[Federal Register Volume 64, Number 240 (Wednesday, December 15, 1999)]
[Notices]
[Pages 70077-70098]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-32311]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 20, 1999, through December 3, 1999. 
The last biweekly notice was published on December 1, 1999 (64 FR 
67330).

Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing

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

[[Page 70078]]

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 14, 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, 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 No. 50-317, Calvert Cliffs 
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland

    Date of amendment request: November 18, 1999.
    Description of amendment request: The proposed amendment revises 
the Unit 1 Heatup Curve (Technical Specification Figure 3.4.3-1), Unit 
1 Cooldown Curve (Technical Specification Figure 3.4.3-2), and Unit 1 
Maximum Power-Operated Relief Valve (PORV) Opening Pressure vs 
Temperature Curve (Technical Specification Figure 3.4.12-1) to change 
fluence level from 2.61 x 10\19\ n/cm \2\ to 4.49 x 1019 n/cm \2\ 
(E>1MeV). This change reflects the new actual fluence level for which 
these curves are valid, and is necessary to extend the

[[Page 70079]]

applicability of the curves for Unit 1 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:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    In accordance with 10 CFR Part 50, Appendix G, the Calvert 
Cliffs pressure/temperature (P-T) limits for material fracture 
toughness requirements of the reactor coolant pressure boundary 
materials were developed using the methods of linear elastic 
fracture mechanics and the guidance found in the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section 
III, Appendix G. The Calvert Cliffs (P-T) limits are based on 
fluence level. The fluence level corresponds to the pressurized 
thermal shock (PTS) screening criteria defined in 10 CFR 50.61 for 
the critical elements. Methods described in the Nuclear Regulatory 
Commission Regulatory Guide 1.99, Revision 2, are used to predict 
the embrittlement effect of neutron irradiation on reactor vessel 
materials. Regulatory Guide 1.99 defines embrittlement effect in 
terms of adjusted reference temperatures (ART), which depends on the 
material property of the PTS critical element.
    The proposed higher fluence level for the Technical 
Specification P-T limits was made possible by the identification of 
a new 10 CFR 50.61 critical element for fracture toughness 
requirements for protection against PTS events. The material 
properties of the new critical element resulted in an increase in 
fluence level from 2.61 x 10 \19\ n/cm \2\ to 4.49 x 1019 n/cm \2\ 
for the ART valves calculated using the material properties of the 
old PTS critical element. the P-T limits analysis remain well within 
the conservative acceptance limits of the ASME Boiler and Pressure 
Vessel Code, Section III, Appendix G. Hence, with the new higher 
fluence level, the 10 CFR Part 50, Appendix G, requirement for 
adequate margin to brittle failure during normal operation, 
anticipated operational occurrences, and system hydrostatic tests, 
for the reactor coolant pressure boundary materials, is maintained.
    Therefore 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 accidents previously evaluated.
    The implementation of the proposed revision has no significant 
effect on either the configuration of the plant, or the manner in 
which it is operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    As discussed above, the P-T limits analysis remain well within 
the conservative acceptance limits of the ASME Boiler and Pressure 
Vessel Code, Section III, Appendix G. Hence, with the new higher 
fluence level, the 10 CFR Part 50, Appendix G, requirement for 
adequate margin to brittle failure during normal operation, 
anticipated operational occurrences, and system hydrostatic tests, 
for the reactor coolant pressure boundary materials, is maintained.
    Therefore, this proposed modification 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 
amendment 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 Acting Section Chief: Victor Nerses.

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 19, 1999.
    Description of amendments request: The amendments request approval 
of changes in the Updated Final Safety Analysis Report (UFSAR) that 
constitute an unreviewed safety question (USQ) as described in 10 CFR 
50.59. Specifically, these changes would be an increase in the 
probability of occurrence of malfunction. Additionally, these changes 
were not previously evaluated in the UFSAR.
    Regulations require that structures, systems, and components 
important to safety be appropriately protected against the effects of 
effects of missiles that might result from equipment failures. Failures 
that could occur in the large turbines of the main turbine-generator 
sets have the potential for producing large high-energy missiles 
(hereinafter called ``turbine missiles''). Both of Baltimore Gas and 
Electric Company's (BGE) turbine generator suppliers studied the 
failure of the rotating elements of their turbine-generators. The UFSAR 
only addresses a turbine missile hitting the Containment Building, 
Control Room, Switchgear Room, and Waste Processing Area. As a result 
of revising the Unit 1 and Unit 2 turbine missile analysis, BGE 
determined that the discussion of turbine missiles in Section 5.3.1 of 
the UFSAR was incomplete. Specifically, it did not discuss the 
probability of a missile from the Unit 1 turbine-generator striking: 1) 
the refueling water tanks; 2) the No. 11 Fuel Oil Storage Tank; or 3) 
plant equipment through various roof slabs or through non-missile-proof 
openings in the missile-proof walls. When these additional targets are 
included, the total target area is increased. If the target area 
increases, the probability of a turbine missile causing equipment 
damage increases. It is this increase in probability that leads to a 
USQ for a turbine missile from Unit 1. Note that by using methodologies 
previously approved by NRC, the revised analysis concludes there is no 
USQ for turbine missiles from the Unit 2 turbine-generator.
    The UFSAR change is considered a USQ for Units 1 and 2 because the 
results of the revised Unit 1 turbine missile analysis for the 
following unprotected rooms or components show an increase in 
probability of occurrence of malfunction not previously evaluated in 
the UFSAR:
    the Refueling Water Tanks;
    the No. 11 Fuel Oil Storage Tank (non-missile-proof);
    the saltwater pumps through roof hatches in the Intake Structure 
roof;
    the roof slabs over the refueling Water Tank Pump Room, the Control 
Room Heating, Ventilation, and Air Conditioning (HVAC) Equipment Room, 
the Spent Fuel Pool Area Ventilation Equipment room, and a portion of 
118 level roof over the fuel cask handling area;
    the Control Room HVAC Room through its non-missile-proof door; and
    the Unit 1 Auxiliary Building 45 Switchgear Room through 
the its non-missile-proof doors.
    The probability of a missile from the Unit 1 turbine-generator 
striking them is a negligible increase in the probability of occurrence 
of malfunction of equipment associated with Unit 1 and 2. Upon approval 
of this request, the UFSAR will be revised to reflect the proposed 
turbine missile description. There is no USQ associated with the Unit 2 
turbine-generator.
    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.
    Regulations require that structures, systems, and components 
important to safety be appropriately protected against the effects 
of missiles that might result from equipment failures. Further that 
could occur in the large turbines of the main turbine-generator sets 
have the potential for producing large high-

[[Page 70080]]

energy missiles (hereinafter called turbine missiles). Both of our 
turbine-generator suppliers studied the failure of the rotating 
elements of their turbine-generators. The UFSAR only addresses 
turbine missile hitting the Containment Building, Control Room, 
Switchgear Room, and Waste Processing Area. As result of revising 
the Unit 1 and Unit 2 turbine missile analysis, we determined that 
the discussion of turbine missiles of the UFSAR was incomplete. From 
the revised analysis, we determined Unit 1 and 2 USQs exist for the 
following unprotected rooms or components (i.e., there is an 
increase in probability of occurrence of malfunction not previously 
evaluated in the UFSAR):
    the Refueling Water Tanks;
    the No. 11 Fuel Oil Storage Tank;
    the Saltwater Pumps through roof hatches in the Intake Structure 
Roof;
    the roof slabs over the Refueling Water Tank Pump Room, the 
Control Room Heating, Ventilation, and Air Conditioning (HVAC) 
Equipment Room, Spent Fuel Pool Area Ventilation Equipment Room, and 
a portion of 118' level roof over the cask handling area;
    the Control Room HVAC Room through its non-missile-proof door; 
and,
    the Unit 1 Auxiliary Building 45' Switchgear Room through its 
non-missile-proof doors.
    The probability of a missile from the Unit 1 turbine-generator 
striking them is a negligible, but greater than zero, increase in 
the probability of occurrence of malfunction of equipment associated 
with Units 1 and 2.
    For Unit 1 High Trajectory Missiles (HTM), the guidance of NUREG 
0800, Standard Review Plan, is used as one acceptable method for 
evaluating the risk. Use of this method is not a commitment to the 
Standard Review Plan and does not incorporate the Standard Review 
Plan into our licensing basis. The revised analysis shows that the 
total target area considered vulnerable to an HTM is less than the 
Standard Review Plan limit of 10,000 ft2 for each unit. 
Therefore, the risk form an HTM is insignificant. Note that all of 
the Units 1, 2, and Common structures listed above are equally 
vulnerable to a Unit 1 HTM. Therefore, any risk increase to the 
plant structures constitutes a USQ for Units 1 and 2.
    For Unit 1 Low-Trajectory Missiles (LTMs), protection for the 
Auxiliary Building is provided by a 3' thick, concrete, missile-
proof wall between the Turbine Building and the Auxiliary building 
(the K-line wall). This wall is 3' thick below the 69' elevation and 
2' thick above the 69' for areas protecting safety-related 
equipment. The revised analysis evaluates the protection of Unit 1 
equipment from a Unit 1 LTM. The 69' Control Room HVAC Equipment 
Room and Unit 1 Auxiliary Building 45' Switchgear Room are protected 
by the missile-proof walls except for the openings at the non-
missile-proof doors. A turbine missile that hits one of these doors 
is assumed to go through them, strike safety-related equipment in 
the room, and cause it to fail. Recall that the Control Room HVAC 
equipment is shared by both units. Therefore, any increase in risk 
of failure of equipment in this room affects both Units 1 and 2.
    The risk associated with a turbine missile to either of these 
doors is calculated using guidance in Regulatory Guide 1.115, 
Revision 1, ``Protection Against Low-Trajectory Turbine Missiles.'' 
This guidance states that the turbine missile hazard should be less 
than 107. The missile hazard rate in the revised risk 
analysis shows that the risk from LTMs from the Unit 1 General 
Electric turbine-generator to the 69' Control Room HVAC Equipment 
Room and Unit 1 Auxiliary Building 45' Switchgear room through these 
non-missile-proof doors is less than 107.
    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 makes no physical changes to the plant. 
Specifically, the proposed change does not add new or modify 
existing plant equipment such that it could become an accident 
initiator different from its current role as an accident initiator. 
The only change made by this activity is the revision of the UFSAR 
to include the revised turbine missile analysis. The UFSAR chapter 1 
drawings correctly depict the location of plant structures and 
components, including the thickness of and the openings in the 
missile-proof wall between the Turbine Building and the Auxiliary 
building (the K-Line Wall). Therefore, the possibility of a new or 
different type of accident is not created by the proposed change.
    3. Would not involve a significant reduction in a margin of 
safety.
    The regulations require an evaluation of turbine missiles to 
ensure that structures, systems, and components important to safety 
be appropriately protected from them. Revised turbine missile 
analysis have been performed consistent with appropriate regulatory 
guidance (Regulatory Guide 1.115 and the Standard Review Plan). The 
results of the revised analysis meet the acceptance criteria of the 
guidance. 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 Acting Section Chief: Victor Nerses.

Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick 
Steam Electric Plant, Unit 1, Brunswick County, North Carolina

    Date of amendment request: November 17, 1999.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 2.1.1.2, ``Reactor Core Safety 
Limits.'' The minimum critical power ratios (MCPR) for single and two 
recirculation loop operation would be increased. In addition, the 
reference in TS 5.6.5, ``Core Operating Limits Report,'' Item b.5, 
would be removed.
    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 amendments do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    The proposed license amendment will establish MCPR Safety Limit 
values of 1.10 for two recirculation loop operation and 1.11 for single 
recirculation loop operation. Additionally, the proposed license 
amendment replaces an expiring cycle-specific reference in the list of 
analytical methods approved for determining core operating limits in 
Specification 5.6.5.b with a reference to a GE [General Electric] 
topical report which has been accepted by the NRC.
    The methods for calculating the MCPR Safety Limit values have been 
previously approved by the NRC and are described in GE's reload 
licensing methodology topical report NEDE-24011-P-A. Use of these 
methods ensures that the integrity of the fuel will be maintained 
during normal operation and that the resulting MCPR Safety Limit values 
satisfy the fuel design safety criteria that less than 0.1 percent of 
the fuel rods experience boiling transition if the safety limits are 
not violated. The change does not require any physical plant 
modifications, physically affect any plant components, or allow the 
plant to be operated any closer to fuel design limits. Therefore, the 
proposed change to the MCPR Safety Limit values and to the list in 
Specification 5.6.5.b of analytical methods approved for determining 
core operating limits results no increase in the probability of a 
previously evaluated accident.
    The consequences of a previously evaluated accident are dependent 
on the initial conditions assumed for the analysis, the behavior of the 
fuel during the accident, the availability and successful functioning 
of the equipment assumed to operate in response to the accident, and 
the setpoints at which these actions are initiated.

[[Page 70081]]

    The methods used for calculating the MCPR Safety Limits have been 
approved by the NRC and are described in GE's reload licensing 
methodology topical report NEDE-24011, ``General Electric Standard 
Application for Reactor Fuel (GESTAR II).'' The proposed MCPR Safety 
Limit values of 1.10 for two recirculation loop operation and 1.11 for 
single recirculation loop operation will ensure that less than 0.1 
percent of the fuel rods will experience boiling transition during any 
plant operation if the limits are not violated. The proposed change to 
the MCPR Safety Limit values does not affect the performance of any 
equipment used to mitigate the consequences of a previously evaluated 
accident. Also, the proposed change does not affect setpoints that 
initiate protective or mitigative actions. No analysis assumptions are 
violated and there are no adverse effects on the factors contributing 
to offsite and onsite dose.
    Based on the determination of the proposed MCPR Safety Limit values 
using conservative NRC-approved methods and the operability of plant 
systems designed to mitigate the consequences of accidents not being 
changed, the proposed change to the MCPR Safety Limit values and to the 
list in Specification 5.6.5.b of analytical methods approved for 
determining core operating limits does not significantly increase the 
consequences of a previously evaluated accident.
    2. The proposed license amendments will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation. This proposed license amendment does 
not involve any physical alteration of plant systems and plant 
equipment will not be operated in a different manner. As a result, 
no new failure modes are being introduced. Therefore, the proposed 
change to the MCPR Safety Limit values and to the list in 
Specification 5.6.5.b of analytical methods approved for determining 
core operating limits will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.

    The margin of safety is established through the design of the 
plant structures, systems, and components; through the parameters 
within which the plant is operated; through the establishment of 
setpoints for actuation of equipment relied upon to respond to an 
event; and through margins contained within the safety analyses.
    The proposed change to the MCPR Safety Limit values and the list 
in Specification 5.6.5.b of analytical methods approved for 
determining core operating limits does not adversely impact the 
performance of plant structures, systems, components, and setpoints 
relied upon to respond to mitigate an accident. As previously 
stated, the methods for calculating the MCPR Safety Limit values 
have been previously approved by the NRC and are described in GE's 
reload licensing methodology topical report NEDE-24011-P-A. Use of 
these methods ensures that the resulting MCPR Safety Limit values 
satisfy the fuel design safety criteria that less than 0.1 percent 
of the fuel rods experience boiling transition if the safety limits 
are not violated. As a result, the proposed changes do not 
significantly impact any safety analysis assumptions or results. 
Based on the assurance that the fuel design safety criteria will be 
met, the proposed changes do not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: November 19, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) for the Harris Nuclear Plant 
(HNP) to incorporate American Society for Testing and Materials (ASTM) 
D3803-1989, ``Standard Test Method for Nuclear-Grade Activated 
Carbon,'' as the standard for testing nuclear-grade activated charcoal. 
Specifically, TS 4.7.6 will be revised for the Control Room Emergency 
Filtration System, TS 4.7.7 will be revised for the Reactor Auxiliary 
Building Emergency Exhaust System, and TS 4.9.12 will be revised for 
the Fuel Handling Building Emergency Exhaust System. These changes are 
being proposed in accordance with NRC Generic Letter (GL) 99-02, 
``Laboratory Testing Of Nuclear-Grade Activated Charcoal,'' dated June 
3, 1999.
    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.
    This proposed change to revise the standard to which activated 
charcoal samples are tested will ensure that testing is accurate and 
repeatable. This will help ensure that the Engineered Safety Feature 
(ESF) ventilation systems are capable of performing their safety 
function. Therefore, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes incorporate ASTM D3803-1989 as the testing 
standard for nuclear-grade activated charcoal samples. This will 
ensure that testing is accurate and repeatable. Plant structures, 
systems, and components will not be operated in a different manner 
as a result of these proposed changes and no physical modifications 
to equipment are involved. Using the improved testing protocol does 
not have the potential for creating the possibility of a new or 
different type of accident from any previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes do not change the manner in which 
structures, systems or components are operated. Revising the 
standard to which activated charcoal samples are tested will ensure 
that testing is accurate and repeatable. This will help ensure that 
the ESF ventilation systems are capable of performing their safety 
function. Therefore, the proposed changes do not involve a reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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.

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

    Date of amendment request: November 12, 1999.
    Description of amendment request: The proposed change revises the 
pressure-temperature limits by revising

[[Page 70082]]

the heatup, cooldown and inservice test limitations for the Reactor 
Pressure Vessel to a maximum of 32 Effective Full Power Years.
    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 of occurrence or consequences of an accident previously 
evaluated?
    The proposed changes do not modify the reactor coolant pressure 
boundary, do not make changes in operating pressure, materials or 
seismic loading. The proposed changes adjust the reference 
temperature for the limiting beltline material to account for 
radiation effects and provide the same level of protection as 
previously evaluated. The proposed changes do not adversely affect 
the integrity of the reactor coolant system (RCS) such that its 
function in the control of radiological consequences is affected. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not create the possibility of a new or 
different kind of accident previously evaluated for Quad Cities 
Nuclear Power Station. No new modes of operation are introduced by 
the proposed changes. The proposed changes will not create any 
failure mode not bounded by previously evaluated accidents. Use of 
the revised pressure-temperature (P-T) curves will continue to 
provide the same level of protection as was previously reviewed and 
approved.
    Further, the proposed changes to the P-T curves do not affect 
any activities or equipment, and are not assumed in any safety 
analysis to initiate any accident sequence for Quad Cities Nuclear 
Power Station. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes reflect an update of the P-T curves to 
extend the Reactor Pressure Vessel (RPV) operating limit to 32 
Effective Full Power Years (EFPY). The revised curves are based on 
the latest American Society of Mechanical Engineers (ASME) guidance 
and actual operational data for the units. This proposed changes are 
acceptable because the ASME guidance maintains the relative margin 
of safety commensurate with that which existed at the time that the 
ASME Section IX Appendix G was approved in 1974. Therefore, the 
proposed changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: November 16, 1999.
    Description of amendment request: The proposed change modifies the 
surveillance requirements for Functional Unit 3 on Table 4.1.A-1 due to 
replacement of the Reactor Pressure Vessel Steam Dome pressure switches 
with analog trip units.
    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:

    Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    During the upcoming refueling outages at Quad Cities Nuclear 
Power Station, Unit 1 and Unit 2, a design change will be 
implemented that upgrades the existing Reactor Vessel Steam Dome-
High instrumentation from a pressure switch to an analog trip unit 
device. Analog trip units are proven technology that are more 
reliable than existing equipment. Analog trip units are used in 
various applications of Quad Cities Nuclear Power Station, including 
the Reactor Protection System (RPS) low water level trip function.
    The proposed change adds a CHANNEL CHECK and 31-day trip unit 
calibration requirement for the Reactor Vessel Steam Dome Pressure--
High RPS trip function. This requirement is not applicable to the 
existing instrumentation because the Barksdale pressure switches are 
non-indicating and do not employ trip units.
    Technical Specification (TS) requirements that govern 
operability or routine testing of plant instruments are not assumed 
to be initiators of any analyzed event because these instruments are 
intended to prevent, detect, or mitigate accidents. Therefore, these 
changes will not involve an increase in the probability of 
occurrence of an accident previously evaluated. Additionally, these 
changes will not increase the consequences of an accident previously 
evaluated because the proposed change does not adversely impact 
structures, systems, or components (SSCs). The planned instrument 
upgrade is a more reliable design than existing equipment. The 
proposed change establishes requirements that ensures components are 
operable when necessary for the prevention or mitigation of 
accidents or transients. Furthermore, there will be no change in the 
types or significant increase in the amounts of any effluents 
released offsite. For these reasons, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes support a planned instrumentation upgrade 
by incorporating Surveillance Requirements required to ensure 
operability. The change does not adversely impact the manner in 
which the instrument will operate under normal and abnormal 
operating conditions. Therefore, these changes provide an equivalent 
level of safety and will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The changes in methods governing normal plant operation are 
consistent with the current safety analysis assumptions. Therefore, 
these changes will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed change supports a planned instrumentation upgrade. 
The proposed change does not affect the probability of failure or 
availability of the affected instrumentation. The addition of a 
CHANNEL CHECK and 31-day trip unit calibration for RPS Functional 
Unit 3 (Reactor Vessel Steam Dome Pressure--High) is a conservative 
change that aligns the surveillance requirements for a planned 
instrumentation upgrade with that of similar instrumentation. 
Therefore, it is concluded that the proposed changes will not result 
in a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: October 13, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.6.1, Table 3.3.6-1, ``Primary 
Containment

[[Page 70083]]

