[Federal Register Volume 64, Number 249 (Wednesday, December 29, 1999)]
[Notices]
[Pages 73083-73108]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-33684]


-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 4, 1999, through December 17, 1999. 
The last biweekly notice was published on December 15, 1999 (64 FR 
70077).

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

[[Page 73084]]

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 January 28, 2000, 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 electronically from 
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room). If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions,

[[Page 73085]]

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 electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: November 22, 1999.
    Description of amendments request: The proposed amendment revises 
Technical Specification (TS) 5.5.11, ``Ventilation Filter Testing 
Program'' for laboratory testing of charcoal in Clavert Cliffs 
engineered safety feature (ESF) ventilation systems to reference the 
latest charcoal testing standard (American Society for Testing and 
Materials [ASTM] D3803-1989, ``Standard Test Method for Nuclear-Grade 
Activated Carbon''). This TS change was requested by the Nuclear 
Regulatory Commission (NRC) in Generic Letter 99-02, ``Laboratory 
Testing of Nuclear-Grade Activated Charcoal,'' and is based on the 
NRC's determination that testing nuclear-grade activated charcoal to 
standards other than ASTM D3803-1989 does not provide assurance for 
complying with the current licensing basis as it relates to the dose 
limits of General Design Criterion 19 of Appendix A to Part 50 of Title 
10 of the Code of Federal Regulations (10 CFR) and Subpart A of 10 CFR 
Part 100. The generic letter provided a sample TS that the NRC 
considers acceptable. The proposed revision to TS 5.5.11 meets the 
intent of the sample TS. Specifically, the proposed change removes the 
reference to testing in accordance with American National Standards 
Institute N510-1975 and changes the allowable methyl iodide penetration 
to an acceptance criterion that is derived from applying a safety 
factor of two to the charcoal filter efficiency assumed in Calvert 
Cliffs design basis dose analysis. The proposed changes will ensure 
that the charcoal filters used in ESF ventilation systems will perform 
in a manner that is consistent with the particular ESF charcoal 
adsorption efficiencies assumed in the analyses of design basis 
accidents.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    This proposed change makes changes to the methods, test 
conditions, and acceptance criteria associated with the performance 
of the laboratory tests of charcoal samples. The effected equipment 
is used to mitigate the consequences of an accident and are not 
accident initiators. This proposed change does not make any changes 
to the method of obtaining the charcoal sample. No structural 
changes or modifications are being made to the ESF ventilation 
equipment. This proposed change does not make any changes to 
equipment, procedures, or processes that increase the likelihood of 
an accident. Therefore, this proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The ESF ventilation systems are designed to mitigate the 
consequences of accidents. The design basis analysis of the 
accidents account to varying degrees for the reduction in airborne 
radioactive material provided by the charcoal filters. The proposed 
change will change the charcoal filter test protocol to ASTM D3803-
1989. The use of this standard will produce more accurate and 
reproducible laboratory test results and provides a more 
conservative estimate of charcoal filter capability. The proposed 
change makes changes to the methyl iodide penetration acceptance 
criteria to ensure that the charcoal filters are capable of 
performing their required safety function for the expected operating 
cycle. The proposed change will make it more likely that the 
charcoal will meet its intended safety function as described in the 
Updated Final Safety Analysis Report. Therefore, the proposed change 
does not significantly increase the consequences of an accident 
previously evaluated.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change will not make any physical changes to the 
plant or changes to the ESF ventilation system operation. The 
proposed change is limited to the ESF ventilation system testing 
protocol, test conditions, and acceptance criteria. These changes 
are administrative in nature. This proposed change does not make any 
changes to the method of obtaining the charcoal sample. This 
proposed change does not cause any ESF ventilation equipment to be 
operated in a new or different manner. No structural changes or 
modifications are being made to the ESF ventilation equipment. This 
proposed change does not create any new interactions between any 
plant components. Therefore, the possibility of a new or different 
type of accident is not created by this proposed change.
    3. Would not involve a significant reduction in a margin of 
safety.
    The safety function of the ESF ventilation systems is to 
mitigate the consequences of accidents by reducing the potential 
release of radioactive material to the environment or the Control 
Room following a design basis accident. The TS requirements for 
laboratory testing of charcoal samples provides assurance that the 
charcoal filters in these systems are capable of reducing airborne 
radioactive material to within acceptable limits. The proposed 
license amendment requires the use of the latest NRC-accepted 
charcoal testing standard and makes changes to the charcoal testing 
methyl iodide removal efficiency acceptance limits in accordance 
with the formula provided by the NRC in Generic Letter 99-02. The 
proposed license amendment continues to provide assurance that the 
charcoal filters are capable of reducing airborne radioactive 
material to within acceptable limits. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Sheri R. Peterson.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: November 22, 1999.
    Description of amendments request: The Baltimore Gas and Electric 
Company (BGE) requests an amendment to implement a change to the 
Calvert Cliffs Nuclear Power Plant (CCNPP) Updated Final Safety 
Analysis Report (UFSAR) that constitutes an unreviewed safety question 
as described in 10 CFR 50.59.
    The change revises the information currently provided within the 
UFSAR on aircraft and their flight paths for Patuxent River Naval Air 
Station (Pax River NAS). The existing information is outdated and does 
not reflect current conditions for aircraft utilizing Pax River NAS. 
Additionally, the UFSAR will be revised to add information

[[Page 73086]]

pertaining to the corporate helipad located northwest of the plant.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The probability of an aircraft crash was not quantified during 
the timeframe of licensing and construction of the plant. As was 
noted previously, the Directorate of Licensing at the U.S. Atomic 
Energy Commission concurred with Baltimore Gas and Electric 
Company's conclusion that no special design provisions were required 
to be incorporated into Calvert Cliffs Nuclear Power Plant (CCNPP) 
because the probability of an aircraft crash affecting the plant was 
acceptably low (implies a probability of less than 10-7/
Year). Therefore, the probability of an aircraft crash affecting the 
plant was acceptably low at less than 10-7/year.
    The probability of an aircraft accident resulting in 
radiological consequences greater than 10 CFR Part 100 exposure 
guidelines was considered to still be below the Standard Review Plan 
(SRP) (NUREG-0800) level of acceptability of 1.0 x 10-7 
per year for CCNPP. The probability of an aircraft accident during 
the timeframe of original construction and licensing of the plant 
was never quantified. Since today's probability of an aircraft 
accident may be higher based on the fact that, at times, aircraft 
going into Patuxent River Naval Air Station fly over the plant, 
where previously they came no closer than seven miles from the plant 
(as described in the UFSAR), the probability of occurrence of an 
accident will conservatively be considered to have increased. 
However, it should be noted that the probability of an aircraft 
accident resulting in radiological consequences greater than 10 CFR 
Part 100 exposure guidelines is still considered to be below 
1.0 x 10-7 cr/yr, which is acceptable since it is within 
SRP Section 3.5.1.6 guidelines. Since the above probability of an 
aircraft accident meets the criteria of SRP Section 3.5.1.6, no 
additional design or procedural protection is required. Note that 
the SRP criteria is only being used as one acceptable method of 
evaluating risk. Use of this method is not a commitment to the SRP 
and does not incorporate the SRP into our licensing basis.
    Changes to the aircraft flight patterns and/or frequency 
(probability) have no affect on the design or method of operating 
equipment necessary to mitigate the consequences of previously 
analyzed accidents. As was noted above, the aircraft hazard was 
considered to be acceptable and, therefore, no additional design or 
procedural protection is required for the plant. Since the aircraft 
hazard is considered acceptable (where additional design features 
are not required), it can be concluded that no action assumed to 
occur within the accident analysis of CCNPP's Updated Final Safety 
Analysis Report Chapter 14 will be degraded or prevented. Therefore, 
it is concluded that the current calculated aircraft hazard will not 
result in an increase of the consequences of an accident preciously 
evaluated in the UFSAR.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    All possible malfunctions have been previously analyzed. 
Aircraft hazard was addressed within the original design of the 
plant. The frequency/probability of an aircraft crash was considered 
to be so low that special design provisions to protect against 
aircraft crashes did not have to be considered during construction 
of CCNPP. The current calculated aircraft hazard is considered to 
still be within SRP Section 3.5.1.6 guidelines. The possibility for 
a malfunction of a different type than preciously evaluated in the 
UFSAR is not created.
    Aircraft accidents were considered within the original plant 
design. The probability of an aircraft accident resulting in 
radiological consequences greater than 10 CFR Part 100 exposure 
guidelines is still considered to be below the level of 
acceptability (per SRP Section 3.5.1.6) and no special design 
provisions are required. Since an aircraft crash is not a design 
basis concern, it is not plausible that the possibility of a new 
accident is created that has not been previously evaluated in the 
UFSAR. There are also no new challenges to safety-related equipment.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in the margin of 
safety.
    The probability of an aircraft crash affecting the plant, at the 
time of original licensing and construction, was so low that no 
special design provisions were needed in the plant for such an 
event. Since aircraft hazards did not have to be considered within 
the design of the plant, no margin of safety was required or 
established for such a hazard. All of the plant equipment and 
initial condition assumptions stipulated within the UFSAR Chapter 14 
accident analysis would not be affected by such an event.
    The calculated probability of an aircraft accident resulting in 
radiological consequences greater than 10 CFR Part 100 exposure 
guidelines, based on today's aircraft hazard, is considered to be 
below the 1.0 x 10-7 per year stipulated within SRP 
Section 3.5.1.6. Therefore, there is still no need for special 
design provisions within the plant to guard against such an event. 
All of the plant equipment and initial condition assumptions 
stipulated within the UFSAR Chapter 14 accident analysis remain 
unchanged. The plant will continue to operate in such a manner that 
will ensure acceptable levels of protection for the health and 
safety of the public.
    Therefore, this proposed change does not significantly reduce 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Sheri R. Peterson.

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: November 23, 1999
    Description of amendments request: The requested amendments would 
change Technical Specification (TS) 5.5.7.c.1, ``Ventilation Filter 
Testing.'' The testing criteria would be changed consistent with the 
NRC request in Generic Letter 99-02, ``Laboratory Testing of Nuclear-
Grade Activated Charcoal.''
    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 license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment revises TS 5.5.7.c.1 to require testing 
of the SGT [Standby Gas Treatment] system charcoal in accordance 
with American Society for Testing and Materials (ASTM) D3803-1989, 
``Standard Test Method for Nuclear-Grade Activated Carbon.'' Per the 
existing TSs, the SGT system charcoal must meet an acceptance 
criteria of < 1.0% penetration of methyl iodide when tested at a 
relative humidity  70%. CP&L performs this testing in 
accordance with the criteria of Regulatory Position C.6.a of 
Regulatory Guide 1.52, Revision 1, 1976, ``Design, Testing, and 
Maintenance Criteria for Engineered Safety Feature Atmosphere 
Cleanup System Air Filtration and Adsorption Units of Light-Water-
Cooled Nuclear Power Plants.'' As stated in Updated Final Safety 
Analysis Report, Section 6.5.1.1, the purpose of the SGT system, 
along with that of the primary and secondary containment, is to 
mitigate accident consequences. 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 key difference associated with the two 
testing protocols is the testing temperature. Specifically, testing 
to a challenge temperature of 30  deg.C per ASTM D3803-1989 versus 
80  deg.C per Regulatory

[[Page 73087]]

Guide 1.52 results in a much more stringent test. Testing at a 
higher temperature tends to eliminate impurities and moisture from 
the sample. This creates the possibility of the charcoal achieving a 
slightly higher efficiency than actual. Other parameter changes will 
not significantly affect charcoal test performance and will result 
in more accurate and reproducible test results.
    The proposed TS change also includes a requirement that the test 
be performed with a face velocity of 61 fpm. A single BSEP SGT 
system train operates at a maximum flow rate of 4200 scfm which 
corresponds to a face velocity of 61 fpm. In accordance with Generic 
Letter (GL) 99-02, this requirement has been included in TS 
5.5.7.c.1.
    As recommended by GL 99-02, the proposed amendment incorporates 
a safety factor of 2 into the allowed methyl iodide penetration 
limit. The existing TS 5.5.7.c.1 acceptance criteria of 99% does not 
account for a safety factor. In previous testing, CP&L has applied 
the safety factor provided by Regulatory Guide 1.52, Revision 1, 
1976, to the laboratory testing results to ensure proper charcoal 
performance. The proposed changes to TS 5.5.7.c.1 require that 
charcoal samples, tested in accordance with the methodology of ASTM 
D3803-1989, show the methyl iodide penetration to be < 0.5%. The 
0.5% penetration limit is derived by applying a safety factor of 2 
to the 99% filtration efficiency assumed in the current bounding 
calculations for offsite radiological dose release limits. As such, 
the acceptance criteria of < 0.5% penetration of methyl iodide 
ensures that 10 CFR 100 offsite dose limits are not exceeded.
    Based on the more stringent testing temperature requirements, 
the new face velocity testing requirement, and the acceptance 
criteria of < 0.5% penetration of methyl iodide, the proposed change 
will not result in an increase in the consequences of an accident 
previously evaluated.
    2. The proposed license amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes revise the required testing methodology for 
SGT system charcoal. The SGT system is not an initiator of any 
accident, and no new accident precursors are created due to the 
change in the charcoal testing methodology. In addition, the change 
does not alter the design, function, or operation of the SGT system. 
Therefore, the proposed change to test SGT system charcoal in 
accordance with ASTM D3803-1989 will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    The proposed amendment upgrades the SGT system charcoal testing 
requirements to those contained in ASTM D3803-1989. The conditions 
employed by ASTM D3803-1989 were selected to approximate operating 
or accident conditions of a nuclear reactor which could reduce the 
performance of activated carbons. The key difference between CP&L's 
current testing protocol and ASTM D3803-1989 is the testing 
temperature. Specifically, testing to a challenge temperature of 
30 deg.C per ASTM D3803-1989 versus 80 deg.C per Regulatory Guide 
1.52 results in a much more stringent test.
    The proposed TS change also includes a requirement that the test 
be performed with a face velocity of 61 fpm. A single BSEP SGT 
system train operates at a maximum flow rate of 4200 scfm which 
corresponds to a face velocity of 61 fpm. In accordance with GL 99-
02, this requirement has been included in TS 5.5.7.c.1.
    As recommended by GL 99-02, the proposed amendment incorporates 
a safety factor of 2 into the allowed methyl iodide penetration 
limit. The existing TS 5.5.7.c.1 acceptance criteria of 99% does not 
account for a safety factor. In previous testing, CP&L has applied 
the safety factor provided by Regulatory Guide 1.52, Revision 1, 
1976, to the laboratory testing results to ensure proper charcoal 
performance. The proposed changes to TS 5.5.7.c.1 require that 
charcoal samples, tested in accordance with the methodology of ASTM 
D3803-1989, show the methyl iodide penetration to be < 0.5%. The 
0.5% penetration limit is derived by applying a safety factor of 2 
to the 99% filtration efficiency assumed in the current bounding 
calculations for offsite radiological dose release limits. As such, 
the acceptance criteria of < 0.5% penetration of methyl iodide 
ensures that 10 CFR 100 offsite dose limits are not exceeded.
    Based on the more stringent testing temperature requirements, 
the new face velocity testing requirement, and the acceptance 
criteria of < 0.5% penetration of methyl iodide, the proposed change 
does not involve a significant [reduction] in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: November 30, 1999.
    Description of amendment request: The amendment revises Technical 
Specifications (TS) Section 5.5.11, Ventilation Filter Testing Program 
(VFTP) testing requirements. The proposed change requires VFTP testing 
be done according to ASTM D3803-1989 protocol in lieu of previous 
standards.
    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:

