1.1.2 for ATRIUM 9X9 fuel. This change is 
effective for Cycle 13 operation only.
    Date of issuance: July 3, 1997
    Effective date: July 3, 1997, to be implemented within 30 days from 
the date of issuance.
    Amendment No.: 151
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications and operating license.
    Date of initial notice in Federal Register: May 29, 1997 (62 FR 
29160). The June 12, June 20, June 25, and July 3, 1997, submittals 
provided clarifying information which did not affect the initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated July 3, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

    Date of application for amendments: September 30, 1996 (TSCR-192), 
as supplemented on November 26 and December 12, 1996, February 13, 
March 5, April 2, April 16, May 9, June 3, June 13 (two letters), and 
June 25, 1997
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 15.3.3, ``Emergency Core Cooling System, Auxiliary 
Cooling Systems, Air Recirculation Fan Coolers, and Containment 
Spray,'' to incorporate allowed outage times similar to those contained 
in NUREG-1431, Revision 1, ``Westinghouse Owner's Group Improved 
Standard Technical

[[Page 40866]]

Specifications,'' and modify the operability requirements for the 
service water and component cooling water systems. TS 15.3.7, 
``Auxiliary Electrical Systems,'' was revised to reflect the changes to 
the service water system operability requirements. These changes ensure 
that TS requirements are the ``lowest functional capability or 
performance levels of equipment required for safe operation of the 
facility,'' as defined in 10 CFR 50.36(c)(2), ``Limiting Conditions for 
Operation.'' Additionally, the amendments change TS 15.3.12, ``Control 
Room Emergency Filtration,'' to revise charcoal filtration efficiencies 
and to include a specific testing standard, and TS 15.5.2, 
``Containment,'' to revise the design heat removal capability of the 
containment fan coolers.
    Date of issuance: July 9, 1997
    Effective date: July 9, 1997, with full implementation prior to 
restart of Unit 2 and Unit 1 and no later 45 days from the date of 
issuance. Implementation includes incorporating changes to TS 
requirements for the service water system, component cooling water 
systems, and control room ventilating system as detailed in an 
application dated September 30, 1996, as supplemented on November 26 
and December 12, 1996, February 13, March 5, April 2, April 16, May 9, 
June 3, June 13 (two), and June 25, 1997, and evaluated in the staff's 
safety evaluation dated July 9, 1997. These amendments are authorized 
contingent on compliance to commitments provided by the licensee, to 
meet the dose limits associated with Title 10, Code of Federal 
Regulations, Part 50, Appendix A, General Design Criterion (GDC) 19 by: 
(1) submitting a license amendment application including supporting 
analyses and evaluations by February 27, 1998, that contains the 
proposed methods for compliance with GDC 19 dose limits under accident 
conditions based on system design and without reliance on the use of 
potassium iodide and/or self contained breathing apparatus, and (2) 
implementing the proposed changes within 2 years of the date that NRC 
approval for the proposed license amendment is granted. Additionally, 
these amendments are authorized contingent on compliance to commitments 
provided by the licensee, to operate Point Beach Nuclear Plant in 
accordance with its service water system analyses and approved 
procedures. Specifically, each unit will utilize only one component 
cooling water (CCW) heat exchanger until such time that analyses are 
completed and the service water system reconfigured as necessary to 
allow operation of one or both units with two heat exchangers in 
service. If two CCW heat exchangers are required in one or both units 
for maintaining acceptable CCW temperature prior to completion of 
necessary analyses to allow operation in the required configuration, 
the service water system will be considered in an unanalyzed condition, 
declared inoperable and action taken as specified by TS 15.3.0.B except 
for short periods of time as necessary to effect procedurally 
controlled changes in system lineups and unit operating conditions.
    Amendment Nos.: 174 and 178
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Licenses and Technical Specifications. Public comments 
requested as to proposed no significant hazards considerations (NSHC): 
Yes (61 FR 58905 dated November 19, 1996; 62 FR 17244 dated April 9, 
1997; and 62 FR 31636 dated June 10, 1997). No comments have been 
received. The June 10, 1997, notice also provided for an opportunity to 
request a hearing by July 10, 1997, but indicated that if the 
Commission makes a final NSHC determination, any such hearing would 
take place after issuance of the amendments. The June 13 and June 25, 
1997, submittals provided clarifying information within the scope of 
the application and did not affect the staff's previous no significant 
hazards considerations determinations. The Commission's related 
evaluation of the amendments, finding of exigent circumstances, and 
final determination of no significant hazards considerations are 
contained in a Safety Evaluation dated July 9, 1997.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    Local Public Document Room location: The Lester Public Library 1001 
Adams Street, Two Rivers, WI 54241

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

    Date of amendment request: April 23, 1997
    Brief description of amendment: This amendment allows the service 
air and breathing air containment penetrations to remain open under 
administrative control during periods of core alterations or movement 
of irradiated fuel inside containment.
    Date of issuance: July 11, 1997
    Effective date: July 11, 1997, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 107
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30648) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 11, 1997. No significant 
hazards consideration comments received: No.
    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

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of

[[Page 40867]]

telephone comments, the comments have been recorded or transcribed as 
appropriate and the licensee has been informed of the public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By August 29, 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 prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, 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. A copy of the petition should also be sent to the Office of the 
General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-001, 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

[[Page 40868]]

absent a determination by the Commission, the presiding officer or the 
Atomic Safety and Licensing Board that the petition and/or request 
should be granted based upon a balancing of the factors specified in 10 
CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of amendment request: July 10, 1997
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications by deleting the requirements of Surveillance 
Requirements (SR) 4.8.1.1.2.h.2 for the diesel fuel oil system. This 
change will result in testing of the diesel fuel oil system in 
accordance with ASME Code Section XI requirements.
    Date of issuance: July 11, 1997
    Effective date: July 11, 1997, with full implementation within 30 
days.
    Amendment No: 132
    Facility Operating License No. NPF-38: Amendment revises the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated July 11, 1997.
    Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn, 
1400 L Street N.W., Washington, D.C. 20005-3502
    Local Public Document Room location:  University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    NRC Acting Project Director: James Clifford, Acting
    Dated at Rockville, Maryland, this 23rd day of July 1997.
    For The Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 97-19910 Filed 7-29-97; 8:45 am]
BILLING CODE 7590-01-F