Isolation Instrumentation.'' This amendment requests that Function 5 on 
Table 3.3.6-1, ``RHR SDC System Isolation,'' be modified by removing 
footnote (d). Footnote (d) states, ``Only the inboard trip system is 
required in Modes 1, 2, and 3, as applicable, when the outboard valve 
control is transferred to the alternate remote shutdown panel and the 
outboard valve is closed.'' The outboard suction valve, RHR-V-8, is no 
longer used as a high/low pressure interface in the residual heat 
removal (RHR) system. Valve RHR-V-9, which is in series with valve RHR-
V-8, is now used as the high/low pressure interface valve. Valve RHR-V-
9 is operable in all modes of operation and therefore, footnote (d) is 
no longer needed. The current footnote (e) will be relettered as 
footnote (d) for consistency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This change involves the probability and consequences of 
accidents associated with the isolation of the RHR SDC [shutdown 
cooling] mode of RHR operation. Isolation is provided if high 
temperatures occur in RHR pump rooms or heat exchanger areas, if 
reactor vessel water level is low, or if reactor vessel pressure is 
high.
    FSAR [Final Safety Analysis Report] Chapter 15, ``Accident 
Analysis,'' describes two events associated with the RHR system 
during SDC operation. FSAR Section 15.1.6, ``Inadvertent Residual 
Heat Removal Shutdown Cooling Operation,'' describes the impact of 
system operation during startup or cool-down when the reactor is 
near critical. The proposed change removes the exemption for the 
second trip system to isolate RHR SDC operation. There will be no 
change in the probability or consequences of this accident as a 
result of the proposed change.
    The second accident is described in FSAR Section 15.2.9, 
``Failure of Residual Heat Removal Shutdown Cooling.'' It postulates 
the failure of the RHR system to function in SDC mode. The 
evaluation assumes a failure of the SDC mode of operation but does 
not disable the remaining modes of RHR operation. The alternate SDC 
paths involve the use of the safety relief valves to establish a 
cooling flow path to the containment suppression pool. That 
evaluated accident does not result in any fuel failure. The proposed 
change will not result in an increase in the probability of fuel 
failures. The evaluated accident does result in normal coolant 
activity being released to the suppression pool through the safety 
relief valves. The proposed activity will not result in a change in 
the release of this coolant activity. The proposed change requires 
the removal of the exemption for the second trip system to isolate 
SDC and will have no impact on the probability or consequences of 
that accident.
    Therefore, the 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 cause any new inadvertent SDC 
startup, loss of water inventory or loss of coolant accidents 
(LOCA). New or different inadvertent RHR SDC startup accidents are 
not possible because this change is only a further restriction on 
system operation. The LOCA during Mode 3 is bounded by the LOCA 
defined for Modes 1 and 2. No new primary system LOCA can be 
initiated because of this 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 removal of an exemption for the second trip system, as 
proposed by this change, will increase the probability that leaks 
and high pressure will be isolated. Therefore, operation of WNP-2 in 
accordance with the proposed amendment will not decrease the margin 
of safety. Therefore, the 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, NW, Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana

    Date of amendment request: October 25, 1999.
    Description of amendment request: The proposed license amendment 
would revise the reactor pressure vessel (RPV) surveillance capsule 
withdrawal schedule for the River Bend Station. The first surveillance 
capsule would be withdrawn at 13.4 effective full power years (EFPY) 
rather than 10.4 EFPY.
    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.
    Pressure-temperature (P/T) limits (RBS Technical Specifications 
Figure 3.4.11-1) are imposed on the reactor coolant system to ensure 
that adequate safety margins against nonductile or rapidly 
propagating failure exist during normal operation, anticipated 
operational occurrences, and system hydrostatic tests. The P/T 
limits are related to the nil-ductility reference temperature, 
RTNDT, as described in ASME [American Society of 
Mechanical Engineers] Section III, Appendix G. Changes in the 
fracture toughness properties of RPV beltline materials, resulting 
from the neutron irradiation and the thermal environment, are 
monitored by a surveillance program in compliance with the 
requirements of 10 CFR [Part] 50, Appendix H. The effect of neutron 
fluence on the shift in the nil-ductility reference temperature of 
pressure vessel steel is predicted by methods given in RG 
[Regulatory Guide] 1.99, [Revision] 2.
    River Bend's current P/T limits, as well as those for the 
planned increase in reactor thermal power (``Power Uprate''), were 
established based on adjusted reference temperatures developed in 
accordance with the procedures prescribed in RG 1.99, [Revision] 2, 
Regulatory Position 1. Calculation of adjusted reference temperature 
by these procedures includes a margin term to ensure conservative, 
upper-bound values are used for the calculation of the P/T limits. 
Revision of the first capsule withdrawal schedule will not affect 
the P/T limits because they will continue to be established in 
accordance with Regulatory Position 1 or other NRC [Nuclear 
Regulatory Commission]-approved procedures. When permitted (two or 
more credible surveillance data sets available), Regulatory Position 
2 (or other NRC-approved) methods for determining adjusted reference 
temperature will be followed.
    This change is not related to any accidents previously 
evaluated. The proposed change is a revision of the first 
surveillance capsule withdrawal time, identified in TRM [Technical 
Requirements Manual] Table 3.4.11-1, from 10.4 EFPY to 13.4 EFPY. 
This change will not affect P/T limits as given in RBS Technical 
Specifications Figure 3.4.11-1 or USAR Figures 5.3-4a and 5.3-4b. 
This change will not affect any plant safety limits or limiting 
conditions of operation. The proposed change will not affect reactor 
pressure vessel performance as no physical changes are involved and 
RBS vessel P/T limits will remain conservative in accordance with RG 
1.99, [Revision] 2 requirements. The proposed change will not cause 
the reactor pressure vessel or interfacing systems to be operated 
outside of their design or testing limits. Also, the proposed change 
will not alter any assumptions previously made in evaluating the 
radiological consequences of accidents. Therefore, the probability 
or

[[Page 70084]]

consequences of accidents previously evaluated will not be increased 
by the proposed change.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the first RPV material surveillance 
capsule withdrawal time in TRM Table 3.4.11-1 from 10.4 EFPY to 13.4 
EFPY. This proposed change does not involve a modification of the 
design of plant structures, systems, or components. The proposed 
change will not impact the manner in which the plant is operated as 
plant operating and testing procedures will not be affected by the 
change. The proposed change will not degrade the reliability of 
structures, systems, or components important to safety (ITS) as 
equipment protection features will not be deleted or modified, 
equipment redundancy or independence will not be reduced, supporting 
system performance will not be downgraded, the frequency of 
operation of ITS equipment will not be increased, and increased or 
more severe testing of ITS equipment will not be imposed. No new 
accident types or failure modes will be introduced as a result of 
the proposed change. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from that 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    As stated in Section 5.3.2 of the River Bend Safety Evaluation 
Report (NUREG-0989), ``Appendices G and H of 10 CFR [Part] 50 
describe the conditions that require pressure-temperature limits and 
provide the general bases for these limits. These appendices 
specifically require that pressure-temperature limits must provide 
safety margins at least as great as those commended in the ASME Code 
[American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code], Section III, Appendix G. * * * Until the results from the 
reactor vessel surveillance program become available, the staff will 
use Regulatory Guide (RG) 1.99, Revision 1 [now Revision 2], to 
predict the amount of neutron irradiation damage.* * * The use of 
operating limits based on these criteria--as defined by applicable 
regulations, codes, and standards--will provide reasonable assurance 
that nonductile or rapidly propagating failure will not occur, and 
will constitute an acceptable basis for satisfying the applicable 
requirements of General Design Criteria (GDC) 31.''
    Bases for RBS Technical Specification 3.4.11 states: ``The P/T 
limits are not derived from Design Basis Accident (DBA) analyses. 
They are prescribed during normal operation to avoid encountering 
pressure, temperature, and temperature rate of change conditions 
that might cause undetected flaws to propagate and cause nonductile 
failure of the RCPB [reactor coolant pressure boundary], a condition 
that is unanalyzed. * * * Since the P/T limits are not derived from 
any DBA, there are no acceptance limits related to the P/T limits. 
Rather, the P/T limits are acceptance limits themselves since they 
preclude operation in an unanalyzed condition.''
    The proposed change will not affect any safety limits, limiting 
safety system settings, or limiting conditions of operation. The 
proposed change does not represent a change in initial conditions, 
or in a system response time, or in any other parameter affecting 
the course of an accident analysis supporting the Bases of any 
Technical Specification. The proposed change does not involve 
revision of the P/T limits but rather a revision of the withdrawal 
time for the first surveillance capsule. The current P/T limits (and 
proposed P/T limits for Power Uprate) were established based on 
adjusted reference temperatures for vessel beltline materials 
calculated in accordance with Regulatory Position 1 of RG 1.99, 
[Revision] 2. P/T limits will continue to be revised as necessary 
for changes in adjusted reference temperature due to changes in 
fluence according to Regulatory Position 1 until two or more 
credible surveillance data sets become available. When two or more 
credible surveillance data sets become available, P/T limits will be 
revised as prescribed by Regulatory Position 2 of RG 1.99, 
[Revision] 2, or other NRC-approved guidance. Therefore, the 
proposed changes do not involve a significant reduction in any 
margins 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: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: October 29, 1999.
    Description of amendment request: The proposed license amendment 
would change the River Bend Station (RBS) Updated Safety Analysis 
Report (USAR), Sections 6.2 and 15.6, to incorporate a revision to the 
calculation of radiological doses following a loss-of-coolant-accident 
(LOCA). The LOCA dose calculation was revised as a result of (1) an 
increase in the calculated positive pressure period (PPP) to account 
for a new phenomenon identified in Information Notice (IN) 88-76, (2) a 
more conservative Suppression Pool water volume value, (3) an 
additional and more conservative liquid leakage term identified in IN 
91-56, and (4) changes to the engineered safety features (ESF) systems 
liquid leakage term.
    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 significantly increase the 
probability or consequences of an accident previously evaluated.
    The analysis changes described by this proposed change to the 
USAR are not initiators to events, and therefore do not involve the 
probability of an accident. These modifications reflect a revision 
to the post-LOCA dose calculation. USAR Section 15.6.5.1.1 states 
that ``There are no realistic, identifiable events which would 
result in a pipe break inside of containment of the magnitude 
required to cause an accident LOCA * * * However, since such an 
accident provides an upper limit estimate to the resultant effects 
for this category of pipe breaks, it is evaluated without the causes 
being identified.'' The analysis itself does not identify an 
initiator, nor is it the initiator, of a LOCA. There was no physical 
change to the plant. The increase to the positive pressure period 
(PPP) was the result of inclusion of phenomena not previously 
included in the analysis documented in the SAR [safety analysis 
report], and does not have any impact on accident probability. The 
inclusion of an NRC [Nuclear Regulatory Commission] Information 
Notice (IN) 91-56 unfiltered liquid leakage term is voluntary and 
conservative in nature and does not represent an additional failure 
that could be construed as an initiator to the event. Therefore, 
this change does not increase the probability of occurrence of an 
accident evaluated previously in the safety analysis report (SAR).
    This proposed change to the USAR does increase the consequences 
of an accident, but the increase is not significant. While the 
calculated off-site and control room doses of a LOCA did increase in 
Revision 1 to the post-LOCA dose calculation (reference 1) [of 
Attachment 1 to the License Amendment request, dated October 29, 
1999], the dose consequences remain below the regulatory limits of 
10 CFR [Part] 100 and 10 CFR [Part] 50, Appendix A, General Design 
Criteria (GDC) 19 as approved per NUREG-0989 and License Amendment 
98. This change first accounts for the potential effect that 
differential temperature has on the PPP assumed in the off-site dose 
analysis. It also conservatively includes an additional liquid 
leakage term to account for concerns documented in NRC IN 91-56. 
Neither of these changes has an appreciable effect on vital area 
access doses. Vital area access dose calculations were not revised 
since they still conservatively reflect the expected doses discussed 
in USAR Section 12.3.2.4. There is no impact on equipment 
qualification associated with the proposed change since other gross 
conservatisms exist in those calculations (e.g., not crediting 
suppression pool scrubbing) compared to the post-LOCA dose 
calculations. Reanalysis of the off-site dose calculation 
demonstrates that the revised doses are increased only slightly and 
remain significantly less than the regulatory