    Carolina Power & Light (CP&L) Company has evaluated the proposed 
Technical Specification change and has concluded that it does not 
involve a significant hazards consideration. The CP&L conclusion is 
in accordance with the criteria set forth in 10 CFR 50.92. The bases 
for the conclusion that the proposed change does not involve a 
significant hazards consideration are discussed below.
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change to Technical Specification Section 5.5.11, 
``Ventilation Filter Testing Program,'' does not involve any 
physical alteration of plant systems, structures or components, 
changes in parameters governing normal plant operation, or methods 
of operation. The proposed change updates the required testing of 
Engineered Safety Features (ESF) ventilation filter systems to more 
recent standards accepted by the NRC and described in Generic Letter 
(GL) 99-02, ``Laboratory Testing of Nuclear-Grade Activated 
Charcoal.'' The NRC has found that charcoal filter test protocols 
other than American Society for Testing and Materials (ASTM) 
standard ASTM D3803-1989 do not assure accurate and reproducible 
test results. Since this proposed change references an acceptable 
testing standard and provides assurance that the current licensing 
basis is met, the proposed change does not involve an increase in 
the probability or consequences of an accident previously analyzed.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures or components, changes in parameters 
governing normal plant operation, or methods of operation. The 
proposed change does not introduce a new mode of operation or 
changes in the method of normal plant operation. The proposed change 
introduces a new testing standard for ESF ventilation system 
charcoal samples removed for testing and does not involve 
manipulation of plant systems to perform the charcoal test. 
Therefore, the possibility of a new or different kind of accident 
from any accident previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change revises the required testing standard for 
ESF ventilation charcoal filter systems and does not alter plant 
design margins or analysis assumptions as described in the Updated 
Final Safety Analysis Report. The proposed change does not affect 
any limiting safety system setpoint, calibration method, or setpoint 
calculation. The

[[Page 73088]]

proposed change is more restrictive with regard to testing protocol 
and less restrictive with respect to the allowed penetration during 
testing of the Control Room ventilation system charcoal. However, 
the allowed increase in penetration is in accordance with the method 
for determining the allowable penetration described in GL 99-02. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William D. Johnson, Vice President and 
Corporate Secretary, Carolina Power & Light Company, Post Office Box 
1551, Raleigh, North Carolina 27602
    NRC Section Chief: Richard P. Correia.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington

    Date of amendment request: November 18, 1999.
    Description of amendment request: The proposed amendment requests a 
revision to Technical Specification (TS) 5.5.7.c. The changes would 
revise the requirements that (1) a sample of the charcoal absorber for 
the standby gas treatment (SGT) system and the control room emergency 
filtration (CREF) system be tested in accordance with American Society 
for Testing and Materials (ASTM) D3803-1986, ``Standard Test Method for 
Nuclear-Grade Activated Carbon'', (2) methyl iodide penetration be less 
than a value of .175% for the SGT system and 1.0% for the CREF system, 
and (3) charcoal absorber testing be conducted at a relative humidity 
of greater than or equal to 70%. As requested by Generic Letter (GL) 
99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal,'' 
Energy Northwest proposed that TS 5.5.7.c be revised so that (1) 
testing of charcoal absorber samples be in accordance with ASTM D3803-
1989 at a specified temperature of 30 deg. Centigrade (C) [86 deg. 
Fahrenheit (F)], (2) methyl iodide penetration to be less than a value 
of 0.5% for the SGT system and 2.5% for the CREF system, (3) testing be 
performed at 70% relative humidity, and (4) a face velocity of 75 feet-
per-minute (fpm) will be specified for the SGT system. In addition, the 
revision to TS 5.5.7.c will note that variations in testing parameters 
are permitted in accordance with the guidance in Table 1 and Section 
A5.2 of ASTM D3803-1989.
    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 SGT System is designed to limit the release of airborne 
radioactive contaminants from secondary containment to the 
atmosphere within the guidelines of 10 CFR 100 in the event of a DBA 
[design basis accident]. The CREF System provides a radiologically 
controlled environment from which the plant can be safely operated 
following a DBA. The proposed amendment will require that charcoal 
from these two ESF [engineered safeguard feature] systems be tested 
to the more conservative standards of ASTM D3803-1989. Using the 
more conservative ASTM D3803-1989 testing standard will provide no 
increase in the probability of an accident previously evaluated.
    The staff considers ASTM D3803-1989 to be the most accurate and 
most realistic protocol for testing charcoal in ESF ventilation 
systems because it offers the greatest assurance of accurately and 
consistently determining the capability of the charcoal. Using the 
more conservative ASTM D3803-1989 testing standard will provide 
greater assurance that the ESF ventilation systems will properly 
perform their safety function, thus assuring no increase in the 
radiological consequences of a DBA.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will 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 will not create a new or different kind of 
accident since it only requires that charcoal from the SGT and CREF 
safety-related filtration systems be tested to the more conservative 
standards of ASTM D3803-1989. Using the more conservative ASTM 
D3803-1989 testing standard will provide even greater assurance that 
the ESF ventilation systems will properly perform their safety 
function, thus helping to minimize the radiological consequences of 
a DBA. The increased margin provided by the more conservative 
testing standard will assure no new or different kinds of accidents 
results from the proposed change.
    Therefore, the operation of WNP-2 in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed amendment requires that more conservative ESF 
charcoal filter testing criteria be used to verify ESF ventilation 
systems are operable. More conservative testing criteria will 
provide greater assurance that the ESF ventilation systems will 
properly perform their safety function, thus helping to minimize the 
radiological consequences of a DBA. Using more conservative testing 
criteria will result in maintaining the current 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.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

Energy Northwest, Docket No. 50-397, WNP-2, Benton County,Washington

    Date of amendment request: November 18, 1999.
    Description of amendment request: The proposed amendment requests a 
revision to subsection 4.3.1.2.b of Technical Specification 4.3, Fuel 
Storage. The change would revise the current wording, which describes 
the spacing of the fuel in the new fuel racks, with wording that would 
limit the number of fuel assemblies that may be stored in the facility 
and establish increased spacing limitations for storage of new fuel 
assemblies in the racks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not increase the consequences of any 
previously analyzed accident or transient, since the arrangement of 
new nuclear fuel in storage racks maintains the effective neutron 
multiplication factor much less than 0.95. The change in 
configuration requirements will not increase the probability of any 
previously analyzed accident, because physical constraints are 
installed in the storage racks when new fuel assemblies are 
inserted, assuring that only certain cells can be used for storage 
of new fuel.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will 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.

[[Page 73089]]

    The proposed change is consistent with a new fuel criticality 
analysis performed in support of a previously implemented change to 
Section 9.1 of the FSAR. A variety of accidents were considered in 
that analysis, and it was determined that the effective neutron 
multiplication factor was well below specified limits for any normal 
or accident case.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The current wording of Technical Specification 4.3.1.2.b was 
determined to not provide sufficient margin of safety to assure that 
the requirements of Technical Specification 4.3.1.2.a would be 
maintained. The proposed amendment modifies the requirements for new 
fuel storage configuration for Technical Specification 4.3.1.2.b, to 
assure the margin of safety is maintained for optimum moderation 
conditions.
    Therefore, operation of WNP-2 in accordance with the proposed 
amendment will not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502.
    NRC Section Chief: Stephen Dembek.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc., 
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of amendment request: August 20, 1999.
    Description of amendment request: The proposed amendment request is 
to incorporate 17 improvements (identified by Technical Specification 
Task Force (TSTF) numbers) to the Improved Standard Technical 
Specifications (TSs), NUREG-1434 (for BWR/6 plants such as the Grand 
Gulf plant), that was part of the basis for the current improved TSs 
for Grand Gulf Nuclear Station (GGNS) that were issued in Amendment 120 
dated February 21, 1995. These improvements to the improved TSs for 
BWR/6 plants such as GGNS are identified by TSTF numbers and are the 
following: (1) TSTF-2, relocate the 10 year sediment cleaning of the 
diesel generator fuel storage tank in Surveillance Requirement (SR) 
3.8.3.6 to the GGNS Updated Final Safety Analysis Report (UFSAR), (2) 
TSTF-5, delete notification, reporting, and restart requirements if a 
safety limit is violated in TSs Section 2.2, (3) TSTF-9, relocate the 
shutdown margin values in Limiting Conditions for Operation (LCO) 3.1.1 
and SR 3.1.1.1 to the Core Operating Limits Report (COLR), (4) TSTF-17, 
extension of the testing frequency for the primary containment airlock 
interlock mechanism from 184 days to 24 months in SR 3.6.1.2.3 and 
deletion of the SR Note, (5) TSTF-18, reword and clarify SR 3.6.4.1.2 
to require only one secondary containment access door per access 
opening to be closed, (6) TSTF-32, move the requirement to ensure that 
``slow'' and withdrawn stuck control rods are appropriately separated 
from LCO 3.1.4 requirements to LCO 3.1.3 Condition A Required Actions, 
(7) TSTF-33, administrative change to clarify the Completion Time for 
LCO 3.1.3 Required Action A.2, (8) TSTF-38, revise and clarify the 
visual surveillance in SR 3.8.4.3 for batteries to specify the 
inspection is for performance degradation, (9) TSTF-45, revise SRs 
3.6.1.3.2 and 3.6.1.3.3 to specify that only Primary Containment 
Isolation Valves which are not locked, sealed, or otherwise secured are 
required to be verified closed, (10) TSTF-60, exempt LCO 3.4.7 on 
Reactor Coolant System Leakage Detection Instrumentation from LCO 3.0.4 
which restricts entry into MODES, or specified conditions with required 
equipment inoperable, (11) TSTF-104, relocate the discussion of 
exceptions in LCO 3.0.4 to the Bases of the TSs, (12) TSTF-118, add a 
sentence to the administrative controls program in TSs Administrative 
Controls Section 5.5.9 that the provisions of SRs 3.0.2 and 3.0.3 
applies to the specified testing frequencies of the Diesel Fuel Oil 
Testing Program, (13) TSTF-153, clarify the exception Notes for LCOs 
3.4.9, 3.4.10, 3.9.8, and 3.9.9 to be consistent with the requirement 
being excepted, (14) TSTF-163, modify SRs 3.8.1.2, 3.8.1.12, 3.8.1.15, 
and 3.8.1.20 for diesel generators to provide minimum volt/Hz limits 
for the 10-second acceptance and detail the current volt/Hz range as 
``steady state'' acceptance criteria, (15) TSTF-166, revise LCO 3.0.6 
to explicitly require an evaluation per the Safety Function 
Determination Program and delete the statement that ``additional * * * 
limitations may be required,'' (16) TSTF-278, LCO 3.8.6 is revised to 
require that battery cell parameters be ``within limits,'' the 
reference to Table 3.8.6-1 is deleted, and a reference to the table is 
added to the Actions Table for LCO 3.8.6, and (17) TSTF-279, delete the 
reference to the ``applicable supports'' from the description of the 
``Inservice Testing Program'' in the Administrative Controls TSs, 
Section 5.5.6. The licensee is proposing the current latest revision 
for each TSTF at the time of application with minor exceptions and/or 
clarification in some cases.
    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 (NSHC). The licensee's NSHC is divided into the following 
five categories (which also list the TSTF changes in each category): 
administrative changes, less restrictive changes--removed detail, less 
restrictive changes--relaxation of required action, less restrictive 
changes--deletion of surveillance requirement, and less restrictive 
changes--relaxation of surveillance frequency. The licensee's category 
NSHCs are presented below:

1. Administrative Changes

    These changes involve reformatting, renumbering, and rewording 
of [TSs], with no change in intent. Since they do not change the 
intent of the [TSs] they are considered to be administrative in 
nature. The GGNS is adopting NRC [Nuclear Regulatory Commission] 
approved TSTF-5, TSTF-18, TSTF-33, TSTF-38, TSTF-104, TSTF-118, 
TSTF-153, TSTF-163, TSTF-166, TSTF-278, and TSTF-279, generic 
changes to the Improved Standard Technical Specifications (ISTS) as 
outlined in NUREG-1434, ``Standard Technical Specifications, BWR/6 
Plants.'' In accordance with the criteria set forth in 10 CFR 50.92, 
EOI [Entergy Operations, Inc.] has evaluated these proposed [TSs] 
changes and determined they do not represent a significant hazards 
consideration. The following is provided in support of this 
conclusion.
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change involves reformatting, renumbering, and 
rewording the existing [TSs]. The reformatting, renumbering, and 
rewording process involves no changes in intent to the [TSs]. The 
proposed changes also involve [TSs] requirements, which are purely 
administrative in nature. As such, this change does not [a]ffect 
initiators of analyzed events or assumed mitigation of accident or 
transient events. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or

[[Page 73090]]

different type of equipment will be installed) or changes in methods 
governing normal plant operation. The proposed change will not 
impose any new or eliminate any old requirements. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    c. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not reduce a margin of safety because 
it has no [a]ffect on any safety analyses assumptions. This change 
is administrative in nature. Therefore, the change does not involve 
a significant reduction in a margin of safety.

2. Less Restrictive Changes--Removed Detail

    GGNS is adopting NRC approved TSTF-2, TSTF-9, and TSTF-32 
generic changes to the Improved Standard Technical Specifications 
(ISTS) as outlined in NUREG-1434, ``Standard Technical 
Specifications, BWR/6 Plants.'' The proposed changes involve moving 
details out of the [TSs] and into the [TSs] Bases, the UFSAR, or the 
Core Operating Limits Report (COLR). The removal of this information 
is considered to be less restrictive because it is no longer 
controlled by the [TSs] change process. Typically, the information 
moved is descriptive in nature and its removal conforms with NUREG-
1434 for format and content.
    In accordance with the criteria set forth in 10 CFR 50.92, the 
EOI has evaluated these proposed [TSs] changes and determined they 
do not represent a significant hazards consideration. The following 
is provided in support of this conclusion.
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates certain details from the [TSs] to 
other documents under regulatory control. The Bases and UFSAR will 
be maintained in accordance with 10 CFR 50.59. In addition to 10 CFR 
50.59 provisions, the [TSs] Bases are subject to the change control 
provisions in the Administrative Controls Chapter of the [TSs]. The 
UFSAR is subject to the change control provisions of 10 CFR 
50.71(e). The COLR is controlled in accordance with TS[s] 5.6.5. The 
controls of TS[s] 5.6.5 will ensure that adequate limits are 
maintained and reported to the NRC. Since any changes to these 
documents will be evaluated, no significant increase in the 
probability or consequences of an accident previously evaluated will 
be allowed. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change will not impose or eliminate any requirements, and 
adequate control of the information will be maintained. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    c. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not reduce a margin of safety because 
it has no [a]ffect on any safety analysis assumptions. In addition, 
the details to be moved from the [TSs] to other documents remain the 
same as the existing [TSs]. Since any future changes to these 
details will be evaluated, no significant reduction in a margin of 
safety will be allowed. A significant reduction in the margin of 
safety is not associated with the elimination of the 10 CFR 50.92 
requirement for NRC review and approval of future changes to the 
relocated details. The proposed change is consistent with the BWR/6 
Standard Technical Specifications, NUREG-1434, issued by the NRC 
Staff, revising the [TSs] to reflect the approved level of detail, 
which indicates that there is no significant reduction in the margin 
of safety.

3. Less Restrictive Changes--Relaxation of Required Action

    GGNS is adopting NRC approved TSTF-60 generic changes to the 
Improved Standard Technical Specifications (ISTS) as outlined in 
NUREG-1434, ``Standard Technical Specifications, BWR/6 Plants.'' The 
proposed changes involve relaxation of the Required Actions in the 
current Technical Specifications (TS).
    Upon discovery of a failure to meet an LCO, the TS specifies 
Required Actions to be completed for the associated Conditions. 
Required Actions of the associated Conditions are used to establish 
remedial measures that must be taken in response to the degraded 
conditions. These actions minimize the risk associated with 
continued operation while providing time to repair inoperable 
features. Some of the Required Actions are modified to place the 
plant in a MODE in which the LCO does not apply. Adopting Required 
Actions from this change is acceptable because the Required Actions 
take into account the operability status of redundant systems of 
required features, the capacity and capability of the remaining 
features, and the compensatory attributes of the Required Actions as 
compared to the LCO requirements. These changes have been evaluated 
to not be detrimental to plant safety.
    In accordance with the criteria set forth in 10 CFR 50.92, the 
EOI has evaluated these proposed [TSs] changes and determined they 
do not represent a significant hazards consideration. The following 
is provided in support of this conclusion.
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relaxes Required Actions. Required Actions 
and their associated Completion Times are not initiating conditions 
for any accident previously evaluated and the accident analyses do 
not assume that required equipment is out of service prior to the 
analyzed event. Consequently, the relaxed Required Actions do not 
significantly increase the probability of any accident previously 
evaluated. The Required Actions in the change have been developed to 
provide assurance that appropriate remedial actions are taken in 
response to the degraded condition considering the operability 
status of the redundant systems of required features, and the 
capacity and capability of remaining features while minimizing the 
risk associated with continued operation. As a result, the 
consequences of any accident previously evaluated are not 
significantly increased. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
Required Actions and associated Completion Times in the change have 
been evaluated to ensure that no new accident initiators are 
introduced. Thus, this change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    c. Does this change involve a significant reduction in a margin 
of safety?
    The relaxed Required Actions do not involve a significant 
reduction in the margin of safety. As provided in the justification, 
this change has been evaluated to minimize the risk of continued 
operation under the specified Condition, considering the operability 
status of the redundant systems of required features, the capacity 
and capability of remaining features, a reasonable time for repairs 
or replacement of required features, and the low probability of a 
DBA [design basis accident] occurring during the repair period. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.

4. Less Restrictive Changes--Deletion of Surveillance Requirement

    GGNS is adopting NRC approved TSTF-45 which is a generic change 
to the Improved Standard Technical Specifications (ISTS) as outlined 
in NUREG-1434, ``Standard Technical Specifications, BWR/6 Plants.'' 
The proposed changes involve deletion of [SRs] in the current 
Technical Specifications (TS).
    The TS require safety systems to be tested and verified Operable 
prior to entering applicable operating conditions. These changes 
eliminate unnecessary TS [SRs] that do not contribute to 
verification that the equipment used to meet the LCO can perform its 
required functions. Thus, appropriate equipment continues to be 
tested in a manner and at a frequency necessary to give confidence 
that the equipment can perform its assumed safety function. These 
changes have been evaluated to not be detrimental to plant safety.
    In accordance with the criteria set forth in 10 CFR 50.92, the 
EOI has evaluated these proposed [TSs] changes and determined they 
do not represent a significant hazards consideration. The following 
is provided in support of this conclusion.

[[Page 73091]]

    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change deletes [SRs]. Surveillance's are not 
initiators to any accident previously evaluated. Consequently, the 
probability of an accident previously evaluated is not significantly 
increased. The equipment being tested is still required to be 
Operable and capable of performing the accident mitigation functions 
assumed in the accident analysis. As a result, the consequences of 
any accident previously evaluated are not significantly [a]ffected. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
remaining [SRs] are consistent with industry practice and are 
considered to be sufficient to prevent the removal of the subject 
Surveillance's from creating a new or different type of accident. 
Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    c. Does this change involve a significant reduction in a margin 
of safety?
    The deleted [SRs] do not result in a significant reduction in 
the margin of safety. As provided in the justification, the change 
has been evaluated to ensure that the deleted [SRs] are not 
necessary for verification that the equipment used to meet the LCO 
can perform its required functions. Thus, appropriate equipment 
continues to be tested in a manner and at a frequency necessary to 
give confidence that the equipment can perform its assumed safety 
function. Therefore, this change does not involve a significant 
reduction in a margin of safety.

5. Less Restrictive Changes--Relaxation of Surveillance Frequency

    GGNS is adopting NRC approved TSTF-17 which is a generic change 
to the Improved Standard Technical Specifications (ISTS) as outlined 
in NUREG-1434, ``Standard Technical Specifications, BWR/6 Plants.'' 
The proposed changes involve the relaxation of Surveillance 
Frequencies in the current Technical Specifications (TS).
    Surveillance Frequencies specify time interval requirements for 
performing surveillance testing. Increasing the time interval 
between Surveillance tests results in decreased equipment 
unavailability due to testing which also increases equipment 
availability. Reduced testing can result in a safety enhancement 
because the unavailability due to testing is reduced and[,] in turn, 
reliability of the [a]ffected structure, system or component should 
remain constant or increase. Reduced testing is acceptable where 
operating experience, industry practice or the industry standards 
such as manufacturers' recommendations have shown that these 
components usually pass the Surveillance when performed at the 
specified interval, thus the frequency is acceptable from a 
reliability standpoint. These changes have been found to be 
acceptable based on a combination of the above criteria and have 
been evaluated to not be detrimental to plant safety.
    In accordance with the criteria set forth in 10 CFR 50.92, the 
EOI has evaluated these proposed [TSs] changes and determined they 
do not represent a significant hazards consideration. The following 
is provided in support of this conclusion.
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relaxes Surveillance Frequencies. The 
relaxed Surveillance Frequencies have been established based on 
achieving acceptable levels of equipment reliability. Consequently, 
equipment which could initiate an accident previously evaluated will 
continue to operate as expected and the probability of the 
initiation of any accident previously evaluated will not be 
significantly increased. The equipment being tested is still 
required to be Operable and capable of performing any accident 
mitigation functions assumed in the accident analysis. As a result, 
the consequences of any accident previously evaluated are not 
significantly [a]ffected. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. Thus, 
this change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    c. Does this change involve a significant reduction in a margin 
of safety?
    The relaxed Surveillance Frequencies do not result in a 
significant reduction in the margin of safety. As provided in the 
justification, the relaxation in the Surveillance Frequency has been 
evaluated to ensure that it provides an acceptable level of 
equipment reliability. Thus, appropriate equipment continues to be 
tested at a Frequency that gives confidence that the equipment can 
perform its assumed safety function when required. Therefore, 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: November 4, 1999. This amendment request 
supercedes the licensee's application of June 10, 1999, in its 
entirety. (64 FR 38025)
    Description of amendment request: The proposed amendment would 
remove the existing filter testing requirements of the Technical 
Specifications (TSs) and replace them with a reference to the 
Ventilation Filter Testing Program which is being added to the 
Administrative Controls section of the Davis-Besse TS. The amendment 
introduces TS 6.8.4.f, ``Ventilation Filter Testing Program,'' and 
removes the specific ventilation filter testing requirements from the 
surveillance requirements of TS 3/4.6.4.4, ``Hydrogen Purge System,'' 
TS 3/4.6.5.1, ``Shield Building Emergency Ventilation System,'' and TS 
3/4.7.6.1, ``Control Room Emergency Ventilation System.'' Also included 
are supporting Bases changes to TS 3/4.6.4.4, TS 3/4.6.5.1, and TS 3/
4.7.6.1
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station(DBNPS), Unit Number 1, in accordance with this change would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no change is being made to any 
accident initiator. The replacement of the specific Technical 
Specification (TS) ventilation filter testing Surveillance 
Requirements for the Containment Hydrogen Purge System (3/4.6.4.4), 
Shield Building Emergency Ventilation System (3/4.6.5.1), and the 
Control Room Emergency Ventilation System (3/4.7.6.1), with a 
reference to the newly created Ventilation Filter Testing Program 
contained in TS Administrative Controls Section 6.8.4.f, Ventilation 
Filter Testing Program, is a removal and relocation of certain TS 
details. The proposed TS 6.8.4.f will, however, add controls to 
maintain similar operation, maintenance, testing and system 
operability for these three ventilation systems. The TS Bases 
changes reflect the use of the Ventilation Filter Testing Program.
    The replacement of ASTM D 3803-1979 with ASTM D 3803-1989 for 
laboratory testing of the charcoal filter samples reflects the NRC 
recommendations in Generic Letter 99-02, ``Laboratory Testing of 
Nuclear Grade Activated Charcoal.'' ASTM D 3803-1989 is

[[Page 73092]]

a more stringent testing standard for charcoal filter testing, than 
the present standard referenced by the TS.
    The increase in allowable charcoal penetration due to the use of 
a safety factor of ``2'' is acceptable as a result of using this 
more stringent testing standard.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The increase in allowable 
charcoal penetration due to the use of a safety factor of ``2'' is 
acceptable as a result of using this more stringent testing 
standard. No physical alterations of the DBNPS are involved, nor are 
plant operating methods being changed. The proposed changes do not 
alter the source term, containment isolation or allowable 
radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not change the way the plant is operated. No new or 
different types of failures or accident initiators are being 
introduced by the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because there are no significant changes to the initial conditions 
contributing to accident severity or consequences. Therefore, there 
are no significant reductions in a margin of safety. Testing under 
the more restrictive requirements of ASTM D 3803-1989 will continue 
to ensure that the ventilation systems will perform their safety 
function.

    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.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: December 1, 1999, as supplemented 
December 15, 1999.
    Description of amendment request: The licensee is requesting to 
revise the Turkey Point Plant Physical Security Plan (PSP) to modify 
the PSP requirements for compensation of a security computer failure, 
and to modify the requirements of the minimum security force staffing. 
The December 1, 1999, submittal supersedes two previous submittals 
dated March 10 and June 8, 1999, regarding the same subject. As a 
result of the proposed changes, License Conditions 3.L. for Turkey 
Point Units 3 and 4 Operating Licenses will be updated to reflect the 
latest revision to the Physical Security Plan dated December 1, 1999. 
In addition, the phrase ``Turkey Point Plant, Units 3 and 4 Security 
Plan'' was revised to ``Turkey Point Physical Security Plan.'' The 
latter changes are administrative in nature.
    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 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    These changes will not significantly affect the ability to 
detect a Protected Area intrusion. These changes do not affect the 
ability of a security response to an overt attack on the plant. 
These changes will not affect the ability of the security force to 
respond to contingency events. Therefore, the proposed changes do 
not affect the probability or consequences of accidents previously 
analyzed.
    (2) 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.
    These changes do not affect the ability of the security force to 
defeat the design basis threat. The composition of the response 
organization is not effected by these changes.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The demonstrated level of dependability of the security system 
ensures that a significant reduction in effectiveness or margin of 
safety does not occur.