[[Page 70085]]

limits. With the IN 91-56 term excluded, the increases are within 
the criteria of less than 10 [percent] of the remaining margin, 
which is the criteria to be applied in the revised 10 CFR 50.59 rule 
for minimal increases in consequences. With the IN 91-56 term 
included, only the 30 day LPZ [low-population zone] thyroid dose 
exceeds the ``minimal increase'' criterion. Note the doses 
documented in Table 1 [of Attachment 1 to the License Amendment 
request, dated October 29, 1999], above, are less than the values 
which had been documented in the SAR prior to the implementation and 
NRC approval of TS [Technical Specifications] Amendment 98. 
Therefore, this change does not significantly increase the 
consequences of an accident previously evaluated in the SAR.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any [previously] analyzed.
    This change does not represent a physical change to the plant. 
It does not involve initiators to any events in the SAR, nor does 
the activity create the possibility for any new accidents. Rather, 
this change is a result of the evaluation of the most limiting LOCA 
which can occur at River Bend. Therefore, this change involves no 
new system interactions and does not create the possibility of an 
accident of a different type than those presently evaluated in the 
SAR.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The off-site dose consequences are calculated in accordance with 
regulatory guidance found in Regulatory Guide 1.3 and the SRP 
[Standard Review Plan], consistent with the analyses submitted to 
and approved by the NRC in support of Technical Specification 
Amendment 98. It is conservatively assumed that 100 [percent] fuel 
failure occurs instantaneously upon a recirculation pipe break, thus 
2 of the 3 fission product barriers are immediately eliminated. 
These assumptions are made without any causes for the failures being 
identified. Containment is assumed to leak at its maximum allowable 
leakage rate (0.26 [percent] per day) for the duration of the event. 
Other leakage terms, such as engineered safety feature (ESF) 
leakage, are assumed to be equal to the Technical Specification 
limit. Since assumptions are made in accordance with Technical 
Specification allowable values and regulatory guidance, this change 
does not reduce the margin of safety as defined in the basis for any 
RBS Technical Specification.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: July 29, 1998, as supplemented by 
letters dated July 29, October 28, and November 11, 1999.
    Description of amendment request: The amendment will revise 
Technical Specification 6.9.1.11.1 by replacing the existing reference 
to the Asea Brown Boveri-Combustion Engineering, Inc. (ABB CE), small 
break loss-of-coolant (SBLOCA) accident emergency core cooling system 
(ECCS) performance evaluation model with the revised model described in 
the topical report CENPD-137, Supplement 2, P-A, April 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. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The SBLOCA ECCS performance evaluation is conducted to 
demonstrate conformance of light water nuclear power reactors to the 
ECCS acceptance criteria of 10 CFR 50.46. The proposed change is 
associated with an analysis performed using the new Supplement 2 
version of the ABB CE SBLOCA Model (S2M). The primary objective of 
the analysis using the new model was to determine the impact of a 
reduction in High Pressure Safety Injection (HPSI) pump flow rate 
due to increased surveillance test measurement uncertainty. NRC 
approval of the new S2M model for use in licensing applications of 
CE design pressurized water reactors was obtained on December 16, 
1997 (Reference 1) [of license amendment request dated July 29, 
1998].
    A comparison of the Waterford 3 results for the limiting SBLOCA 
scenario using the new S2M model against the criteria of 10 CFR 
50.46(b) is summarized below:

------------------------------------------------------------------------
           Parameter                    Result             Criterion
------------------------------------------------------------------------
Peak Cladding Temperature......  1929 deg.F.........  2200 deg.F
Maximum Cladding Oxidation.....  8.09%..............  17%
Core-wide Cladding Oxidation...  <0.58%.............  1%
Coolable Geometry Maintained...  Yes................  Yes
------------------------------------------------------------------------

    These results remain within the criteria of 10 CFR 50.46. Thus, 
application of the new S2M model to the ECCS at Waterford will not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    The proposed change will not create any new system connections 
or interactions. Thus, no new modes of failure are introduced. The 
revised methods used in the new SBLOCA evaluation model and their 
impact has been reviewed and approved by the NRC (Reference 1) [of 
license amendment request dated July 29, 1998]. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change does not alter the ability of the ECCS to 
maintain compliance with 10 CFR 50.46 criteria. The revised methods 
used in the new SBLOCA evaluation model and their impact has been 
reviewed and approved by the NRC (Reference 1) [of license amendment 
request dated July 29, 1998]. Therefore, the proposed change will 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn, 
1400 L Street NW., 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: September 7, 1999.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 3/4.3.2.1, ``Safety 
Features

[[Page 70086]]

Actuation System Instrumentation,'' Table 3.3-4, ``Safety Features 
Actuation System Instrumentation Trip Setpoints,'' to remove the ``Trip 
Setpoint'' values and modify the ``Allowable Values'' for Containment 
Pressure-High and Containment Pressure-High-High, and would change TS 
3/4.3.2, ``Reactor Protection System and Safety System 
Instrumentation,'' to reflect the above change.
    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 (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit No. 1, in accordance with these changes 
would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because the proposed changes do not 
change any accident initiator, initiating condition, or assumption.
    The proposed changes would revise Technical Specification (TS) 
Table 3.3-4, Safety Features Actuation System Instrumentation Trip 
Setpoints, to administratively remove from TS the ``Trip Setpoint'' 
values for Instrument String Functional Unit ``b'', Containment 
Pressure--High, and Functional Unit ``c'', Containment Pressure--
High-High, and also modify the TS ``Allowable Values'' entry for 
these same Functional Units, consistent with updated calculations 
using current setpoint methodology. The Trip Setpoint values removed 
from TS will be maintained in DBNPS-controlled documents. The 
proposed changes to Limiting Condition for Operation (LCO) 3.3.2.1 
and Bases 3/4.3.1 and 3/4.3.2 are associated with these changes.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
invalidate assumptions used in evaluating the radiological 
consequences of an accident, do not alter the source term or 
containment isolation, and do not provide a new radiation release 
path or alter radiological consequences.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not introduce a new or different accident initiator or 
introduce a new or different equipment failure mode or mechanism.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes establish an error analysis that has 
been shown to adequately preserve 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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: November 2, 1999.
    Description of amendment request: The proposed amendment would: (1) 
relocate the Boric Acid Addition Tank System (BAAS) and Borated Water 
Storage Tank requirements of Technical Specification (TS) 3/4.1.2.8, 
Reactivity Control Systems--Borated Water Sources--Shutdown, in their 
entirety to the Davis-Besse Nuclear Power Station Updated Safety 
Analysis Report (USAR) Technical Requirements Manual (TRM); (2) 
relocate the BAAS requirements of TS 3/4.1.2.9, Reactivity Control 
Systems--Borated Water Sources--Operating, to the USAR TRM, except for 
portions applicable to the BWST which are proposed to be deleted 
because they are redundant to the existing provisions of TS 3/4.5.4, 
Emergency Core Cooling Systems--Borated Water Storage Tank; (3) modify 
TS 3/4.1.2.1, Reactivity Control Systems--Borated Water Sources--
Shutdown, by deleting references to TS 3.1.2.8; (4) incorporate 
corresponding changes to the TS index; and (5) incorporate 
corresponding changes to the TS Bases.
    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 proposed changes 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. No previously analyzed accident scenario is 
changed, and initiating conditions remain as previously analyzed.
    The proposed changes would relocate the Boric Acid Addition 
System (BAAS) and Borated Water Storage Tank (BWST) requirements of 
Technical Specification (TS) 3/4.1.2.8 in their entirety to the 
Davis-Besse Nuclear Power Station (DBNPS) Updated Safety Analysis 
Report (USAR) Technical Requirements Manual (TRM). The proposed 
changes would also relocate the BAAS requirements of TS 3/4.1.2.9 to 
the USAR TRM. The portions of TS 3/4.1.2.9 applicable to the BWST 
are proposed to be deleted because they are completely redundant to 
the existing provisions of TS 3/4.5.4, Emergency Core Cooling 
Systems--Borated Water Storage Tank. Associated with these changes, 
TS 3/4.1.2.1 is proposed to be revised to delete references to TS 
3.1.2.8. The appropriate changes to the TS Index are also proposed, 
as well as changes to TS Bases     3/4.1.2. The proposed changes are 
also consistent with the improved ``Standard Technical 
Specifications--Babcock and Wilcox Plants,'' NUREG-1430, Revision 1.
    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 proposed changes do 
not alter the source term, containment isolation or allowable 
radiological releases.
    The chemical addition system, which includes the BAAS, is not 
credited for mitigation of any USAR Chapter 6 or Chapter 15 
accidents. The BWST is credited for mitigation of USAR Chapter 6 and 
Chapter 15 accidents, as part of the Emergency Core Cooling System 
(ECCS). However, the BWST's requirements concerning ECCS are 
provided in separate TS 3/4.5.4, that is not proposed for change.
    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, and no new or 
different failure modes have been defined for any plant system or 
component important to safety. No new or different types of failures 
or accident initiators are introduced by the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes are administrative in nature, 
consisting of deletion and/or relocation of certain TS requirements 
into licensee-controlled documents, and have no bearing on the 
margin of safety which exists in the present TS or USAR.

    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.

[[Page 70087]]

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

    Date of amendment request: November 2, 1999.
    Description of amendment request: The proposed amendment would: (1) 
modify Technical Specification (TS)     3/4.3.2.1, Safety Features 
Actuation System Instrumentation, Table 3.3-4, Safety Features 
Actuation System Instrumentation Trip Setpoints, to remove ``Trip 
Setpoint'' values for Instrument String Functional Unit ``f,'' Borated 
Water Storage Tank (BWST) Level; (2) modify TS 3/4.3.2.1, Table 3.3-4, 
Functional Unit ``f,'' Allowable Values, to make it consistent with 
updated calculations using current setpoint methodology; (3) modify 
Limiting Condition for Operation (LCO) 3.3.2.1, Safety Features 
Actuation System Instrumentation to reflect removal of the ``Trip 
Setpoint'' for this Functional Unit; (4) change the footnote associated 
with TS 3/4.3.2.1, Table 3.3-4, Functional Unit ``f,'' Allowable 
Values, to indicate that the Allowable Values apply to the Channel 
Functional Test and no longer applies to the Channel Calibration; (5) 
modify TS     3/4.1.2.9, Reactivity Control Systems--Borated Water 
Sources--Operating, and TS 3/4.5.4, Emergency Core Cooling Systems--
Borated Water Storage Tank, to increase the minimum BWST water level; 
and (6) make corresponding changes to TS Bases 3/4.1.2, Boration 
Systems, 3/4.3.1 and 3/4.3.2, Reactor Protection System and Safety 
System Instrumentation, and 3/4.5.4, Borated Water Storage Tank.
    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 proposed changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because the proposed changes do not 
change any accident initiator, initiating condition, or assumption.
    The proposed changes would revise Technical Specification (TS) 
Table 3.3.4, Safety Features Actuation System Instrumentation Trip 
Setpoints, to administratively remove from the TS the ``Trip 
Setpoint'' values for Instrument String Functional Unit ``f,'' 
Borated Water Storage Tank (BWST) Level, and also modify the TS 
``Allowable values entry for this same Functional Unit, consistent 
with updated calculations using current setpoint methodology. The 
Trip Setpoint values removed from the TS will be maintained in 
Davis-Besse Nuclear Power Station (DBNPS)-controlled documents. The 
proposed changes to Limiting Condition for Operation (LCO) 3.3.2.1 
and Bases 3/4.3.1 and 3/4.3.2 are associated with these changes.
    Associated with the above changes, TS     3/4.1.2.9 and TS 3/
4.5.4 are proposed to be revised to increase the minimum available 
BWST borated water volume requirement as specified in LCO 
3.1.2.9.b.1 and LCO 3.5.4.a. The proposed changes to Bases 3/4.1.2 
and Bases 3/4.5.4 are associated with these changes. These changes 
are consistent with the revised setpoint analyses.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
invalidate assumptions used in evaluating the radiological 
consequences of an accident, do not alter the source term or 
containment isolation, and do not provide a new radiation release 
path.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not introduce a new or different accident initiator or 
introduce a new or different equipment failure mode or mechanism.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes establish an error analysis that has 
been shown to adequately preserve the margin of safety, and the trip 
setpoint values removed from the TS will be maintained in the DBNPS 
Updated Safety Analysis Report, with proposed changes subject to the 
regulatory requirements of 10 CFR 50.59.

    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.

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

    Date of amendment request: November 8, 1999.
    Description of amendment request: The proposed amendment would 
relocate Technical Specification (TS) 6.5.1, Station Review Board, and 
TS 6.5.2, Company Nuclear Review Board, to Davis-Besse Updated Safety 
Analysis Report Chapter 17.2, Quality Assurance During the Operations 
Phase, also known as the Quality Assurance Program. The proposed 
changes are consistent with the recommendations in NRC Administrative 
Letter 95-06, ``Relocation of Technical Specification Administrative 
Controls Related to Quality Assurance.''
    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 proposed changes would:

    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions or assumptions are affected by the proposed changes to 
Section 6.0, Administrative Controls, of the Technical 
Specifications (TS).
    The proposed changes to relocate the detailed listings of TS 
Section 6.5.1, Station Review Board (SRB), and TS 6.5.2, Company 
Nuclear Review Board (CNRB), to the Davis-Besse Nuclear Power 
Station (DBNPS) Quality Assurance Program in Chapter 17 of the 
Updated Safety Analysis Report are consistent with the NRC's 
guidance in NUREG-1430, ``Standard Technical Specifications--Babcock 
and Wilcox Plants,'' Revision 1 and NRC Administrative Letter 95-06, 
``Relocation of Technical Specification Administrative Controls 
Related to Quality Assurance,'' dated December 12, 1995. These TS 
being relocated will remain subject to the controls of other NRC 
regulations (e.g., 10 CFR 50.54(a)). The proposed changes to the TS 
Index reflect the relocation of TS 6.5.1 and TS 6.5.2. These are 
administrative changes that do not reduce the duties or 
responsibilities of the SRB and CNRB in ensuring the safe operation 
of the DBNPS.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected by the proposed changes. As described 
above, these changes are consistent with the improved ``Standard 
Technical Specifications--Babcock and Wilcox Plants'' (NUREG-1430 
Revision 1) and Administrative Letter 95-06, and are administrative 
changes. The proposed changes do not alter the source term, 
containment isolation, or allowable releases. The proposed changes, 
therefore, will not increase the radiological consequences of a 
previously evaluated accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes, which involve the administrative location for listing SRB 
and CNRB responsibilities. The proposed changes do not alter any 
accident scenarios.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes are administrative and do not reduce or 
adversely affect the capabilities of any plant structures, systems 
or components to perform their nuclear safety functions.

    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

[[Page 70088]]

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.

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

    Date of amendment request: November 1, 1999
    Description of amendment request: The proposed license amendment is 
prescribed by the requested actions of Generic Letter 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The 
proposed amendment will modify the existing Ventilation Filter Testing 
Program contained in Technical Specification 5.5.7.c by replacing the 
reference to ASTM D3803-1986, the standard for charcoal filter testing 
for ESF ventilation systems, with ASTM D3803-1989. The proposed 
amendment will also incorporate the suggested safety factor for 
charcoal filter efficiency regarding methyl iodide penetration.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to reference American Society for Testing 
and Materials (ASTM) D3803-1989, ``Standard Test Method for Nuclear-
Grade Activated Carbon,'' for laboratory testing of Engineered 
Safety Features (ESF) ventilation systems in lieu of ASTM D3803-1986 
is prescribed by the requested actions of Generic Letter (GL) 99-02, 
``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The use 
of ASTM D3803-1989 allows for increased accuracy in monitoring the 
degradation of ESF ventilation system activated carbon (charcoal) 
over time and is a reproducible method for determining the realistic 
capability of charcoal. The 1989 standard is endorsed by the NRC and 
is considered to be more stringent regarding testing criteria than 
the previous referenced standard (1986). GL 99-02 encourages 
addressees, if necessary, to amend their Technical Specifications 
(TS) to reference ASTM D3803-1989 for charcoal filter laboratory 
testing for ESF ventilation systems. In response to the referenced 
GL, the proposed change modifies the existing Perry Nuclear Power 
Plant (PNPP) Ventilation Filter Testing Program (VFTP) contained in 
the PNPP TS to reference ASTM D3803-1989 as the standard for 
charcoal filter laboratory testing for ESF ventilation systems. In 
addition, the proposed change incorporates the safety factor 
suggested within GL 99-02 for charcoal filter efficiency with 
respect to methyl iodide penetration. The proposed change provides 
assurance for compliance with the current licensing basis regarding 
dose limits of General Design Criteria (GDC) 19 of Appendix A to 10 
CFR 50 and 10 CFR 100. The proposed change ensures originally stated 
design criteria are met and therefore does not affect the precursors 
for accidents or transients analyzed in Chapter 15 of the PNPP 
Updated Safety Analysis Report (USAR). With the proposed change, the 
radiological consequences are the same as previously stated in the 
USAR. Therefore, the implementation of the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to reference ASTM D3803-1989 for the 
laboratory testing of charcoal filters of ESF ventilation systems in 
lieu of ASTM D3803-1986 is prescribed by the requested actions of GL 
99-02. ASTM D3803-1989 is endorsed by the NRC and is considered a 
more stringent testing standard than the previous referenced 
standard, ASTM D3803-1986. In addition, the proposed change 
incorporates the safety factor suggested within GL 99-02 for 
charcoal filter efficiency with respect to methyl iodide 
penetration. The proposed change provides assurance for compliance 
with the current licensing basis regarding dose limits of GDC 19 of 
Appendix A to 10 CFR 50 and 10 CFR 100. The proposed change does not 
change the assumptions used in any accident analysis and no new or 
different kind of accident is created. The proposed change ensures 
originally stated design criteria are met and therefore does not 
affect the precursors for accidents or transients analyzed in 
Chapter 15 of the PNPP USAR. Therefore, the implementation of the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is prescribed by the requested actions of GL 
99-02. The use of ASTM D3803-1989 allows for increased accuracy in 
monitoring the degradation of ESF ventilation systems charcoal over 
time and is a very accurate and reproducible method for determining 
the realistic capability of charcoal. ASTM D3803-1989 is considered 
a more stringent testing standard than the previous referenced 
standard, ASTM D3803-1986. Additionally, as specified in GL 99-02, a 
safety factor of 2 has been utilized in the calculation of the 
revised allowable penetration based upon the credited efficiency 
approved by the NRC. The proposed change provides assurance for 
compliance with the current licensing basis regarding dose limits of 
GDC 19 of Appendix A to 10 CFR 50 and 10 CFR 100. Therefore, the 
implementation of the proposed change does not involve a reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: November 1, 1999
    Description of amendment request: Technical Specification 
Surveillance Requirement (SR) 3.6.1.7.4 requires that each containment 
spray nozzle be verified unobstructed on a 10-year frequency. The 
proposed amendment would revise the frequency for SR 3.6.1.7.4 from 
once every 10 years to only those conditions when maintenance is 
performed which could result in nozzle blockage.
    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 would not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change revises the surveillance frequency from 
every 10 years to following maintenance that could result in nozzle 
blockage. Analyzed events are initiated by the failure of plant 
structures, systems or components. The containment spray system is 
not considered as an initiator of any analyzed event. The proposed 
change does not have a detrimental impact on the integrity of any 
plant structure, system or component that initiates an analyzed 
event. The proposed change will not alter the operation of, or 
otherwise increase the failure probability of any plant equipment 
that initiates an analyzed accident. As a result, the probability of 
any accident previously evaluated, is not significantly increased.
    The proposed change revises the Surveillance Frequency. Reduced 
testing is acceptable where operating experience has shown that 
these components usually pass the Surveillance when performed at the 
specified interval, thus the frequency is acceptable from a 
reliability standpoint. The proposed containment spray nozzle 
Surveillance Frequency has been established based on achieving 
acceptable levels of equipment

[[Page 70089]]

reliability. This change does not affect the plant design. Due to 
the plant design, the spray header is maintained dry and alarmed on 
water intrusion. Formation of significant corrosion products is 
unlikely. Due to its location at the top of the containment, 
introduction of foreign material from exterior to the header is 
unlikely. Since maintenance that could introduce foreign material is 
the most likely cause for obstruction, testing or inspection 
following such maintenance would verify the nozzle(s) being 
unobstructed, and the system would be capable of performing its 
safety function. As a result, the consequences of any accident 
previously evaluated are not significantly affected.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed change would not create the possibility of a new 
of 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.
    3. The proposed change would not involve a significant reduction 
in a margin of safety.
    The margin of safety for this system is based on the capacity of 
the spray headers. Since the system is not susceptible to corrosion 
induced obstruction or obstruction from external to the system, and 
performance of maintenance on the system would require evaluation of 
the potential for nozzle blockage and the need for a test or 
inspection, the spray header nozzles will not become blocked in the 
event that the safety function is required. Therefore, the capacity 
of the system would remain unaffected. Hence, this change does not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: November 17, 1999.
    Description of amendment request: The proposed amendments for St. 
Lucie, Units 1 and 2, will revise the current 72-hour action completion 
allowed outage time (AOT) specified in Technical Specification (TS) 
3.8.1.1, Action ``b,'' to allow 14 days to restore an inoperable 
emergency diesel generator set to operable status. The proposed AOT is 
based on an integrated review and assessment of plant operations, 
deterministic design basis factors, and an evaluation of overall plant 
risk using probabilistic safety assessment techniques.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments for St. Lucie Unit 1 and Unit 2 will 
extend the action completion/allowed outage time (AOT) for a single 
inoperable Emergency Diesel Generator (EDG) from 72 hours to 14 
days. The EDGs are designed as backup AC power sources for essential 
safety systems in the event of a loss of offsite power. As such, the 
EDGs are not accident initiators, and an extended AOT to restore 
operability of an inoperable diesel generator would not 
significantly increase the probability of occurrence of accidents 
previously analyzed.
    The proposed technical specification revisions involve the AOT 
for a single inoperable EDG, and do not change the conditions, 
operating configuration, or minimum amount of operating equipment 
assumed in the plant safety analyses for accident mitigation. Plant 
defense-in-depth capabilities will be maintained with the proposed 
AOT, and the design basis for electric power systems will continue 
to conform with 10 CFR 50, Appendix A, General Design Criterion 17. 
In addition, a Probability Safety Assessment (PSA) was performed to 
quantitatively assess the risk-impact of the proposed amendment for 
each unit. The impact on the early radiological release probability 
for design basis events was also evaluated and it is concluded that 
the risk contribution from this proposed AOT is small and consistent 
with regulatory risk-assessment acceptance guidelines. Therefore, 
operation of either facility in accordance with its proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendments will not change the physical plant or 
the modes of operation defined in either facility license. The 
changes do not involve the addition of new equipment or the 
modification of existing equipment, nor do they alter the design of 
St. Lucie plant systems. Therefore, operation of either facility in 
accordance with its proposed amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendments are designed to improve EDG reliability 
by providing flexibility in the scheduling and performance of 
preventive and corrective maintenance activities. The surveillance 
intervals or the operability requirements are not changed by the 
proposal; only the AOT for a single inoperable EDG will be extended. 
The proposed changes do not alter the basis for any technical 
specification that is related to the establishment of, or the 
maintenance of, a nuclear safety margin, and design defense-in-depth 
capabilities are maintained. An integrated assessment of the risk 
impact of extending the AOT for a single inoperable EDG has 
determined that the risk contribution is small and is within 
regulatory guidelines for an acceptable TS change. Therefore, 
operation of either facility in accordance with its proposed 
amendment would not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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.