    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. The staff has also reviewed the changes to License 
Conditions 3.L. for Turkey Point Units 3 and 4 Operating Licenses, as 
well as the change of the security plan title. Based on this review, 
the staff finds that the changes are administrative in nature and that 
they meet the three criteria discussed above. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Correia.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: October 6, 1999.
    Description of amendment request: This proposed Technical 
Specification TS change will revise the Cooper Nuclear Station (CNS) TS 
Sections 1.0, ``Use and Application,'' 3.6, ``Containment Systems,'' 
Bases 3.0, ``Limiting Condition for Operation (LCO) Applicability,'' 
Bases 3.6, ``Containment Systems,'' and 5.5, ``Programs and Manuals,'' 
to adopt the implementation requirements of 10 CFR Part 50, Appendix J, 
Option B, for the performance of Type A, B, and C containment leakage 
rate testing. Contingent upon the Nuclear Regulatory Commission's 
(NRC's) approval of the proposed TS change, the licensee is also 
requesting the NRC to grant the withdrawal of two exemptions. These 
exemptions were previously granted under Option A to 10 CFR Part 50, 
Appendix J; however, under Option B they are no longer required.
    The proposed TS change also contains line-item changes for TS 
requirements addressing containment airlock interlocks, primary and 
secondary containment isolation valves and power-operated automatic 
valves. These changes, along with the specific change to implement 
Option B, have been previously approved by the NRC through submittals 
made by the Nuclear Energy Institute-sponsored TS Task Force.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Implement 10 CFR 50 Appendix J, Option B.
    There is no increase in the probability or consequences of an 
accident since there is no work that would affect containment 
integrity. The testing of containment isolation valves and other 
containment penetration sealing devices are not postulated as an 
accident precursor or initiating event.
    The NRC has concluded, prior to approving Option B, that 
performance-based testing would eliminate or modify prescriptive 
regulatory requirements for which the burden is marginal-to-safety. 
Reviews and analyses considered by the NRC are presented in NUREG-
1493, ``Performance-Based Containment Leak-Test Program, Final 
Report,'' September 1995 (Attachment 2, Reference 12 [of the October

[[Page 73093]]

6, 1999, application]). The historical leakage rate test results for 
Cooper and for the nuclear industry support extension of the testing 
frequencies and demonstrate that structural integrity has been 
maintained.
    Type A testing is capable of determining the total leakage from 
both local leakage paths and gross containment leakage paths. The 
Type B and C testing has consistently provided accurate leakage 
rates for valves and penetrations. Administrative controls govern 
maintenance and testing such that there is very low probability that 
unacceptable maintenance or alignments can occur. Prior to and 
following maintenance on primary containment isolation valves and 
penetrations, a local leak rate test is required to be performed. As 
a result, Type A testing is not required to accurately quantify the 
leakage through containment penetrations.
    Extension of testing frequency of containment airlock interlock 
mechanism from 18 months to 24 months.
    There is no increase in the probability or consequences of an 
accident since there is no work that would affect containment 
integrity. The testing of containment airlock interlocks, isolation 
valves and other containment penetration sealing devices is not 
postulated as an accident precursor or initiating event.
    This changed the testing of the containment airlock interlocks 
from 18 months to 24 months. This testing is only performed during 
periods of reactor shut down and the primary containment is de-
inerted. Thus this change plus the allowance from SR [Surveillance 
Requirement] 3.0.2, provides a total of 30 months, which corresponds 
to the overall airlock leakage test frequency under Option B. In 
this fashion, the interlock can be tested in a Mode where the 
interlock is not required.
    Clarify the Containment Isolation Valve (CIV) surveillance to 
apply to only automatic isolation valves.
    The Bases for SR 3.6.1.3.5 state that the isolation time test 
ensures the valve will isolate in time period less than or equal to 
that assumed in the safety analysis. There may be valves credited as 
containment isolation valves, which are power operated, that do not 
receive a containment isolation signal. These valves do not have an 
isolation time as assumed in the accident analyses since they 
require operator action. However, these valves are tested in 
accordance with the Inservice Test Program as required. Therefore 
this change reduces the potential for misinterpreting the 
requirements of this SR while maintaining the assumptions of the 
accident analysis.
    Based on the above discussion, there is no increase in the 
probability or consequences of an accident, since this change 
provides clarification of the applicability of the SR and has no 
affect on those automatic valves with operating times assumed in the 
accident analysis.
    Allow administrative means of position verification for locked 
or sealed valves.
    It is sufficient to assume that the initial establishment of 
component status (e.g., isolation valve closed) was performed 
correctly. Subsequently verification is intended to ensure the 
component has not been inadvertently repositioned. Given that the 
function of locking, sealing or securing components is to ensure the 
same avoidance of inadvertent repositioning, the periodic re-
verification should only be a verification of the administrative 
control that ensures that the component remains in the required 
state. It would be inappropriate to remove the lock, seal, or other 
means of securing the component solely to perform an active 
verification of the required state. There is no increase in the 
probability or consequences of an accident since the function of 
locking, sealing, or securing components is to ensure that these 
devices are not inadvertently repositioned.
    Therefore, the proposed change described above does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated in the USAR [updated safety analysis 
report].
    The proposed change will not create the possibility of a new or 
different kind of accident than evaluated in the USAR.
    The proposed change involves individual proposed changes related 
to the implementation of 10 CFR 50 Appendix J, Option B, the 
extension of testing frequency of the containment airlock interlock, 
clarification of the CIV surveillance to apply to only automatic 
isolation valves, and the allowance of administrative means of 
position verification for locked or sealed valves. The proposed 
change does not result in any physical change to plant structures, 
systems, or components. The proposed change does not alter the form, 
fit, or function of any equipment or components credited in the 
accident analyses described in the USAR. The performance history of 
containment testing verifies that containment integrity has been 
maintained.
    The frequency changes allowed by the implementation of the 
applicable proposed TS changes will not significantly decrease the 
level of confidence in the ability of the containment to limit 
offsite doses to allowable values. No accident or malfunction can be 
the result of the allowed changes to test schedule or frequency.
    Since the proposed changes will not directly impact equipment, 
procedures or operations, the changes will not create the 
possibility of any new or different kind of accident from any 
accident previously evaluated in the USAR.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident.
    The proposed change will not involve a significant reduction in 
a margin of safety.
    The reason for performing containment leakage rate testing is to 
assure that the leakage paths are identified, and that any accident 
release will be restricted to those paths assumed in the safety 
analysis. The purpose for the schedule is to assure that containment 
integrity is verified on a periodic basis. Implementation of Option 
B to provide flexibility in the scheduled requirements does not mean 
that containment integrity will be compromised.
    The NRC has concluded, prior to approving Option B, that 
performance-based testing would eliminate or modify prescriptive 
regulatory requirements for which the burden is marginal-to-safety. 
Reviews and analyses considered by the NRC are presented in NUREG-
1493, ``Performance-Based Containment Leak-Test Program, Final 
Report,'' September 1995 (Attachment 2, Reference 12). The 
historical leakage rate test results for CNS and for the nuclear 
industry support extension of the testing frequencies and 
demonstrate that structural integrity has been maintained.
    Administrative controls govern position verification for locked 
or sealed valves such that there is a very low probability that 
unacceptable alignment can occur.
    When the containment airlock interlock is opened during times 
the interlock is required, the operator first verifies that one door 
is completely shut before attempting to open the other door. 
Therefore, the interlock is not challenged except during actual 
testing of the interlock. Therefore, it should be sufficient to 
ensure proper operation of the interlock by testing the interlock on 
a 24 month interval.
    There may be valves credited as containment isolation valves, 
which are power operated, that do not receive a containment 
isolation signal. These valves do not have an isolation time as 
assumed in the accident analyses since they require operator action. 
However, these valves are tested in accordance with the Inservice 
Test Program as required and as such there will be no reduction in a 
margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: December 6, 1999.
    Description of amendment request: Changes are proposed to Technical 
Specification (TS) Section 2.1.1.2 for the safety limit minimum 
critical power ratio (SLMCPR). The proposed changes to TS 2.1.1.2 
revise the SLMCPR values from 1.06 to 1.08 for two recirculation loop 
operation, and from 1.07 to 1.09 for single recirculation loop 
operation.
    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:


[[Page 73094]]


    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Evaluation: The basis for the Safety Limit Minimum Critical 
Power Ratio (SLMCPR) is to ensure that at least 99.9% of all fuel 
rods in the core avoid transition boiling if the SLMCPR limit is not 
violated. The revised SLMCPR values preserve the existing margin to 
transition boiling and thus the probability for fuel damage is not 
increased. The determination of a revised SLMCPR Technical 
Specification value does not affect the assumptions of accidents 
previously evaluated; or initiate, or affect initiators, of 
accidents previously evaluated. The proposed revisions to SLMCPR are 
based on the use of methodology previously accepted by the NRC for 
calculating SLMCPR and do not change the definition of SLMCPR. Thus, 
the probability of an accident previously evaluated is not 
increased.
    The revised SLMCPR values do not affect the design or operation 
of any system, structure, or component in the facility. No new or 
different type of equipment is installed by this change. The 
proposed revision does not change or alter the design assumptions 
for systems, structures, or components used to mitigate the 
consequences of an accident. Thus, he dose consequences of an 
accident previously evaluated are not increased.
    Therefore, the proposed license amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    Evaluation: The SLMCPR ensures that at least 99.9% of all fuel 
rods in the core avoid transition boiling if the SLMCPR limit is not 
violated. The revised SLMCPR values preserve the existing margin to 
transition boiling. The proposed revisions to SLMCPR are based on 
the use of methodology previously accepted by the NRC for 
calculating SLMCPR and do not change the definition of SLMCPR. The 
proposed revision does not change the design or operation of any 
system, structure, or component. No new or different type of plant 
equipment is installed by this change. The proposed revision does 
not involve a change to plant operation or allowable plant operating 
modes. The calculational methodology used to determine a revised 
SLMCPR Technical Specification value cannot initiate or create a new 
or different type of accident.

    Therefore, the proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed license amendment create a significant 
reduction in the margin of safety?
    Evaluation: The SLMCPR ensures that at least 99.9% of all fuel 
rods in the core avoid transition boiling if the SLMCPR limit is not 
violated. The revised SLMCPR values were calculated using a 
methodology previously accepted by the NRC, and preserve the 
existing margin to transition boiling and thus the margin of safety 
to fuel failure. The proposed change does not involve a relaxation 
of the criteria or basis used to establish safety limits, or a 
relaxation in the criteria or bases for the limiting conditions for 
operation. The assumptions and methodologies used in the plant 
accident analysis remain unchanged. Therefore, the proposed change 
does not create 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.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Section Chief: Robert A. Gramm.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: December 16, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specification (TS) Safety Limit Minimum Critical 
Power Ratio (SLMCPR) values for two recirculation pump and single-loop 
operation, delete cycle specific footnotes, update the single-loop 
operation Average Planar Heat Generation rate limiting values, correct 
a typographical error, and delete an obsolete reference to Siemens 
fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    GE [General Electric] has recently revised their single loop 
operation (SLO) analysis review procedures to add an additional 
requirement that the peak cladding temperature (PCT) during a LOCA 
[loss-of-coolant accident] initiated while in SLO should be bounded 
by the PCT for a LOCA initiated while in dual loop operation. This 
desired result is enforced by revising the SLO MAPLHGR [maximum 
average planar linear heat generation rate] ``multipliers'' found in 
Technical Specification 3.11.A from the current value of 0.85 for 
all fuel to values of 0.78 for GE10 fuel and 0.80 for GE11 and GE12 
fuel. This change ensures that the condition that the Upper Bound 
PCT does not exceed 1600  deg.F (as required by the NRC-approved 
SAFER methodology for performing ECCS [emergency core cooling 
system] LOCA calculations) is satisfied even if a LOCA were to occur 
while operating in SLO. This change does not alter the method of 
operating the plant and does not increase the probability of an 
accident initiating event or transient. These limits are established 
to preserve required margins.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    SLMCPR is a TS numerical value designed to ensure that 
transition boiling does not occur in greater than 99.9% of all fuel 
rods in the core during the limiting postulated transient. A change 
in SLMCPR cannot create the possibility of any new type of accident. 
SLMCPR values for the new fuel cycle are calculated using previously 
transmitted methodology. Similarly, changes to the SLO MAPLHGR 
multiplier values are designed to ensure that the PCT resulting from 
a LOCA while operating in SLO are bounded by the PCT from a LOCA 
while operating in dual loop operation. Thus, a change in these 
multipliers cannot create the possibility of any new type of 
accident. This multiplier update results from application of GE's 
current standard methodology for this analysis.
    The proposed changes result only from a specific analysis for 
the Monticello core reload design and deletion of a cycle specific 
reference for the values. These changes do not involve any new or 
different 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 amendment will not involve a significant 
reduction in the margin of safety.
    SLMCPR calculations are based on ensuring that greater than 
99.9% of all fuel rods in the core avoid transition boiling if the 
limit is not violated. Proposed SLMCPRs preserve required margin to 
transition boiling and fuel damage in the event of a postulated 
transient. Fuel licensing acceptance criteria for SLMCPR 
calculations apply to Monticello Cycle 20 in the same manner as 
applied in previous cycles. The revised SLMCPR values do not change 
the method of operating the plant and have no effect on the 
probability of an accident-initiating event or transient because 
these limits are established to preserve required margin.
    Fuel licensing acceptance criteria for SLMCPR calculations apply 
to Monticello Cycle 20 in the same manner as previously applied. 
SLMCPRs prepared by GE using methodology previously transmitted to 
the NRC ensure that greater than 99.9% of all fuel rods in the core 
will avoid transition boiling if the limit is not violated, thereby 
preserving fuel cladding integrity. The operating MCPR limit is set 
appropriately above the safety limit value to ensure

[[Page 73095]]

adequate margin when the cycle specific transients are evaluated.
    Application of new SLO MAPLHGR multiplier values ensures that 
SLO LOCA results are bounded by those for dual loop operation and 
thus maintain or improve the margin of safety for LOCA analyses.
    Therefore, the proposed TS changes do not involve a reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: August 5, 1999.
    Description of amendment request: The proposed amendment would add 
a license condition denoting NRC approval of the Trojan Nuclear Plant 
(TNP) License Termination Plan.
    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 requested license amendment does not authorize additional 
plant activities beyond those that already may be conducted under 
the approved TNP Decommissioning Plan and the Defueled Safety 
Analysis Report (DSAR). Accident analyses are included in the 
approved TNP Decommissioning Plan and incorporated into the TNP 
DSAR. No systems, structures, or components that could initiate or 
be required to mitigate the consequences of an accident are affected 
by the proposed change in any way not previously evaluated in the 
approved TNP Decommissioning Plan and DSAR. Therefore, the proposed 
change is administrative in nature and as such does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested license amendment does not authorize additional 
plant activities beyond those that already may be conducted under 
the approved TNP Decommissioning Plan and the DSAR. Accident 
analyses are included in the approved TNP Decommissioning Plan and 
incorporated into the DSAR. The proposed change does not affect 
plant systems, structures, or components in any way not previously 
evaluated in the approved TNP Decommissioning Plan and DSAR, and no 
new or different failure modes will be created. Therefore, the 
proposed change is administrative in nature and as such 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.
    Approval of the TNP License Termination Plan by license 
amendment is administrative in nature since the decommissioning and 
fuel storage activities described in the TNP license Termination 
Plan are consistent with those in the approved TNP Decommissioning 
Plan and DSAR. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRC Section Chief: Michael T. Masnik.