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: November 23, 1999.
    Description of amendment request: The proposed license amendments 
are submitted in response to Generic Letter (GL) 99-02, Laboratory 
Testing of Nuclear-Grade Activated Charcoal, which requires that 
American Society for Testing and Materials (ASTM) D3803-1989 be used 
for testing both new and used charcoal in engineered safety feature 
applications. The proposed amendments would modify Technical 
Specification (TS) 3/4.6.3, EMERGENCY CONTAINMENT FILTERING SYSTEM, TS 
3/4.6.6, POST ACCIDENT CONTAINMENT VENT SYSTEM, and TS 3/4.7.5, CONTROL 
ROOM EMERGENCY VENTILATION SYSTEM.

[[Page 70090]]

    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.
    The probability of occurrence of an accident previously 
evaluated for Turkey Point is not altered by the proposed TS changes 
because no physical modifications are being made to the plant.
    The proposed change requires that new and used charcoal in the 
plant engineered safety feature (ESF) ventilation systems be tested 
in accordance with ASTM D3803-1989, at a temperature of 30  deg.C 
and a relative humidity of 95%. The use of a new or different test 
standard to satisfy the charcoal surveillance test requirement does 
not change the radiological consequences of any previously evaluated 
accident. The adoption of the ASTM standard will, however, require 
that future charcoal samples from the emergency containment filters 
be tested for methyl iodide removal rather than elemental iodine 
removal as permitted by previous test protocols. The revised test 
method will provide a more uniform test program for the ESF filters, 
and will not adversely affect the filters affinity for elemental 
iodine removal. The adoption of the ASTM standard for laboratory 
analysis of the ESF charcoal does not impact the design bases of the 
ESF systems, alter post-accident source terms, or modify the removal 
efficiencies credited in the facility dose calculations.
    The ASTM standard is very stringent and has been shown to 
provide a more reliable measure of the ability of charcoal to 
fulfill its intended design function, i.e., to remove radioiodine in 
any chemical form from the attendant plant gas stream, than previous 
test protocols. Consequently, the adoption of the ASTM standard for 
laboratory analysis of the ESF charcoal will ensure that Turkey 
Point is operated in a manner consistent with the licensing basis of 
the facility as it relates to the protection of the public and the 
control room operators during radiological accidents.
    Based on the above, it is concluded that the proposed amendment 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    (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 previously evaluated.
    The proposed change does not create a new or different type of 
accident for Turkey Point because no physical plant changes are 
being made, and no compensatory measures are imposed that would 
create a new failure scenario. The proposed change only imposes a 
more stringent surveillance requirement for both new and used 
charcoal in the plant ESF ventilation systems. Since no new failure 
modes are associated with the proposed changes, the activity does 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed license amendment adopts a more stringent standard 
for performing laboratory surveillance tests on both new and used 
charcoal in the ESF ventilation systems. Given the increased 
accuracy of the proposed test standard, the amendment also supports 
the adoption of revised acceptance criteria having a lower safety 
factor to the plant safety analysis limits. The composite change 
does not impact the design bases of the ESF systems, alter post-
accident source terms, or modify the removal efficiencies credited 
in the facility dose calculations
    The margin of safety associated with operation of the ESF 
ventilation systems is established by the facility dose calculations 
and the acceptance criteria for system performance defined in 10 CFR 
100 and Criterion 19 of Appendix A to 10 CFR 50. The proposed 
amendments will not change this acceptance criteria nor the 
calculated dose limits used to establish the current plant-licensing 
basis.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Richard P. Corriea.

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

    Date of application for amendment: October 12, 1999.
    Brief description of amendment: The proposed amendment would revise 
the Appendix B Environmental Protection Plan of the Crystal River Unit 
3 (CR-3) Operating License. The changes incorporate requirements from a 
biological opinion (BO) issued by the National Marine Fisheries Service 
(NMFS). The BO reviews the effects of the cooling water intake system 
on species of sea turtles protected by the Endangered Species Act 
(ESA). Additionally, other administrative changes are proposed to 
Appendix B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed changes to the CR-3 EPP are administrative in 
nature and reflect the information provided in the NMFS BO. These 
changes do not affect the initial conditions, assumptions, or 
conclusions of the CR-3 accident analyses. In addition, the proposed 
changes do not affect the operation or performance of any equipment 
assumed in the accident analyses. Therefore, the proposed changes 
would not significantly increase the probability or consequences of 
an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from previously evaluated accidents?
    The proposed changes are administrative in nature and reflect 
information provided by the NMFS BO regarding the incidental taking 
of species of sea turtles protected by the ESA. These changes do not 
impact or alter the configuration or operation of the facilities and 
do not create any new modes of operation. Therefore, the proposed 
changes would not create the possibility of a new or different kind 
of accident.
    3. Involve a significant reduction in a margin of safety?
    As indicated above, the proposed changes do not change the 
configuration or operation of the plant and do not affect the CR-3 
accident analyses. The proposed changes are administrative in nature 
and do not affect any margin of safety for CR-3. Therefore, the 
proposed changes would 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, General Counsel (MAC-
BT15A), Florida Power Corporation, P. O. Box 14042, St. Petersburg, 
Florida 33733-4042.
    NRC Section Chief: Richard Correia.

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

    Date of amendment request: October 29, 1999.
    Description of amendment request: The proposed license amendment 
would modify the Technical Specifications (TSs) to: (1) Add operating 
limits for make-up tank (MUT) level and pressure in a new figure 3.3.1; 
(2) add surveillance requirements for the MUT pressure instrument 
channel; (3) change the frequency of calibration for the MUT level 
instrument from F (every 24 months) to R (refueling interval); (4) 
change the frequency of calibration for

[[Page 70091]]

the high pressure injection (HPI) and low pressure injection (LPI) flow 
instruments; and (5) make minor editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not represent a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The changes included in this LCA [License Change Application] 
impose new requirements for MU/HPI system operation and testing and 
extension of calibration frequencies for the MUT level, HPI flow and 
LPI flow instruments. These changes could not result in initiation 
of any accident previously evaluated. Therefore, the probability of 
an accident could not be affected by changes to the MU/HPI system.
    As described in the list of benefits for operation with the MU/
HPI cross-connect valves open, listed in Section III.B above 
[Section III.B of the October 29, 1999 application], the purpose of 
changing the operation of the MU/HPI system was to preclude the 
possibility of HPI pump damage. The addition of surveillance 
requirements for the MUT pressure instrument and the addition of LCO 
[limiting conditions for operation] limits on MUT level and pressure 
along with an appropriate action statement and AOT [allowed outage 
time] will ensure that gas entrainment of the MUT does not occur. 
The proposed change in instrument calibration frequencies will 
continue to maintain the required accuracy of the MUT level, HPI 
flow, and LPI flow instruments.
    Minor editorial changes are included in this request to improve 
clarity and readability of the T.S. and could not adversely affect 
plant operation.
    Therefore, the proposed changes will not adversely impact the 
reliability of the MU/HPI system and could not represent a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This LCA does not involve the addition of any new hardware. 
Along with minor editorial changes, the requested changes involve 
MU/HPI system operation and testing, which could only affect RCS 
[reactor coolant system] coolant inventory changes during operation 
and the ability to provide protection in the event of a Loss of 
Coolant Accident (LOCA). The full spectrum of LOCAs has been 
evaluated in the FSAR [Final Safety Analysis Report]. Therefore, no 
new accident scenarios have been created.
    The additional controls on MUT level and pressure provided by 
this LCA will ensure that a malfunction of a different type, gas 
entrainment of the MU/HPI pumps, will not occur. These limits on MUT 
level and pressure ensure that the initial conditions assumed for 
ECCS [emergency core cooling system] operation are maintained. The 
T.S. limits maintain the accident analysis initial conditions such 
that no operator action is required to meet NPSH [net positive 
suction head] or to avoid gas entrainment during ECCS operation with 
the postulated single failure as required by the TMI-1 licensing 
basis (Reference 14) [of the October 29, 1999, application].
    Extension of the calibration frequencies for the HPI level, HPI 
flow, and LPI flow will continue to maintain the accuracy of these 
instruments and could not create the potential for any new accident 
that has not been evaluated.
    Minor editorial changes are included in this request to improve 
the clarity and readability of the T.S. and could not adversely 
affect plant operation.
    Therefore, these changes do not create the potential for any 
accident different from those that have been evaluated.
    3. These proposed changes do not involve a significant reduction 
in a margin of safety.
    This LCA includes changes to the MU/HPI system operation and 
testing and an extension of the calibration frequency for certain 
instrument[s]. The requested changes will serve to maintain the 
proper system initial conditions to ensure the ability of the MU/HPI 
system to provide protection in the event of a Loss of Coolant 
Accident (LOCA) and maintain the required instrument accuracy for 
the instruments where changes to a refueling interval frequency are 
being requested. NRC guidance for addressing the effect on increased 
surveillance intervals on instrument drift and safety analysis 
assumptions presented in GL [generic letter] 91-04 has been 
addressed in enclosure 1A above [of the October 29, 1999, 
application].
    Minor editorial changes are included in this request to improve 
the clarity and readability of the T.S. and could not adversely 
affect plant operation.
    These changes, which are consistent with the TMI-1 licensing and 
design basis requirements, do not result in a degradation of safety 
related equipment, and therefore, 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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Sheri R. Peterson.

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 17, 1999.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) values for methyl iodide 
penetration for the main control room environmental control system and 
the standby gas treatment system. Also, editorial revisions are being 
made to portions of TS Section 5.0 to reference the correct sections of 
Regulatory Guide 1.52.
    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 the consequences of a previously evaluated 
accident.
    This proposed revision makes changes to Technical Specification 
(TS) Section 5.5.7, ``Ventilation Filter Testing Program'' (VFTP). 
The references to sections in the Regulatory Guide 1.52, Revision 2 
for VFTP are being corrected. Additionally, the proposed revision 
also changes the allowable methyl iodide penetration percent for the 
carbon in the Standby Gas Treatment (SGT) and the Main Control Room 
Environmental Control (MCREC) systems when tested in accordance with 
ASTM DS3803-1989. This is based on the values that would be derived 
using a factor of safety of 2 between the credited and tested carbon 
efficiencies. This safety factor is contained in the Generic Letter 
99-02. The Generic Letter allows the reduction of the factor of 
safety between the credited and tested carbon efficiencies from 5 
(for systems with heaters) and 7 (for systems without heaters) to 2 
(for systems with or without heaters) when tested per ASTM D-3803-
1989. Since the factor of safety of 2 is maintained, this change 
does not involve a significant increase in the probability or the 
consequences of a previously evaluated event. The changes in the 
section references to Regulatory Guide 1.52 Revision 2 for the 
Ventilation Filter Testing Program (VFTP) are considered to be 
editorial corrections.
    2. The change does not involve a significant increase in the 
probability of or the consequences of an event not previously 
analyzed.
    This proposed revision makes changes to TS Section 5.5.7, 
``Ventilation Filter Testing Program'' (VFTP). The section 
references to Regulatory Guide 1.52 Revision 2 for the Ventilation 
Filter Testing Program (VFTP) are being corrected. The change in the 
allowable methyl iodide penetration percent is based

[[Page 70092]]

on the values that would be derived using the safety factor of 2 
contained in Generic Letter 99-02. The Generic Letter will reduce 
the factor of safety between the credited and tested carbon 
efficiencies from 5 (for systems with heaters) and 7 (for systems 
without heaters) to 2 if tested per ASTM D-3803-1989. Since the 
credited carbon efficiencies in the dose calculations are not being 
compromised, this change will not involve a significant increase in 
the probability of, or the consequences of an event not previously 
analyzed.
    The changes in the section references to Reg. Guide 1.52 are 
editorial and thus do not significantly increase the probability of, 
or the consequences of a previously unanalyzed event.
    3. The change does not significantly reduce the margin of 
safety.
    The change in the allowable methyl iodide penetration percent 
implements the Generic Letter's carbon efficiency safety factor of 2 
between the credited and the tested carbon efficiencies. Per the 
generic letter, it is acceptable to use this new safety factor since 
the new standard is more accurate and demanding than previous ones. 
Therefore, the proposed revision will not significantly reduce the 
margin of safety. The changes in the section references for 
Regulatory Guide 1.52 Revision 2 are considered to be editorial 
corrections.

    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.

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

    Date of amendment request: November 15, 1999 (TS 99-016).
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) for Watts Bar Unit 1 to: (1) 
revise the Watts Bar TS and associated TS Bases for TS 3.6.11.5 to 
change the methodology and frequency for sampling the ice condenser ice 
bed (stored ice) and (2) add a new TS 3.6.11.7 and associated TS Bases 
to address sampling requirements for all ice additions to the ice bed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The Proposed Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated.
    The only analyzed accidents of possible consideration in regards 
to changes potentially affecting the ice condenser are a loss of 
coolant accident (LOCA) and a main steam line break (MSLB) inside 
containment. However, the ice condenser is not postulated as being 
the initiator of any LOCA or MSLB. This is because it is designed to 
remain functional following a design basis earthquake, and the ice 
condenser does not interconnect or interact with any systems that 
interconnect or interact with the reactor coolant or main steam 
systems. Since the proposed changes to the TS and TS Bases are 
solely to revise and provide clarification of the ice sampling and 
chemical analysis requirements, and are not the result of or require 
any physical change to the ice condenser, then there can be no 
change in the probability of an accident previously evaluated in the 
Safety Analysis Report (SAR).
    In order for the consequences of any previously evaluated event 
to be changed, there would have to be a change in the ice 
condenser's physical operation during a LOCA or MSLB, or in the 
chemical composition of the stored ice. The proposed changes do not 
alter either from existing requirements, except to add an upper 
limit on boron concentration, which is the bounding value for the 
Hot Leg Switchover timing calculation. Though the frequency of the 
existing surveillance requirement for sampling the stored ice is 
changed from once every 18 months to once every 54 months, the 
sampling requirements are strengthened overall with (1) the 
requirement to obtain one randomly selected sample from each ice 
condenser bay (24 total samples) rather than nine ``representative'' 
samples, and (2) the addition of a new surveillance requirement to 
verify each addition of ice meets the existing requirements for 
boron concentration and pH value. The only other change is to 
clarify that each sample of stored ice is individually analyzed for 
boron concentration and pH, but that the acceptance criteria for 
each parameter is based on the average values obtained for the 24 
samples. This is consistent with the bases for the boron 
concentration of the ice, which is to ensure the accident analysis 
assumptions for containment sump pH and boron concentration are not 
altered following complete melting of the ice condenser. 
Historically, chemical analysis of the stored ice has had a very 
limited number of instances where an individual sample did not meet 
the boron or pH requirements, with all subsequent evaluations 
(follow up sampling) showing the ice condenser as a whole was well 
within these requirements. Requiring chemical analysis of each 
sample is provided to preclude the practice of melting all samples 
together before performing the analysis, and to ensure the licensee 
is alerted to any localized anomalies for investigation and 
resolution without the burden of entering a 24 hour ACTION 
Condition, provided the averaged results are acceptable. Thus, based 
on the above, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    B. The Proposed Change Does Not Create The Possibility Of A New 
Or Different Kind Of Accident From Any Accident Previously 
Evaluated.
    Because the TS and TS Bases changes do not involve any physical 
changes to the ice condenser, any physical or chemical changes to 
the ice contained therein, or make any changes in the operational or 
maintenance aspects of the ice condenser as required by the Tech 
Specs, there can be no new accidents created from those already 
identified and evaluated.
    C. The Proposed Change Does Not Involve A Significant Reduction 
In A Margin Of Safety.
    The ice condenser Technical Specifications ensure that during a 
LOCA or SLB the ice condenser will initially pass sufficient air and 
steam mass to preclude over pressurizing lower containment, that it 
will absorb sufficient heat energy initially and over a prescribed 
time period to assist in precluding containment vessel failure, and 
that it will not alter the bulk containment sump pH and boron 
concentration assumed in the accident analysis. Since the proposed 
changes do not physically alter the ice condenser, but rather only 
serve to strengthen and clarify ice sampling and analysis 
requirements, the only area of potential concern is the effect these 
changes could have on bulk containment sump pH and boron 
concentration following ice melt. However, this is not affected 
because there is no change in the existing requirements for pH and 
boron concentration, except to add an upper limit on boron 
concentration. This upper limit is the bounding value for the Hot 
Leg Switchover timing calculation. Averaging the pH and boron values 
obtained from analysis of the individual samples taken is not a new 
practice, just one that was not consistently used by all ice 
condenser plants. Using the averaged values provides an equivalent 
bulk value for the ice condenser, which is consistent with the 
accident analysis for the bulk pH and boron concentration of the 
containment sump following ice melt. Changing the performance 
frequency for sampling the stored ice does not reduce any margin of 
safety because (1) the newly proposed surveillance (SR 3.6.15.7) 
ensures ice additions meet the existing boron concentration and pH 
requirements, (2) there are no normal operating mechanisms, 
including sublimation, that reduce the ice condenser bulk pH and 
boron concentration, and (3) the number of required samples has been 
increased from nine to 24 (one randomly selected ice basket per 
bay), which is approximately the same number of samples that would 
have been taken in the same time period under the existing 
requirements. Thus, it can be concluded that the proposed TS and TS 
Bases changes do not involve a significant reduction in the margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 70093]]

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 Correia.

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

    Date of amendment request: November 20, 1998 and July 19, 1999 
(TS99-014).
    Description of amendment request: The proposed amendment would 
revise the Watts Bar Nuclear plant Unit 1 Technical Specifications (TS) 
and associated TS Bases to alter the acceptance criteria in 
Surveillance Requirement (SR) 3.6.11.4 and to revise the Bases for TS 
3.6.12. The changes would replace the current visual inspection 
requirement that uses a 0.38 inch ice/frost buildup criterion with a 
visual surveillance program that provides an increased confidence level 
that flow blockage in ice condenser baskets does not exceed the 15 
percent assumed in the accident analyses. The proposed amendment dated 
July 19, 1999 is considered to supercede and replace entirely a 
proposed amendment dated November 20, 1998 on this same subject.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Neither the TS amendment nor the TS Bases changes can increase 
the probability of occurrence of any analyzed accident because they 
are not the result or cause of any physical modification to ice 
condenser structures, and for the current design of the ice 
condenser, there is no correlation between any credible failure of 
it and the initiation of any previously analyzed event.
    Regarding the consequences of analyzed accidents, the ice 
condenser is an engineered safety feature designed, in part, to 
limit the containment subcompartment and steel containment vessel 
pressures immediately following the initiation of a LOCA [loss-of-
coolant accident] or HELB [high energy line break]. Conservative 
subcompartment pressure analysis shows this criteria will be met if 
the reduction in the flow area per bay provided for ice condenser 
air/steam flow channels is less than or equal to 15 percent, or if 
the total flow area blocked within each lumped analysis section is 
less than or equal to the 15 percent assumed in the safety analysis. 
The present 0.38 inch frost/ice buildup surveillance criteria only 
addresses the acceptability of any given flow channel, and has no 
direct correlation between flow channels exceeding this criteria and 
percent of total flow channel blockage. In fact, it was never the 
intent of the current SR to make such a correlation. If problems 
were encountered in meeting the 0.38 inch criteria, it was expected 
that additional inspection and analysis, such as provided in the 
proposed amendment, would be performed to make such a determination.
    Verifying an ice bed is left with less than or equal to 15 
percent flow channel blockage at the conclusion of a refueling 
outage assures the ice bed will remain in an acceptable condition 
for the duration of the operating cycle. During the operating cycle, 
a certain amount of ice sublimates and reforms as frost on the 
colder surfaces in the Ice Condenser. However, frost does not 
degrade flow channel area. The surveillance will effectively 
demonstrate operability for an allowed 18 month surveillance period. 
Therefore, limiting ice bed flow channel blockage to less than or 
equal to 15 percent ensures operation is consistent with the 
assumptions of the design basis accident (DBA) analyses. Thus, the 
proposed amendment for flow blockage determination provides the 
necessary assurance that flow channel requirements are met without 
additional evaluations, and thus will not increase the consequences 
of a LOCA or HELB.
    In regard to the TS 3.6.12 Bases change, clarifying that 
Condition B does not apply when personnel are standing on or opening 
doors for a short duration to perform surveillances or minor 
maintenance activities, such as ice removal, does not increase 
analyzed accident consequences. These are not new or additional 
actions to those performed previously, the probability of an 
accident versus the time to perform these actions is small, the 
number of personnel involved is small, and their duration is 
generally much less than the four hour frequency of Required Action 
B.1 (monitor maximum ice condenser temperature). Therefore, these 
activities do not adversely affect ice bed sublimation, melting, or 
ice condenser flow paths. However, if during these activities any 
door is determined to be restrained, not fully closed from a 
previous activity, or otherwise not operable, then separate entry 
into Condition B is required for each door so identified.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    For such a possibility to exist, there would have to be either a 
physical change to the ice condenser, or some change in how it is 
operated or physically maintained. None of the above is true for the 
proposed TS amendment and TS Bases change.
    There is no change to the existing design requirements or 
inputs/results of any accident analysis calculations.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    Design Basis Accident analyses have shown that with 85 percent 
of the total flow area available (uniformly distributed), the ice 
condenser will perform its intended function. Thus, the safety limit 
for ice condenser operability is a maximum 15 percent blockage of 
flow channels. SR 3.6.11.4 currently uses a specific value of 0.38 
inch buildup to determine if unacceptable frost/ice blockage exists 
in the ice condenser. However, this specific value does not have a 
direct correlation to the safety limit for blockage of ice condenser 
flow area. The proposed TS amendment requires more extensive visual 
inspection (33 percent of the flow area/bay) than is currently 
described (2 flow channels/bay) in the TS Bases for SR 3.6.11.4, 
thus providing greater reliability and a direct relationship to the 
analytical safety limits. Changing the TS to implement a 
surveillance program that is more reliable and uses acceptance 
criteria of less than or equal to 15 percent flow blockage, as 
allowed by the TMD [transient mass distribution] analysis, will not 
reduce the margin of safety of any TS.
    Additionally, verifying an ice bed is left with less than or 
equal to 15 percent flow channel blockage at the conclusion of a 
refueling outage assures the ice bed will remain in an acceptable 
condition for the duration of the operating cycle. During the 
operating cycle, a certain amount of ice sublimates and reforms as 
frost on the colder surfaces in the Ice Condenser. However, frost 
has been determined to not degrade flow channel flow area. Thus, 
design limits for the continued safe function of containment 
subcompartment walls and the steel containment vessel are not 
exceeded due to this change.
    The change made to TS 3.6.12 Bases does not affect the margin of 
safety as defined in any TS as it does not involve design 
specifications or acceptance criteria. This change only adds a 
clarifying note that entry into Condition B is not required solely 
because of actions (standing on and opening intermediate/upper deck 
doors) necessary for the performance of required ice condenser 
surveillances, maintenance, or routine activities. This does not 
preclude entry into Condition B during performance of these 
activities should an intermediate deck door or upper deck door 
otherwise be determined inoperable.