Power Authority of the State of New York, Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: November 24, 1999.
    Description of amendment request: The proposed amendment would 
delete Section 4.7.D.1.e of Appendix A (Technical Specifications (TSs)) 
to the James A. FitzPatrick Operating License to eliminate the 
surveillance requirement for partially stroking of the plant Main Steam 
Isolation Valves (MSIVs) twice a week. The MSIVs will continue to be 
fully stroked with a frequency that is in accordance with the In-
Service Testing (IST) Program per TS 4.7.D.1.d, which is consistent 
with the Boiling-Water Reactor Standard Technical Specification and the 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code. The proposed changes include associated administrative changes to 
Section 4.7.D.1.d, and to Bases Section 4.7.D of the TSs.
    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 will not significant[ly] increase the 
probability or consequences of any previously evaluated accidents.
    This proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
This proposed change deletes the requirement to exercise the MSIVs 
twice per week. The twice per week exercise involves partial closure 
of each individual MSIV and subsequent reopening to the full open 
position.
    The safety function of the MSIV is to isolate the main steam 
line in case of a steam line break, Control Rod Drop Accident or 
Loss of Coolant Accident in order to limit the loss of reactor 
coolant and/or the release of radioactive materials. The MSIVs 
perform a safety function which mitigates the consequences of 
accidents: however, an event can be initiated by the inadvertent 
closure of MSIVs. Therefore, eliminating excessive operation of the 
MSIVs reduces the probability of an inadvertent closure. Also, the 
surveillance which is being deleted does not test the safety 
function of the MSIVs. The safety function is tested during the full 
stroke fast closure test. Since deleting the twice per week exercise 
of the valves is not considered to have any effect on the 
reliability of the MSIVs to perform there safety function, there is 
no increase in the consequences of any postulated accidents. 
Therefore, deleting the requirement for twice per week exercising of 
the MSIVs does not significantly increase the probability or 
consequences of any previously evaluated accidents.
    (2) The proposed change will not create the possibility of a new 
or different kind of accident.
    The safety function of the MSIV is to isolate the main steam 
line in case of a steam line break, Control Rod Drop Accident, or 
Loss of Coolant Accident in order to limit the loss of reactor 
coolant and/or the release of radioactive materials. The MSIVs 
perform a safety function which mitigates the consequences of 
accidents: however, an event can be initiated by the inadvertent 
closure of MSIVs. The inadvertent closure of the MSIVs event has 
been previously evaluated in Chapter 14 of the James A. FitzPatrick 
Final Safety Evaluation Report (FSAR). The surveillance which is 
being deleted does not test the safety function of the MSIVs. The 
safety function is tested during the full stroke fast closure test. 
Since the MSIVs perform a mitigating safety function, and the MSIV 
full stroke fast closure test adequately tests the safety function, 
elimination of the twice per week exercise will not create any new 
or different kind of accident.
    (3) The proposed change will not involve a significant reduction 
in a margin of safety.
    The safety function of the MSIV is not tested during the twice 
per week exercise. The ability of the MSIVs to perform their safety 
function is tested during the MSIV full stroke fast closure test in 
accordance with the IST Program. Therefore, deletion of the 
requirement does not reduce the margin of safety. The exercising of 
the MSIVs was

[[Page 73096]]

originally specified in order to detect binding of the pilot valve. 
The type of pilot valve that was susceptible to binding was replaced 
and there is no longer any need for frequent exercising of the 
MSIVs. The full closure test of the MSIVs in accordance with the IST 
Program adequately demonstrates that the MSIVs and their pilot 
valves are not binding and that the MSIVs will perform their safety 
function. Additionally, reducing the frequency of MSIV operation 
reduces the probability of inadvertent scrams and transients, and 
challenges to relief valves, providing a net addition to the margin 
of safety. The full stroke fast closure test is considered to be 
sufficient. It is the only test required by the ASME Boiler and 
Pressure Vessel Code and the BWR Standard Technical Specifications. 
Therefore, eliminating the twice per week exercise of the MSIVs does 
not significantly reduce any margin of safety.
    The proposed change will not increase the probability or 
consequences of any previously analyzed accident, introduce any new 
or different kind of accident previously evaluated, or reduce 
existing margin to safety. Therefore, the proposed license amendment 
will not involve a significant hazards consideration.

    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.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Alexander W. Dromerick (Acting Section Chief).

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

    Date of amendment request: November 24, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to implement Filtration, 
Recirculation, and Ventilation System (FRVS) and Control Room Emergency 
Filtration (CREF) System charcoal filter testing requirements that are 
consistent with the U. S. Nuclear Regulatory Commission requirements 
delineated in Generic Letter 99-02, ``Laboratory Testing of Nuclear-
Grade Activated Charcoal.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS change does not involve any physical changes to 
plant structures, systems or components (SSC). The CREF and FRVS 
systems will continue to function as designed. The CREF and FRVS 
systems are designed to mitigate the consequences of an accident, 
and therefore, can not contribute to the initiation of any accident. 
The proposed TS surveillance requirement changes implement testing 
methods that more appropriately demonstrate charcoal filter 
capability and establish acceptance criteria, which ensure that Hope 
Creek's licensing and design basis assumptions are met.
    In addition, this proposed TS change will not increase the 
probability of occurrence of a malfunction of any plant equipment 
important to safety, since the manner in which the CREF and FRVS 
systems are operated is not affected by these proposed changes. The 
proposed surveillance requirement acceptance criteria ensure that 
the FRVS and CREF safety functions will be accomplished. Therefore, 
the proposed TS changes would not result in the increase of the 
consequences of an accident previously evaluated, nor do they 
involve an increase in the probability of an accident previously 
evaluated.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS changes do not involve any physical changes to 
the design of any plant SSC. The design and operation of the CREF 
and FRVS systems are not changed from that currently described in 
Hope Creek's licensing basis. The CREF and FRVS systems will 
continue to function as designed to mitigate the consequences of an 
accident. Implementing the proposed charcoal filter testing methods 
and acceptance criteria does not result in plant operation in a 
configuration that would create a different type of malfunction to 
the CREF and FRVS systems than any previously evaluated. In 
addition, the proposed TS changes do not alter the conclusions 
described in Hope Creek's licensing basis regarding the safety 
related functions of these systems.
    Therefore, the proposed TS change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes contained in this submittal would implement 
TS requirements that: (1) Are consistent with the requirements 
delineated in Generic Letter 99-02; (2) implement testing methods 
that adequately demonstrate charcoal filter capability; and (3) 
establish acceptance criteria consistent with Hope Creek's licensing 
basis. The ability of CREF and FRVS to perform their safety 
functions is not adversely affected by these proposed changes. 
Therefore, the proposed TS change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

Southern California Edison Company, et al., Docket Nos. 50-206, 50-361, 
and 50-362, San Onofre Nuclear Generating Station, Units 1, 2, and 3, 
San Diego County, California

    Date of amendment requests: December 2, 1999 (Unit 1--PCN 267, 
Units 2 and 3--PCN 506).
    Description of amendment requests: This amendment application is a 
request to revise the Unit 1 Technical Specifications Section D6, 
Administrative Controls, to be consistent with the San Onofre Units 2 
and 3 Technical Specification Section 5.0, Administrative Controls, and 
incorporate changes related to certified fuel handlers and 10 CFR 
50.54(x), administrative control of working hours and working hour 
deviation approvals, position titles and responsibilities and 
organizational description reference, qualifications for a multi-
discipline supervisor, quality assurance program control of review and 
audit and record retention procedures, high radiation area controls, 
description of the plant configuration for environmental protection, 
and environmental protection related document reporting.
    This amendment application also requests to revise the Unit 2 and 
Unit 3 Technical Specifications, Section 5.0, Administrative Controls, 
to incorporate changes related to the operating organization, working 
hours deviation approvals, qualifications for a multi-discipline 
supervisor, the schedule for submitting Technical Specification Bases 
changes, reference to American Society of Mechanical Engineers (ASME) 
code class components, steam generator inspection reporting, Core 
Operating Limits Report references, high radiation area controls, 
offsite dose calculation manual change control reference, and 
environmental protection related document reporting.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?

[[Page 73097]]

    No. This proposed change is to revise the administrative 
controls section of the San Onofre Units 1, 2 and 3 technical 
specifications. To the extent practicable, the San Onofre Unit 1 
technical specification Section D6, Administrative Controls, is made 
consistent with the San Onofre Units 2 and 3 technical specification 
Section 5.0, Administrative Controls. This change allows the 
handling of key administrative controls to be consistent on site. 
Certain position titles have been revised, and the cognizant Vice 
President has been included as an approver of deviations from the 
work hours and reviewer of overtime hours. The Vice President--
Business and Financial Services is identified to be responsible for 
Unit 1 decommissioning. The specification allowing the certified 
fuel handlers to implement 10 CFR 50.54(x) is removed since this is 
now included in the regulations. The qualification requirements for 
a multi-discipline supervisor consistent with the American National 
Standards Institute [ANSI] standard have been added to the staff 
qualifications section. The schedule for submitting technical 
specification Bases changes is revised to be consistent with the NRC 
approved exemption to 10 CFR 50.71(e) for submitting Updated Final 
Safety Analysis Report (UFSAR) updates. A reference to Class 1, 2, 
and 3 ASME code components is removed from the technical 
specifications and maintained in the Licensee Controlled 
Specifications (LCS) and the inservice inspection and testing 
program. The Units 2 and 3 steam generator inspection reporting 
requirements are revised to refer to the technical specification 
requirement. The Core Operating Limits Report (COLR) section is 
revised to include references to 2 topical reports related to the 
reload analysis technology transfer and the NRC's evaluation of the 
technology transfer. The sections on high radiation are revised to 
be consistent with Regulatory Guide 8.38 which provides an 
acceptable method for controlling access to high radiation areas. 
The environmental protection section of the San Onofre Unit 1 
technical specifications is revised to reflect the current status of 
the discharge system. The environmental protection sections for Unit 
1 and Units 2/3 are further revised by including a 30 day timeframe 
for providing the NRC copies of reports related to unusual or 
important environmental events and deleting the requirement to 
provide the NRC copies of proposed changes and renewal applications 
for NPDES permits.
    All of these changes are being made to provide consistency and 
flexibility in the handling of site programs, and update and clarify 
the administrative controls. There are no equipment changes or 
modifications to the plant associated with these changes that would 
affect the probability or consequences of accidents at all three 
units.
    Therefore, this change does not affect the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated?
    No. This proposed change is to revise the administrative 
controls sections of the San Onofre Units 1, 2, and 3 technical 
specifications. The changes provide consistency and flexibility in 
the handling of site programs, and update and clarify the 
administrative controls. There is no administrative change being 
made that could create a new or different accident at any of the 
three units and there is no plant or equipment modification 
associated with this change.
    Therefore, this change does not create the possibility of a new 
or different type of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety?
    No. This change revises the administrative controls sections of 
the San Onofre Units 1, 2, and 3 technical specifications. The 
changes provide consistency and flexibility in the handling of site 
programs, and update and clarify the administrative controls. There 
is no change to plant equipment associated with this change. This 
change does not affect any margin of safety.

    Therefore, this change does not involve a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chiefs: Michael Masnik (Unit 1); Stephen Dembek (Units 
2 and 3).

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: November 24, 1999 (PCN-274).
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 3.3.11, ``Post Accident Monitoring 
Instrumentation (PAMI).'' Specifically, the proposed change would 
extend the PAMI channel calibration surveillance frequency from 18 
months to 24 months to accommodate a 24-month fuel cycle. All PAMI 
instruments would then be on a 24-month calibration interval, which 
removes the need for Surveillance Requirement (SR) 3.3.11.5. Therefore, 
the licensee also proposes to delete SR 3.3.11.5.
    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?
    Response: No.
    The proposed license amendment[s] to extend the calibration 
surveillance frequency of Post Accident Monitoring Instrumentation 
(PAMI) instrumentation [are] being made to support plant operation 
with a 24-month fuel cycle.
    Increasing the calibration intervals for PAMI instrumentation to 
30 months [24 months plus the 25% surveillance interval extension 
allowed by SR 3.0.2] does not affect the initiation or probability 
of any previously analyzed accident. Increasing the calibration 
interval will not affect the integrity of any of the principal 
barriers against radiation release (fuel cladding, reactor vessel, 
and containment building). The ability of the plant to mitigate the 
consequences of any previously analyzed accidents is not adversely 
affected.
    PAMI instrumentation provides to the operators both qualitative 
and quantitative information used in accident mitigation and for the 
safe shutdown of the plant. Instrumentation which provides 
qualitative information is unaffected by a change in instrument 
accuracy induced by drift due to the increased surveillance interval 
because no explicit value is required by the Emergency Operating 
Instructions (EOIs). Instrumentation that provides quantitative 
information (i.e., decision points) in the EOIs have been evaluated. 
This evaluation resulted in no changes to any operating 
instructions. This evaluation of the proposed change to the 
surveillance interval demonstrates that licensing basis safety 
analyses acceptance criteria and San Onofre Nuclear Generating 
Station (SONGS) Units 2 and 3 EOI criteria will continue to be met.
    The proposed new surveillance frequency for these instrument 
channels was evaluated using the guidance of Generic Letter 91-04. 
The basis for the change includes a quantitative evaluation of 
instrument drift for PAMI instrumentation providing quantitative 
information to the EOIs. Also, loop accuracy/setpoint calculations 
for these instruments were updated to accommodate the extended 
surveillance period. Analyses and evaluations completed to assess 
the proposed increase in the surveillance interval demonstrate that 
the effectiveness of these instruments in fulfilling their 
respective functions is maintained. Technical Specifications Channel 
Checks and Channel Functional Checks for the subject channels, will 
continue to be performed to provide assurance of instrument channel 
OPERABILITY.
    Therefore, the proposed amendment[s do] not involve a 
significant increase in the probability or consequences of any 
previously analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    Response: No.
    The increased calibration surveillance interval for PAMI 
instrumentation is justified based on evaluation of past equipment