    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: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Section Chief: Richard P. Correia.

[[Page 70094]]

Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing

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

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: November 8, 1999.
    Description of amendment request: The amendment changed action 
statements, definitions, and footnotes pertaining to the Technical 
Specifications for primary containment leakage and primary containment 
purge system to allow an alternative approach to the existing 
requirement.
    Date of publication of individual notice in Federal Register: 
November 16, 1999 (64 FR 62228).
    Expiration date of individual notice: December 16, 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 
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).

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

    Date of application for amendment: September 7, 1999, as 
supplemented on October 1, 1999.
    Brief description of amendment: This amendment reassigns the 
responsibilities of the Site Manager, who works for the Westinghouse 
Electric Company (a contractor to CBS), to the TR-2 Decommissioning 
Project Director, who works for CBS.
    Date of issuance: November 23, 1999.
    Effective Date: November 23, 1999.
    Amendment No: 10.
    Facility License No. TR-2: This amendment changes the 
decommissioning plan.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56529).
    The Commission has issued a Safety Evaluation for this amendment 
dated November 23, 1999.
    No significant hazards consideration comments received: No.

Commonwealth Edison Company, Docket No. 50-254, Quad Cities Nuclear 
Power Station, Unit 1, Rock Island County, Illinois

    Date of application for amendment: March 30, 1999.
    Brief description of amendment: The amendment revises the Technical 
Specifications by changing Surveillance Requirement 4.6.E.2 to allow a 
one-time extension of the 18-month requirement to pressure set test or 
replace one half of the Main Steam Safety Valves to an interval of 24 
months.
    Date of issuance: November 30, 1999.
    Effective date: Immediately, to be implemented within 60 days.
    Amendment No.: 191.
    Facility Operating License No. DPR-29: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 5, 1999 (64 FR 
24194).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 30, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 6, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) to expand the allowable values for 
Interlocks P-6 (Intermediate Range Neutron Flux) and P-10 (Power Range 
Neutron Flux) in TS 3.3.1, Table 3.3.1-1, Function 16, Reactor Trip 
System Interlocks, as recommended by Westinghouse.
    Date of issuance: November 30, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-189; Unit 2-170.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27319).
    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.

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

    Date of application for amendment: May 5, 1999.
    Brief description of amendment: This amendment conforms the license 
to reflect the transfer of Operating License NPF-58 for the Perry 
Nuclear Power Plant, Unit 1, to the extent held by Duquesne Light 
Company, to the Cleveland Electric Illuminating Company as previously 
approved by an Order dated September 30, 1999.
    Date of issuance: December 3, 1999.
    Effective date: December 3, 1999.

[[Page 70095]]

    Amendment No.: 108.
    Facility Operating License No. NPF-58: This amendment revised the 
operating license.
    Date of initial notice in Federal Register: June 14, 1999 (64 FR 
31879).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 30, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 27, 1999, as supplemented 
October 4, 1999.
    Brief description of amendments: Revises the Technical 
Specifications (TS) to extend the allowed outage time, on a one-time 
basis, for an inoperable emergency diesel generator from 72 hours to 7 
days, to replace the Unit 3 diesel engine radiators prior to April 
2000. The revision applies to Turkey Point Unit 3 only, however, Unit 4 
is included administratively because the TS are combined for both 
Units.
    Date of issuance: November 19, 1999.
    Effective date: As of date of issuance, to be implemented prior to 
April 2000.
    Amendment Nos.: 202 and 196.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the TS.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46441). The supplemental letter of October 4, 1999, provided 
clarification information that did not change the original no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 19, 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 8, 1998.
    Brief description of amendments: The proposed amendments would 
change the Technical Specifications for both units to place tighter 
restrictions on the allowed outage time for the refueling water storage 
tank water level instrumentation.
    Date of issuance: November 30, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 232 and 215.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1999 (64 FR 
47532). 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.

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 10, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.4.7 so that the surveillance requirement does 
not need to be performed when the reactor is defueled with no forced 
circulation. The revision to TS 3/4.4.7 also includes changes to Tables 
3.4-1 and 4.4-3. TS Table 4.4-3 is revised to change the reactor 
coolant system (RCS) chemistry sampling frequency from three times per 
7 days with a maximum interval of 72 hours to a frequency of at least 
once per 72 hours. An editorial change to Unit 1 Tables 3.4-1 and 4.4-3 
relocates the asterisk for the footnote to a position adjacent to the 
parameter ``dissolved oxygen,'' from its current position next to the 
allowable chemistry limit in Table 3.4-1 and the analysis frequency in 
Table 4.4-3. An editorial change also corrects the footnote for Table 
3.4-1 for Unit 1 and Unit 2 by making the word ``limit'' plural, as it 
applies to both the steady-state and transient limits. Surveillance 
Requirement 4.11.2.2 is revised to delete the phrase ``by analysis of 
the Reactor Coolant System noble gases.''
    Date of issuance: November 19, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 231 and 214.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54376).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 19, 1999.
    No significant hazards consideration comments received: No.

PECO Energy Company, Docket No. 50-352, Limerick Generating Station, 
Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: June 7, 1999.
    Brief description of amendment: The amendment revised the technical 
specifications (TSs) to reflect the permanent deactivation in the 
closed position of the ``wet'' instrument reference leg isolation valve 
HV-61-102. Specifically, TS Table 3.6.3.1, ``Primary Containment 
Isolation Valve,'' and its associated notations were revised to reflect 
this current plant configuration.
    Date of issuance: November 18, 1999.
    Effective date: As of its date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 138.
    Facility Operating License No. NPF-39. This amendment revised the 
TSs.
    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54380).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 18, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 15, 1999, as 
supplemented January 18 and October 22, 1999.
    Brief description of amendment: The amendment provides a revision 
to the Technical Specifications for the FitzPatrick Nuclear Power Plant 
by modifying the description of what constitutes an acceptable Local 
Power Range Monitor calibration.
    Date of issuance: November 22, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 257.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1999 (64 FR 
11965).
    The January 18, 1999, and October 22, 1999, letters 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 November 22, 1999.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 22, 1999.

[[Page 70096]]

    Brief description of amendment: This amendment changes the 
Technical Specifications by extending the pressure-temperature (P-T) 
limit curves to 24 effective full-power years (EFPY) and 32 EFPY. The 
current P-T limit curves are valid through 16 EFPY.
    Date of issuance: November 29, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 258.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43775).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 29, 1999.
    No significant hazards consideration comments received: No.

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

    Date of amendments request: June 30 1997, as supplemented by 
letters of February 22, March 19, June 30, and October 4, 1999.
    Brief Description of amendments: The amendments change the 
Technical Specifications (TS) to clarify surveillance requirements for 
the control room emergency filtration system, penetration room 
filtration system, and related storage pool ventilation system. The 
changes also revised the required number of radiation monitoring 
instrumentation channels, and deleted the containment purge exhaust 
filter TS.
    Date of issuance: November 23, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 145 and 136.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1999 (64 
FR 47870).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 23, 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: September 21, 1999.
    Brief description of amendment: The amendment increases the 
required volume of stored fuel in the diesel fuel oil storage tank as a 
result of a conservative recalculation of diesel generator fuel 
consumption.
    Date of Issuance: November 22, 1999.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 180.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56537). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated November 22, 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: September 21, 1999, as supplemented by 
letter dated November 5, 1999.
    Brief description of amendment: The amendment extended the 
effective full implementation date by six months, from December 31, 
1999, to June 30, 2000, for Amendment No. 120 issued March 22, 1999, 
that approved a modification to increase the storage capacity of spent 
fuel assemblies at the site. The extension is due to delays fabricating 
and installing the new fuel storage racks.
    Date of issuance: November 30, 1999.
    Effective date: November 30, 1999, to be implemented by June 30, 
2000.
    Amendment No.: 129.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 20, 1999 (64 FR 
56538). The supplemental letter of November 5, 1999, 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 
published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 30, 1999.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an

[[Page 70097]]

opportunity for public comment. If comments have been requested, it is 
so stated. In either event, the State has been consulted by telephone 
whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and 
electronically from the ADAMS Public Library component on the NRC Web 
site, http://www.nrc.gov (the Electronic Reading Room).
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By January 14, 1999, 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW, Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).

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

    Date of application of amendments: November 17, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications to modify the definition

[[Page 70098]]

of steam generator repair limit for axial tube imperfections detected 
between the primary side surface of the tube sheet clad and the end of 
the tube.
    Date of Issuance: December 3, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-308; Unit 2-308; Unit 3-308.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes. The NRC published a public notice of the proposed 
amendments, issued a proposed finding of no significant hazards 
consideration and requested that any comments on the proposed no 
significant hazards consideration be provided to the staff by the close 
of business on December 2, 1999. The notice was published in the 
``Greenville News,'' Greenville, SC; and the ``Anderson Independent-
Mail,'' Anderson, SC, on November 24, 1999. No comments have been 
received.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, consultation with the State of South Carolina, 
and final no significant hazards consideration determination are 
contained in a Safety Evaluation dated December 3, 1999.
    Attorney for licensee: Richard W. Blackburn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington DC 20005.
    NRC Section Chief: Richard L. Emch, Jr.

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 application for amendments: November 10, 1999 (PCN-510).
    Brief description of amendments: The amendments modify the 
Technical Specification Limiting Condition for Operation 3.4.9.b to 
delete the phrase stating that two groups of pressurizer heaters be 
``capable of being powered from an emergency power supply.
    Date of issuance: November 22, 1999.
    Effective date: November 22, 1999.
    Amendment Nos.: Unit 2-161; Unit 3-152.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes. The NRC published a public notice of the proposed 
amendments, issued a proposed finding of no significant hazards 
consideration, and requested that any comments on the proposed no 
significant hazards consideration be provided to the staff by close of 
business November 19 , 1999. The notice was published in the ORANGE 
COUNTY REGISTER on November 15-16, 1999. No public comments were 
received.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
November 22, 1999.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-32311 Filed 12-14-99; 8:45 am]
BILLING CODE 7590-01-P