[[Page 73098]]

performance and does not require any plant hardware changes or 
changes in normal system operation. Changing the calibration 
interval for this instrumentation has no means of creating the 
possibility of a new or different kind of accident. There are no new 
decision points or operator responses required to support existing 
accident mitigation strategies.
    Therefore, there are no new failure modes introduced as a result 
of extending these surveillance intervals, and the proposed 
amendment[s do] not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    Response: No.
    The proposed change to the calibration surveillance interval was 
evaluated using the criteria of 95% probability/95% confidence level 
for process sensor drift.
    PAMI instrumentation are used to provide indication following 
certain hypothetical accident conditions and are used in EOIs for 
trending and to initiate operator action at certain decision points. 
Instrument uncertainty calculations have been updated for PAMI 
instrumentation used for EOI decision points as appropriate. Updated 
calculations show that the total loop uncertainty for PAMI evaluated 
either decreased or remained the same. These updated calculations 
demonstrate that applicable accuracy requirements for SONGS 2 and 3 
are satisfied with the proposed new surveillance intervals.
    Changing the calibration interval for these channels does not 
affect the margin of safety for previously analyzed accidents. 
Therefore, the proposed amendment[s do] not involve a significant 
reduction in a margin of safety.
    Based on the responses to these three criteria, Southern 
California Edison (SCE) has concluded that the proposed amendment[s 
involve] no significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment requests: December 13, 1999 (PCN-507).
    Description of amendment requests: San Onofre Nuclear Generating 
Station (SONGS) Units 2 and 3 are currently licensed for operation for 
40 years commencing with issuance of their construction permits. The 
licensee proposes to amend the SONGS Units 2 and 3 operating licenses 
to revise the expiration dates of these licenses to 40 years from the 
date of issuance of the operating licenses. Thus, these amendment 
applications request that the SONGS Unit 2 operating license expiration 
date be changed from October 18, 2013, to February 16, 2022, and the 
SONGS Unit 3 operating license expiration date be changed from October 
18, 2013, to November 15, 2022.
    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?
    Response: The proposed change does not involve any changes to 
the design or operation of the San Onofre Nuclear Generating Station 
(SONGS) 2 and 3 which may affect the probability or consequences of 
an accident evaluated in the Updated Final Safety Analysis Report 
(UFSAR). SONGS 2 and 3 were designed and constructed on the basis of 
a forty (40) year life. The accidents analyzed in the UFSAR were 
postulated on the basis of a 40 year life. No changes will be made 
that could alter the design, construction, or postulated scenarios 
regarding accident initiation and/or response. Existing 
surveillance, inspection, testing and maintenance practices and 
procedures ensure that degradation in plant equipment, structures, 
and components will be identified and corrected throughout the life 
of the plant. The effect of aging of electrical equipment, in 
accordance with 10 CFR50.49, has been incorporated into the plant 
maintenance and surveillance procedures. Therefore, the probability 
or consequences of a postulated accident previously evaluated in the 
UFSAR are not increased as a result of the proposed change.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The proposed change does not involve any changes to 
the physical structures, components, or systems of SONGS 2 and 3. 
Existing surveillance, inspection, testing, and maintenance 
practices and procedures will assure full operability for the 
plant's design lifetime of 40 years. Continued operation of SONGS 2 
and 3 in accordance with these approved procedures and practices 
will not create a new or different kind of accident.
    (3) Involve a significant reduction in a margin of safety?
    Response: There are no changes in the design, design basis, or 
operation of SONGS 2 and 3 associated with the proposed change. 
Existing surveillance, inspection, testing, and maintenance 
practices and procedures provide assurance that any degradation of 
equipment, structures, or components will be identified and 
corrected throughout the lifetime of the plant. These measures 
together with the continued operation of SONGS 2 and 3 in accordance 
with the Technical Specifications assure an adequate margin of 
safety is preserved on a continuous basis. Therefore, the proposed 
change does not result in a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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: November 30, 1999.
    Description of amendment request: The proposed amendments would 
change Technical Specification Surveillance Requirement 3.8.1.12 to 
remove the restriction which prevents performance of the diesel 
generator 24-hour run while operating in either Mode 1 or Mode 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or the consequences of a previously evaluated 
event for the following reasons:
    The primary function of the diesel generators is to supply 
emergency power to the safety-related equipment necessary to safely 
shut down the plant in case of a design basis event, such as a loss 
of coolant accident (LOCA) concurrent with a loss of offsite power 
(LOSP). The diesels are not designed to prevent such an event. 
Accordingly, the probability of a LOCA/LOSP event is not increased 
by allowing the performance of the 24-hour run with the reactor 
operating.
    It is possible that, with a diesel generator connected to its 
bus, an electrical disturbance will travel through the system and 
affect the other busses. This is most likely to happen when 
initially connecting the diesel to the bus. However, the 
surveillance procedures require that diesel generator output voltage 
be synchronized with the bus prior to the diesel output breaker 
being closed in, thus reducing the chance of an electrical 
distribution system disturbance.

[[Page 73099]]

    If a LOCA occurred concurrent with an LOSP while a diesel 
generator is connected to the bus in its 24-hour run, the diesel 
logic automatically realigns itself to the Standby mode of 
operation, allowing the diesel to supply power to the emergency bus. 
A Technical Specifications surveillance requirement tests this 
feature. Also, the proposed specification prevents the test from 
being performed unless the other two diesel generators are operable; 
this includes suspending the surveillance if one of the other 
available diesels becomes inoperable during the actual test. This 
restriction will ensure that two diesels are available to safely 
shut down the plant if necessary.
    Additionally, this amendment request does not affect any other 
system or piece of equipment necessary to prevent or mitigate the 
consequences of previously evaluated events. As a result, the 
consequences of a LOCA/LOSP event are not increased.
    2. The proposed changes do not create the possibility of an 
accident of a new or different kind from any previously evaluated 
based upon the following:
    This proposed modification to SR 3.8.1.12 does not introduce any 
new modes of operation or testing. In fact, each diesel generator is 
already connected to its respective bus during operation to satisfy 
SR 3.8.1.2, the monthly test. In the monthly test, the diesel is run 
loaded for 1 hour, connected to the grid, with the unit in 
operation. Therefore, allowing the 24 hour test to be performed for 
the diesels introduces nothing new with respect to diesel testing, 
and as a result, the possibility of a new type of event is not 
created.
    3. The change does not significantly reduce the margin of safety 
for the following reasons:
    The probability of an electrical disturbance affecting plant 
operation while connecting the diesel to the bus is minimized by the 
fact that the diesel's output voltage and phase angle are 
synchronized with those of the grid prior to being tied to the 
emergency bus. Based on engineering judgement, with the diesel 
synchronized and running connected to the grid, the likelihood of an 
electrical disturbance being transferred through the system and 
causing a plant transient is very small. Furthermore, since only one 
diesel will be tied to the bus in either Mode 1 or Mode 2, neither 
of the other two diesel generators will be affected by the 
disturbance.
    If a LOCA/LOSP occurred during the 24-hour run, the diesel 
generator's auto-logic would take the diesel out of the test mode. 
This feature is tested once per 18 months per Technical 
Specifications. With the diesel no longer in test, it would be free 
to once again tie itself to the bus. Additionally, only one diesel 
will be tied to the line during a 24-run performed with the reactor 
operating, with other diesel generators available to supply power to 
their respective emergency busses. This ensures two diesels are 
available to shut down the plant and maintain it in a safe 
condition.
    Other precautions will also be placed into plant procedures; 
specifically, the 24-hour run will not be performed on line during 
periods of severe weather or during grid instabilities.
    For the above reasons, the proposed Technical Specifications 
change will 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.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: November 18, 1999.
    Description of amendment request: The proposed amendments would 
revise technical specification surveillance requirements 4.7.7, 4.7.8, 
and 4.9.12, on the control room makeup and cleanup filtration system 
and the fuel handling building exhaust air system, from a requirement 
that laboratory analysis of charcoal filter samples meets the 
laboratory testing criteria of Regulatory Position C.6.a of Regulatory 
Guide 1.52, ``Design, Testing, and Maintenance Criteria for 
Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air 
Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power 
Plants,'' Revision 2, March 1978, to a requirement that the analysis 
meets the laboratory testing criteria of American Society for Testing 
and Materials ASTM D3803-1989, ``Standard Test Method for Nuclear-Grade 
Activated Carbon.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change revises the test protocol for Engineered 
Safety Feature charcoal filters from ASTM D3803-1979 to ASTM D3803-
1989. The change in protocol is a conservative change in that the 
revised test conditions will more accurately reflect the 
functionality of the charcoal filters under accident conditions. 
There is no change in plant configuration or components. The tests 
are conducted under laboratory conditions, so that change in 
protocol has no effect on plant operation. There is no change in how 
samples are taken to be used in analyses.
    Based on the above, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change revises the test protocol for Engineered 
Safety Feature charcoal filters from ASTM D3803-1979 to ASTM D3803-
1989. The change in protocol is a conservative change in that the 
revised test conditions will more accurately reflect the 
functionality of the charcoal filters under accident conditions. 
There is no change in plant configuration or components. The tests 
are conducted under laboratory conditions, so that change in 
protocol has no effect on plant operation. There is no change in how 
samples are taken to be used in analyses.
    Based on the above, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change revises the test protocol for Engineered 
Safety Feature charcoal filters from ASTM D3803-1979 to ASTM D3803-
1989. The change in protocol is a conservative change in that the 
revised test conditions will more accurately reflect the 
functionality of the charcoal filters under accident conditions. 
There is no change in plant configuration or components. The tests 
are conducted under laboratory conditions, so that change in 
protocol has no effect on plant operation. There is no change in how 
samples are taken to be used in analyses.

    Based on the above, the margin of safety is not significantly 
reduced by this change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: December 6, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification Definition 1.9, ``Core Alterations,'' to 
explicitly define core alterations as the movement of any fuel, 
sources, or reactivity control components within the reactor vessel 
with the vessel head removed and fuel in the vessel.

[[Page 73100]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not involve an increase in the 
probability or consequences of an accident previously evaluated. The 
proposed change does not involve any physical changes to the 
facility. The change to the definition of core alterations is 
consistent with that used in NUREG-1431, Revision 1, ``Improved 
Standard Technical Specifications for Westinghouse Plants.'' The 
proposed revision to the definition of core alterations will not 
affect the Technical Specifications Section 3/4.9, ``Refueling 
Operations'', requirements which ensure the core remains 
subcritical, nor will any Limiting Condition for Operation required 
for core alterations or the movement of fuel be changed. The 
proposed change will not affect any safety margin or safety limit 
applicable to the facility. Therefore, the proposed change does not 
involve an increase in the probability or consequences of any 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not affect any previously evaluated 
accident scenario, nor does it create any new accident scenarios. 
The proposed change is a clarifying revision to the definition of 
core alterations only, and will not alter any of the currently 
approved refueling operation activities, nor will it create any new 
refueling operation activities.
    Since the proposed change does not impact operation of the 
facility as presently approved, no possibility exists for a new or 
different kind of accident from those previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    South Texas Project Technical Specification 3/4.9.1, ``Boron 
Concentration'', ensures that the reactor will remain subcritical 
(Keff  0.95) during core alterations and that 
uniform boron concentration is maintained for reactivity control in 
the water volume having direct access with the reactor vessel. The 
proposed change in the definition of core alterations will allow 
``non-reactive'' components, such as cameras, lights, tools, movable 
incore detector thimbles, 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 negligible effect on core reactivity, 
and will not affect reactor coolant system boron concentration. 
Therefore, operations using these types of components will not 
adversely affect Keff or the shutdown margin. 
Additionally, reactor subcriticality status is continuously 
monitored in the control room during Operating Mode 6, as specified 
in Specification 3/4.9.2, ``Instrumentation''. Thus, there will be 
no reduction in a margin of safety resulting from the proposed 
change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of application for amendments: October 14, 1999 (TS 99-12).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79 
(Unit 2) by revising the Technical Specification (TS) surveillance 
requirements for steam generator tube integrity by incorporating an 
alternate repair criteria for axial primary water stress corrosion 
cracking at dented tube support plate intersections.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Operation of Sequoyah Units 1 and 2, in accordance with the 
proposed license amendment, does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Examination of crack morphology for primary water stress 
corrosion cracking (PWSCC) at dented intersections has been found to 
show one or two microcracks well aligned with only a few uncorroded 
ligaments and little or no other inside diameter axial cracking at 
the intersection. This relatively simple morphology is conducive to 
obtaining good accuracy in Non-destructive Examination (NDE) sizing 
of these indications. Accordingly, alternate repair criteria is 
established based on crack length and average and maximum depth 
within the thickness of the tube support plate (TSP) or limited 
extension outside the thickness of the TSP.
    The application of the alternate repair criteria (ARC) requires 
a condition monitoring assessment. If all indications satisfy the 
structural limits with regard to bounding lengths and average 
depths, the condition monitoring burst pressure requirements are 
satisfied.
    In addition, an operational assessment is performed to determine 
the length/depth repair bases. The crack profiles are projected to 
the end of the operating cycle for comparison with acceptance limits 
(i.e., length limit and average depth limit). Burst pressures are 
calculated from the depth profiles by searching the total crack 
length for the partial length that results in the lowest burst 
pressure. Because the burst pressure can be lower than that for the 
longest acceptable crack length at its average depth, a fixed repair 
limit is not established. The repair bases is obtained by projecting 
the crack profile to the end of the next operating cycle and 
determining if the burst pressure for the projected profile meets 
the burst pressure margin requirements defined by [Westinghouse 
Topical Report] WCAP-15128, Revision 1, dated August 1999. If the 
projected end-of-cycle (EOC) burst margin requirements are 
satisfied, the indication is left in service. Thus, the repair limit 
relative to length and average depth assures that the operational 
assessment requirements are satisfied.
    Crack length limits are established in the WCAP to assure that 
crack extension and growth outside of the TSP provides adequate 
margin against burst for the free-span crack (i.e., 3DPNO 
burst capability is maintained) in addition to the total crack 
length. A repair limit is also established in the WCAP for maximum 
depth to provide a high confidence that the indication will not 
progress through the wall at the end of an operating cycle.
    Based on the above, the proposed amendment does not result in 
any increase in the probability or consequences of an accident 
previously evaluated within the Sequoyah FSAR [Final Safety Analysis 
Report].
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed S/G [steam generator] tube ARC 
does not introduce any significant changes to the plant design 
basis. A single or multiple tube rupture event would not be expected 
in a S/G in which the plugging criteria has been applied. Both 
condition monitoring and operational assessments are completed as 
part of the implementation of ARC to determine that structural and 
leakage margin exists prior to returning S/Gs to service following 
inspections. If the condition monitoring requirements are not 
satisfied for burst or leakage, the causal factors for EOC 
indications exceeding the expected values will be evaluated. The 
methodology and application of this ARC will continue to ensure that 
tube integrity is maintained during all plant conditions consistent 
with the requirements of draft RG [Regulatory Guide] 1.121 and 
Revision 1 of RG 1.83.
    A S/G tube rupture event is one of a number of design basis 
accidents that are analyzed as part of a plant's licensing basis. In 
the analysis of a S/G tube rupture event, a bounding primary-to-
secondary leakage rate equal to the operational leakage limits in 
the TSs, plus the leak rate associated with the double ended rupture 
of a single tube, is

[[Page 73101]]

assumed. For other design basis accidents such as a main steam line 
break and loss of alternating current power, the tubes are assumed 
to retain their structural integrity and exhibit primary-to-
secondary leakage within the limits assumed in Final Safety Analysis 
Report (FSAR) accident analyses. The proposed ARC does not result in 
an accident leakage rate in excess of that assumed or calculated in 
SQN's current accident analyses.
    Even under severe accident conditions, the potential for 
significant leakage would be expected to be small and not 
significantly different than for other degradation mechanisms 
repaired to 40 percent depth limits. It is concluded that 
application of the proposed ARC for PWSCC at dented TSP locations 
results in a negligible difference from current 40-percent repair 
limits.
    TVA continues to implement a maximum operating condition leak 
rate limit of 150 gallons per day (0.1 gallons per minute) per S/G 
to preclude the potential for excessive leakage during all plant 
conditions.
    The possibility of a new or different kind of accident from any 
previously evaluated is not created because S/G tube integrity is 
maintained by inservice inspection and effective primary-to-
secondary leakage monitoring.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Tube repair limits provide reasonable assurance that tubes 
accepted for continued service without plugging or repair will 
exhibit adequate tube structural and leakage integrity during 
subsequent plant operation. The implementation of the proposed ARC 
is demonstrated to maintain S/G tube integrity consistent with the 
criteria of draft NRC Regulatory Guide 1.121. The guidelines of RG 
1.121 describe a method acceptable to the NRC staff for meeting 
General Design Criteria (GDC) 2, 4, 14, 15, 31, and 32 by ensuring 
the probability or the consequences of S/G tube rupture remain 
within acceptable limits. This is accomplished by determining the 
limiting conditions of degradation of S/G tubing, for which tubes 
with unacceptable cracking should be removed from service.
    Upon implementation of the proposed ARC, even under the worst-
case conditions, the occurrence of PWSCC at the tube support plate 
elevations is not expected to lead to a S/G rupture event during 
normal or faulted plant conditions. All tubes are shown to retain 
the margins of safety against burst consistent with the safety 
factor margins implicit in the stress limit criteria of Section III 
of the American Society of Mechanical Engineers [ASME] Code, for all 
service loading conditions. In addition, all tubes have been shown 
to retain a margin of safety against gross failure or burst 
consistent with the stress limits of [Paragraph] NB-3225 of Section 
III of the ASME Code under postulated accident conditions concurrent 
with a safe shutdown earthquake.
    In addressing the combined effects of loss-of-coolant accident 
plus safe shutdown earthquake on the S/G component (as required by 
GDC 2), it has been determined that tube collapse will not occur in 
the Sequoyah S/Gs. This analysis is discussed in WCAP 13990, dated 
May 1994. No tubes are excluded from the application of the proposed 
ARC.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to the plant safety analyses as defined in the 
FSAR or TSs.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: November 8, 1999.
    Brief description of amendments: The proposed amendments would 
change Technical Specification 5.5.11, ``Ventilation Filter Testing 
Program (VFTP)'' to include the requirement for laboratory testing of 
Engineered Safety Feature (ESF) Ventilation System charcoal samples per 
American Society for Testing and Materials (ASTM) D3803-1989 and the 
application of a safety factor of 2.0 to the charcoal filter efficiency 
assumed in the plant design-basis dose analyses.
    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) Do the proposed changes involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes only involve the laboratory testing 
methodology performed on activated charcoal to help determine 
whether the charcoal in the filtration units can remain in place or 
[if it] require[s] replacement.
    Generic Letter 99-02 intends to standardize the way nuclear-
grade activated charcoal is tested throughout the industry in order 
to provide conservative filtration results as well as uniform and 
repeatable tests. The purpose is to ensure the filtration systems 
protect the Operators in the Control Room (GDC [General Design 
Criterion] 19) as well as the public (10CFR100), in the event of a 
radiological accident scenario.
    The charcoal adsorber sample laboratory testing per ASTM D3803-
1989 is more stringent than the current testing practice and more 
accurately demonstrates the required performance of the adsorbers 
following a design ba[s]is LOCA [loss of coolant accident]. No 
Licensing Basis Accidents or mitigation capability will be affected 
by incorporation of these changes.
    Therefore, this change will not result in a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    (2) Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    Plant procedures are only altered to the extent that the revised 
specification will allow different reference standards for testing 
activated charcoal. These changes ensure continued support of the 
safety related ESF filtration equipment and do not affect their 
failure or failure modes.
    Therefore, this change will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    (3) Do the proposed changes involve a significant reduction in a 
margin of safety?
    None of the changes being proposed alter the environmental 
conditions maintained in the areas supported by the ESF filtration 
systems during normal operations and following an accident. Also 
these changes will not cause an increase in radiological releases 
through the Primary Plant Ventilation Exhaust System. As a result, 
the margin of safety for these functions remains the same.

    Therefore, 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Section Chief: Robert A. Gramm.

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

    Date of application request: December 3, 1999 (ULNRC-04158).
    Description of amendment request: The proposed amendment requested 
changes to Section 5.6.6, ``Reactor Coolant System (RCS) Pressure and 
Temperature Limits Report (PTLR),'' of the improved Technical 
Specifications (ITS) that were issued on May 28, 1999, in Amendment No. 
133. The current Technical Specifications (CTS) remain in effect until 
the ITS are implemented on or before April 30, 2000. The proposed 
changes to the ITS would approve the use of the PTLR by the licensee to 
make changes to the plant pressure temperature limits and low 
temperature overpressure protection

[[Page 73102]]

limits without prior NRC staff approval in accordance with Generic 
Letter 96-03, ``Relocation of the Pressure Temperature Limit Curves and 
Low Temperature Overpressure Protection System Limits,'' dated January 
31, 1996. The proposed changes are: (1) Add the word criticality to ITS 
Subsection 5.6.6.a as one of the reactor conditions for which RCS 
pressure and temperature limits will be determined, (2) add the phrase 
``and COMS PORV,'' where COMS PORV stands for cold overpressure 
mitigation system power operated relief valve, to the the introductory 
paragraph of ITS subsection 5.6.6.b to show that the analytical methods 
listed in the subsection are also for the COMS PORV, and (3) replace 
the two documents listed in ITS subsection 5.6.6.b by the reference to 
the future NRC letter that approves the use of the PTLR and the 
Westinghouse Topical Report, WCAP-14040-NP-A, Revision 2, ``Methodology 
Used to Develop Cold Overpressure Mitigating System Setpoints and RCS 
Heatup and Cooldown Limit Curves,'' dated January 1996, that provides 
the methodology that will be used by the licensee in using the PTLR 
report. The current plant pressure temperature limits and low 
temperature overpressure protection limits are in the CTS and were 
approved in Amendment No. 124, which was issued April 2, 1998.
    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 submits the PTLR, which contains the 
relocated CTS heatup and cooldown, and COMS PORV limits and the 
methodology used to calculate them, and the added references into 
ITS 5.6.6. The proposed change is administrative in nature since it 
is a movement of information from the CTS to a licensee controlled 
document, and has prior NRC staff approval. The PTLR contains the 
limit curves and the ITS requires more restrictive actions to be 
taken when the limiting conditions for operation are not met than is 
currently required by the CTS. The heatup and cooldown, and COMS 
PORV limits within the PTLR will be implemented and controlled per 
Callaway Plant programs and procedures and changes to the PTLR will 
be performed per requirements of 10 CFR 50.59 to ensure that change 
to these limits in the future will 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.
    As stated earlier, the movement of the heatup and cooldown, and 
COMS PORV limits from the CTS to the PTLR has no influence or 
impact, nor does it contribute in any way to the probability or 
consequences of an accident. No safety-related equipment, safety 
function, or plant operations will be altered as a result of this 
proposed change. The proposed change is administrative in nature 
since it is a movement of requirements from the CTS to a licensee 
controlled document, the PTLR, and the change adds references into 
the ITS incorporating the licensee controlled document. Therefore, 
the possibility of a new or different kind of accident from any 
accident previously evaluated is not created.
    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 
an analyzed event. The margin of safety presently provided by the 
CTS remains unchanged. 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 protective functions. 
Therefore, the proposed change is administrative in nature and does 
not impact the operation of Callaway Plant in a manner that involves 
a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: November 5, 1999, as supplemented on 
December 3, 1999.
    Description of amendment request: This proposed change revises the 
applicability for the reactor power distribution limits and the Average 
Power Range Monitor (APRM) gain adjustments. The applicability is 
proposed to be revised to operation at 25% Rated Thermal 
Power (RTP).
    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 Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change does not involve an increase in the 
probability or consequences of an accident previously evaluated 
because the revisions standardize and make consistent the 
applicability and actions for the reactor power distribution limits 
in the current Technical Specifications. Since reactor operation 
with these revised Specifications is fundamentally unchanged, no 
design or analytical acceptance criteria will be exceeded. As such, 
this change does not impact initiators of analyzed events or assumed 
mitigation of accident or transient events. The structural and 
functional integrity of plant systems is unaffected. Therefore, the 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change does not affect any parameters or conditions 
that could contribute to the initiation of any accident. No new 
accident modes are created. No safety-related equipment or safety 
functions are altered as a result of these changes. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment will not involve a 
significant reduction in a margin of safety.
    At thermal power levels < 25% RTP, the reactor is operating with 
substantial margin to the reactor power distribution limits [and 
this margin is unchanged]. The proposed change does not impact 
operation at power levels  25% RTP and has no effect on 
any safety analysis assumption or initial condition. Thus, the 
margin of safety required for safety analyses [is] maintained. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: James W. Clifford.

[[Page 73103]]

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

    Date of amendment request: November 15, 1999 (TSCR 202).
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TSs) in order to extend the 
required frequency of the control rod exercise test (TS 15.4.1, Table 
15.4.1-2, Item 10) from the current frequency of every 2 weeks to 
quarterly.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of an accident previously evaluated.
    Relaxing the frequency of performance for a surveillance does 
not result in any hardware changes, nor does it significantly 
increase the probability of occurrence for initiation of any 
analyzed events since the function of the equipment has remained 
unchanged. The proposed frequency has been determined to be adequate 
based on industry operating data as supported by the conclusions 
reached in NUREG 1366 and NRC GL [Generic Letter] 93-05.
    Surveillance tests are intended to provide assurance of 
continued component operability. The frequency of performance of a 
surveillance does not significantly increase the consequences of an 
accident, as a change in frequency does not change the response of 
the equipment in performing its specified function (i.e. the overall 
functional capabilities of the rod control system will not be 
modified). Increasing the interval of control rod exercise testing 
will reduce the possibility of inadvertent testing related [to] 
reactor trips and dropped rods, and resulting in fewer challenges to 
safety systems and resultant plant transients.
    This change does not involve a significant increase in the 
consequences of an accident or event previously evaluated because 
the source term, containment isolation or radiological releases are 
not being changed by the proposed revision. Existing system and 
component redundancy and operation is not being changed by the 
proposed change. The assumptions used in evaluating the radiological 
consequences in the PBNP Final Safety Analysis Report are not 
invalidated. Therefore, this change does not affect the consequences 
of previously evaluated accidents.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This change does not introduce nor increase the number of 
failure mechanisms of a new or different type of accident than those 
previously evaluated since there are no physical changes being made 
to the facility. The design and design basis of the facility remain 
unchanged. The plant safety analyses remain unchanged. All equipment 
important to safety will continue to operate as designed. Component 
integrity is not challenged. The changes do not result in any event 
previously deemed incredible being made credible. The changes do not 
result in more adverse conditions nor result in any increase in 
challenges to safety systems. Therefore, operation of the Point 
Beach Nuclear Plant in accordance with the proposed amendment will 
not create the possibility of a new or different type of accident 
from any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not involve a significant reduction in 
a margin of safety.
    The proposed change does not involve a significant reduction in 
the margin of safety because existing component redundancy is not 
being changed by this proposed change. There are no changes to 
initial conditions contributing to accident severity or 
consequences. The proposed surveillance frequency, as supported by 
past test results, continues to provide the required assurance of 
operability, such that safety margins established through the design 
and facility license, including the Technical Specifications, remain 
unchanged. Therefore, there are no significant reductions in a 
margin of safety introduced by this proposed amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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 notice was previously published as a separate 
individual notice. The notice content was the same as above. It was 
published as an individual notice either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. It is 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.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: October 20, 1999.
    Brief description of amendment: The amendment changed the footnote 
to the Improved Technical Specifications associated with the Design 
Features Fuel Storage Specification 4.3.1.1.b which required that 2300 
ppm boron be maintained in the Spent Fuel Pool.
    Date of publication of individual notice in Federal Register: 
November 19, 1999 (64 FR 63346).
    Expiration date of individual notice: December 20, 1999.

Notice of Issuance of Amendments to Facility Operating Licenses

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

[[Page 73104]]

Street, NW., Washington, DC, and electronically from the ADAMS Public 
Library component on the NRC Web site, http://www.nrc.gov (the 
Electronic Reading Room).

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 application for amendments: September 14, 1999.
    Brief description of amendments: The amendments approve the 
administrative changes to PVNGS TS 5.5.2, Primary Coolant Sources 
Outside Containment, to delete the references to the post-accident 
sampling return piping of the radioactive waste gas system and the 
liquid radwaste system, and TS 5.6.2, Annual Radiological Environmental 
Operating Report, to delete the administrative requirement to include 
in the report certain TLD [thermoluminescence dosimeter] results that 
are no longer available.
    Date of issuance: November 24, 1999.
    Effective date: November 24, 1999, to be implemented within 60 
days.
    Amendment Nos.: Unit 1--122, Unit 2--121, Unit 3--121.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56528).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 24, 1999.
    No significant hazards consideration comments received: No.

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 application for amendment: October 21, 1999.
    Brief description of amendment: This amendment revises Technical 
Specifications (TS) for the Shearon Harris Nuclear Power Plant by 
implementing selected improvements described in NRC Generic Letter (GL) 
93-05, ``Line-Item Technical Specifications To Reduce Surveillance 
Requirements For Testing During Power Operation,'' dated September 27, 
1993.
    Date of issuance: December 17, 1999.
    Effective date: December 17, 1999.
    Amendment No: 93.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 17, 1999 (64 
FR 62705).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 17, 1999.
    No significant hazards consideration comments received: No.

CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz 
Mill, Pennsylvania

    Date of application for amendment: September 15, 1999, as 
supplemented on October 4, 1999.
    Brief description of amendment: This amendment changes the 
decommissioning Technical Specifications dealing with controls for 
ingress, egress, and equipment removal from containment.
    Date of issuance: December 7, 1999.
    Effective Date: December 7, 1999.
    Amendment No: 11.
    Facility License No. TR-2: This amendment changes the 
decommissioning Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59798).
    The Commission has issued a Safety Evaluation for this amendment 
dated December 7, 1999.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Inc., Docket No. 50-003, 
Indian Point Nuclear Generating Station, Unit 1, Buchanan, New York

    Date of application for amendment: July 20, 1999.
    Brief description of amendment: The amendment would revise the 
Technical Specifications to change the senior license requirements for 
the Operations Manager.
    Date of issuance: December 15, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No: 46.
    Facility Operating License No. DPR-5: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 2, 1999 (64 
FR 49027).
    The July 20, 1999, letter providing clarifying information that did 
not change the scope of the original application and proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated December 15, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 5, 1999, as supplemented 
June 22 and July 30, 1999.
    Brief description of amendments: These amendments conform the 
licenses to reflect the transfer of Operating Licenses Nos. DPR-66 and 
NPF-73 for the Beaver Valley Power Station Unit Nos. 1 and 2, to the 
extent held by Duquesne Light Company (DLC) to the Pennsylvania Power 
Company, and the operating authority under the licenses from DLC to 
FirstEnergy Nuclear Operating Company as previously approved by an 
Order dated September 30, 1999.
    Date of issuance: December 3, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 226 and 104.
    Facility Operating License Nos. DPR-66 and NPF-73: These amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: June 14, 1999 (64 FR 
31880).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 30, 1999. The June 22 and July 
30, 1999, supplements were within the scope of the initial application 
as originally noticed.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of application for amendment: September 14, 1999.
    Brief description of amendment: This amendment eliminates License 
Condition 2.C.10 of the Operating License regarding controls over the 
containment air locks during plant outages and modifies License 
Condition 2.F of the Operating License regarding reporting requirements 
for violations of the Technical Specifications and the Environmental 
Protection Plan.
    Date of issuance: December 15, 1999.
    Effective date: December 15, 1999.
    Amendment No.: 109.
    Facility Operating License No. NPF-58: This amendment revised the 
Operating License.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59803).
    The Commission's related evaluation of the amendment is contained 
in a

[[Page 73105]]

Safety Evaluation dated December 15, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 29, 1999, as supplemented 
August 27, October 29, and November 3, 1999.
    Brief description of amendment: The amendment clarifies the 
authority to possess certain types of radioactive materials and 
components at either Unit 1 or Unit 2. Following the transfer of the 
Three Mile Island, Unit 1 (TMI-1), operating license to AmerGen, these 
items, under the amendment, may continue to be moved between the TMI-1 
and TMI-2 units as they currently are.
    Date of issuance: December 9, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 217.
    Facility Operating License No. DPR-50: Amendment revised the 
License.
    Date of initial notice in Federal Register: July 12, 1999 (64 FR 
37572). The August 27, October 29, and November 3, 1999, letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the amendment beyond the scope of the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 9, 1999.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, (TMI-2) Middletown, Pennsylvania

    Date of application for amendment: June 29, 1999, as supplemented 
by letters dated August 27, October 29, and November 3, 1999.
    Brief description of amendment: The amendment adds a provision to 
the license conditions to ensure that the storage of certain types of 
radioactive materials and components at Three Mile Island (TMI), Unit 
2, pursuant to the TMI, Unit 1 license, does not result in a source 
term that would exceed the limits in the TMI, Unit 2 Post-Defueling 
Monitored Storage Safety Analysis Report.
    Date of issuance: December 14, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 53.
    Facility Operating License No. DPR-73: Amendment revised the 
License.
    Date of initial notice in Federal Register: July 12, 1999 (64 FR 
37572). The August 27, October 29, and November 3, 1999, supplements 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination or expand 
the amendment beyond the scope of the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 14, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 23, 1999, as supplemented 
July 30, August 9, August 20, October 7, and October 11, 1999.
    Brief description of amendment: The amendment replaces references 
to Illinois Power Company in the Operating License with references to 
AmerGen Energy Company, LLC, to reflect the transfer of the license as 
approved by an Order dated November 24, 1999.
    Date of issuance: December 15, 1999.
    Effective date: December 15, 1999.
    Amendment No.: 123.
    Facility Operating License No. NPF-62: The amendment revised the 
Operating License.
    Date of initial notice in Federal Register: August 19, 1999 (64 FR 
45290).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 24, 1999.
    Comments received: Yes. Comments received from The Environmental 
Law and Policy Center of the Midwest were addressed in the staff's 
safety evaluation.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: September 23, 1999, as 
supplemented October 11 and November 10, 1999.
    Brief description of amendments: The amendments provide approval to 
move steam generator sections through the auxiliary building and to 
disengage crane travel interlocks, and provide relief from performance 
of Technical Specification Surveillance Requirement 4.9.7.1. The loads 
to be moved are in support of the Unit 1 Steam Generator Replacement 
Project.
    Date of issuance: December 7, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 233 and 216.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1999 (64 FR 
57665). The October 11, 1999, submittal provided corrected TS pages. 
The November 10, 1999, submittal was in response to a NRC request for 
additional information dated October 26, 1999, and provided clarifying 
information to the original submittal. This information was within the 
scope of the original Federal Register notice and did not change the 
staff's initial proposed no significant hazards considerations 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 7, 1999.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: October 1, 1999, as 
supplemented November 19, 1999.
    Brief description of amendments: The amendments involve the 
resolution of an unreviewed safety question related to certain small-
break loss-of-coolant accident scenarios for which there may not be 
sufficient containment recirculation sump water inventory to support 
continued operation of the emergency core cooling system and 
containment spray system pumps during and following switchover to cold 
leg recirculation. Resolution of this issue consists of a combination 
of physical plant modifications, new analyses of containment 
recirculation sump inventory, and resultant changes to the accident 
analyses to ensure sufficient water inventory in the containment 
recirculation sump. The amendments would also change the Technical 
Specifications dealing with the refueling water storage tank inventory 
and temperature, the required amount of ice in each ice basket in the 
containment, and the delay to start the containment air recirculation/
hydrogen skimmer fans.
    Date of issuance: December 13, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 234 and 217.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.

[[Page 73106]]

    Date of initial notice in Federal Register: October 29, 1999 (64 FR 
58458).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 13, 1999.
    No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: February 12, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specifications to (1) allow reactor vessel hydrostatic and leakage 
tests when reactor coolant temperature is above 212 deg.F without 
maintaining primary containment integrity and (2) establish a limit and 
a surveillance requirement on reactor coolant activity when reactor 
coolant temperature is above 212 deg.F, the reactor is not critical, 
and primary containment has not been established.
    Date of issuance: November 24, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 107.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 24, 1999 (64 FR 
14283).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 24, 1999.
    No significant hazards consideration comments received: No.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of application for amendment: September 30, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specification surveillance periodicity requirements for the control 
room emergency filtration system.
    Date of issuance: December 8, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 108.
    Facility Operating License No. DPR-22. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59805).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 8, 1999.
    No significant hazards consideration comments received: No.

PECO Energy Company, Public Service Electric and Gas Company,Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket No. 
50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County, 
Pennsylvania

    Date of application for amendment: March 1, 1999, as supplemented 
June 14, October 1 and October 6, 1999.
    Brief description of amendment: The amendment supports the 
installation of a digital Power Range Neutron Monitoring system and the 
incorporation of the long-term thermal-hydraulic stability solution 
hardware.
    Date of issuance: October 14, 1999.
    Effective date: Effective as of date of issuance and shall be 
implemented prior to restart from the Peach Bottom Atomic Power 
Station, Unit 3, October 1999 refueling outage.
    Amendment No.: 234.
    Facility Operating License No. DPR-56: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29711). The June 14, October 1 and October 6, 1999, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated October 14, 1999.
    No significant hazards consideration comments received: No.

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: April 6, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specifications by removing the words ``three individual underground'' 
and ``underground'' from the limiting conditions for operation when 
referring to the emergency diesel generator fuel oil storage tanks in 
Sections 3.7.A.5 and 3.7.F.4.
    Date of issuance: December 7, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 198.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29713).
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 7, 1999.
    No significant hazards consideration comments received: No.

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

    Dates of amendments request: March 12, 1998, as supplemented by 
letters of April 24, 1998, August 20, 1998, November 20, 1998, February 
3, 1999, February 20, 1999, April 30, 1999 (two letters), May 28, 1999, 
June 30, 1999, July 27, 1999, August 19, 1999, August 30, 1999, 
September 15, 1999, and September 23, 1999.
    Brief description of amendments: The amendments fully convert SNC's 
Current TS (CTS) to Improved TS (ITS) based on NUREG-1431, ``Standard 
Technical Specifications, Westinghouse Plants,'' Revision 1, of April 
1995. The amendments add two new Additional Conditions to Appendix C of 
the Unit 1 and Unit 2 Facility Operating Licenses. The first new 
Additional Condition authorizes SNC to relocate certain CTS 
requirements to SNC-controlled documents. The second new condition 
addresses the schedule for performing new and revised ITS 
surveillances.
    Date of issuance: November 30, 1999.
    Effective date: As of the date of issuance and shall be implemented 
no later than March 31, 2000.
    Amendment Nos.: 146 and 137.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments fully 
convert SNC's CTS to ITS.
    Dates of initial notices in Federal Register: May 25, 1999 (64 FR 
28218) and August 25, 1999 (64 FR 46443). The supplemental letters 
dated April 24, 1998, August 20, 1998, November 20, 1998, February 3, 
1999, February 20, 1999, April 30, 1999 (two letters), May 28, 1999, 
June 30, 1999, July 27, 1999, August 19, 1999, August 30, 1999, 
September 15, 1999, and September 23, 1999, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determinations.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 30, 1999.
    No significant hazards consideration comments received: No.

[[Page 73107]]

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: April 28, 1999.
    Brief description of amendments: The amendments revised Vogtle's 
operating licenses to allow the licensee to establish containment 
hydrogen monitoring within 90 minutes of initiation of a safety 
injection following a loss-of-coolant accident, compared to the current 
30 minute requirement.
    Date of issuance: December 8, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 110 and 88.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43779).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 8, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 28, 1998, as supplemented by 
letters dated May 31 and October 21 (2 letters), 1999.
    Brief description of amendments: The amendments authorize the 
revision of the South Texas Project updated final safety analysis 
report (UFSAR) to allow the use of operator action to reduce the steam 
generator power-operated relief valve setpoint consistent with the 
revised small-break loss-of-coolant accident analysis for the 
replacement Delta 94 SGs.
    Date of issuance: December 14, 1999.
    Effective date: December 14, 1999. Revisions will be incorporated 
into the next UFSAR update in accordance with the schedule in 10 CFR 
50.71(e).
    Amendment Nos.: Unit 1--119, Unit 2--107.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
authorize revision of the UFSAR.
    Date of initial notice in Federal Register: September 9, 1998 (63 
FR 48268).
    The May 31 and October 21 (2 letters), 1999, supplements provided 
additional clarifying information. One of the October 21, 1999, 
supplements also provided a revised UFSAR pages. This information was 
within the scope of the original application and Federal Register 
notice and did not change the staff's initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 14, 1999.
    No significant hazards consideration comments received: No.

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: September 30, 1999.
    Description of amendment request: The amendments revise the 
operating licenses to remove license conditions that have become 
outdated, are no longer applicable, or are redundant, and to 
consolidate license conditions which currently exist in two locations 
in each units license.
    Date of issuance: December 16, 1999.
    Effective date: December 16, 1999.
    Amendment Nos.: 237, 262, and 222.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68: 
Amendments revised the licenses.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59807).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 16, 1999.
    No significant hazards consideration comments received: No.

TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: February 27, 1998, as supplemented by 
letters dated June 10, 1998, and October 22, 1999.
    Brief description of amendments: The amendments change the 
refueling water storage tank (RWST) low-low level setpoints in 
Technical Specification Table 3.3.2-1, ``Engineered Safety Feature 
Actuation System Instrumentation,'' to increase the volume of water 
available to containment spray pumps when the containment spray system 
switches to the recirculation mode of operation.
    Date of issuance: December 8, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 73 and 73.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 15, 1998 (63 FR 
38205). The October 22, 1999, supplement provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the 
application beyond the scope described in the initial notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 8, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 18, 1999.
    Brief description of amendment: The amendment revises the reactor 
core spiral reloading pattern such that it begins around a source range 
monitor. The offloading pattern is the reverse sequence.
    Date of Issuance: December 14, 1999.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 181.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 8, 1999 (64 
FR 48867).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated December 14, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: September 23, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) by deleting the test requirements for 
snubbers from the TSs. These requirements are already included in the 
Point Beach Nuclear Plant In-Service Inspection Program.
    Date of issuance: December 6, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 191 and 196.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.

[[Page 73108]]

    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71977).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 6, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 29, 1998, as supplemented by 
letters dated July 29 and October 21, 1999.
    Brief description of amendment: The amendment revised (1) the 
reactor coolant system (RCS) heatup and cooldown limit curves in 
Figures 3.4-2 and 3.4-3 and cold overpressure mitigation system power-
operated relief valve setpoint limit curve in Figure 3.4-4 of the 
current TSs, and (2) the list of references in Section 5.6.6 on the RCS 
pressure temperature limits report (PTLR) in the improved TSs. The 
improved TSs were issued in Amendment No. 123, dated March 31, 1999, to 
replace the current TSs, but have not yet been implemented. The 
revision to Section 5.6.6 of the improved TSs replaced the previous 
references to NRC documents giving criteria for the above limit curves 
in the current TSs by the references to (1) the NRC letter of December 
2, 1999, that approved the use of the PTLR of Generic Letter 96-03, 
``Relocation of the Pressure Temperature Limit Curves and Low 
Temperature Overpressure Protection System Limits,'' dated January 31, 
1996, for WCGS, and (2) WCAP-14040-NP-A, ``Methodology Used to Develop 
Cold Overpressure Mitigation System Setpoints and RCS Heatup and 
Cooldown Limit Curves.'' The PTLR will provide the methodology for the 
licensee to revise the heatup and cooldown and setpoint limit curves 
for WCGS in the future without prior staff approval, after the improved 
TSs are implemented and have replaced the current TSs. The improved TSs 
are to be implemented by December 31, 1999.
    Date of issuance: December 7, 1999.
    Effective date: December 7, 1999, to be implemented by December 31, 
1999.
    Amendment No.: 130.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9023) and September 8, 1999 (64 FR 48869). The October 21, 1999, 
supplemental letter provided additional clarifying information, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 7, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 8, 1999.
    Brief description of amendment: The amendment corrects 15 errors in 
the improved Technical Specifications that was issued in Amendment No. 
123 on March 31, 1999. In addition, four corrections to Table LG, 
``Details Relocated from Current Technical Specifications [CTS],'' that 
was attached to the safety evaluation dated March 31, 1999, issued with 
Amendment No. 123 were made.
    Date of issuance: December 16, 1999.
    Effective date: December 16, 1999, to be implemented December 31, 
1999.
    Amendment No.: 131.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 16, 1999 (64 
FR 62231).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 16, 1999.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 8th day of December 1999.

    For the Nuclear Regulatory Commission.
Suzanne C. Black,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-33684 Filed 12-28-99; 8:45 am]
BILLING CODE 7590-01-P