[Federal Register Volume 65, Number 56 (Wednesday, March 22, 2000)]
[Notices]
[Pages 15375-15397]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-6913]


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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 February 26, 2000, through March 10, 2000. 
The last biweekly notice was published on March 8, 2000 (65 FR 12286).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By April 21, 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

[[Page 15376]]

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. 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).

Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power 
Station, Unit 2, Grundy County, Illinois

    Date of amendment request: April 30, 1999.
    Description of amendment request: The proposed amendment would 
revise the expiration date of the operating license to allow 40 years 
of operation from the original date of issuance.
    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?
    The programs to detect incipient failures or degraded 
performance such as Inservice Inspection, Inservice Testing, and 
Environmental Qualification programs, for example, remain in place 
and unchanged. The thermal cycles and reactor vessel toughness are 
within the 40-year design margin and will remain within those 
margins for the total operating period proposed by the amendment. No 
equipment is added, modified, or removed as a result of this 
amendment. Therefore there is no increase in the probability of an 
occurrence. No changes are made to the assumptions on which the 
UFSAR accident and transient analyses are based. Therefore, there is 
no reason for an increase in the consequences of any of the analyzed 
conditions which could lead to an increase in Onsite or Offsite dose 
consequences.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence of 
consequences of an accident previously evaluated.
    Does the change create [the] possibility of a new or different 
kind of accident from any previously evaluated?
    No systems, structures, or components are changed by this 
amendment. No procedures that operate, maintain, or surveil them are 
changed. No provisions of the license or the technical 
specifications are modified or relaxed.
    Therefore, the proposed amendment does nor create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Does the change involve a significant reduction in the margin of 
safety?
    No assumptions are changed for any analysis as a result of this 
amendment. No system, structure, or component is changed by this 
amendment. This amendment does not change the results of accident 
and transient analyses previously evaluated.
    Therefore, the proposed amendment does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendment involves 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. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: February 21, 2000.
    Description of amendment request: The proposed amendments would 
change the condensate storage tank (CST) low level setpoint to prevent 
entrainment of air in the high pressure coolant injection (HPCI) pump 
suction line when taking suction from the CST. The amendments would 
also revise the surveillance requirements for the CST level 
instruments.
    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 or consequences of an accident previously evaluated?
    The Condensate Storage Tank (CST) water level and the 
installation of new pressure type switches are not precursors to 
accidents or transients described in the Updated Final Safety 
Analysis Report (UFSAR). The proposed changes will maintain the 
operability of the High Pressure Coolant Injection (HPCI) system, 
thus the HPCI system will continue to function as designed. Any 
failure of the new switches will still cause realignment of the HPCI 
suction from

[[Page 15377]]

the CST to the Torus as currently designed. Therefore, the proposed 
changes in water level and the installation of a new type switch 
will not result in 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?
    For a system to create the possibility of a new and different 
accident, the proposed changes would have to require the system to 
operate in a mode or configuration that is different from the 
original design. The installation of the new switches does not alter 
the current logic configuration. The new switches will continue to 
function and initiate a transfer from the CSTs to the Torus as the 
suction source as originally designed. The proposed changes to the 
Technical Specifications (TS) will ensure that the HPCI suction 
transfer will occur before any air is entrained into the pump 
suction line. This is accomplished by ensuring that the water level 
in the CSTs does not reach the vortex limit before the transfer of 
the HPCI pump suction from the CSTs to the Torus is complete. No new 
functional failure modes will be introduced upon implementation of 
the proposed changes. Therefore, the possibility of a new or 
different kind of accident has not been created.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed changes to the CST Level-Low trip setpoint and 
installation of the new pressure switches provide assurance that air 
entrainment and vortexing will be prevented during HPCI operation. 
By maintaining an increased volume in the CSTs, the probability of a 
HPCI system malfunction due to air entrainment or vortexing is 
decreased. The installation of the new pressure type switches does 
not change the current logic configuration. The new switches will be 
calibrated at a frequency to ensure that the probability of 
unacceptable instrument drift is maintained at an acceptable level. 
Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: February 28, 2000.
    Description of amendment request: The proposed amendments would 
increase the Technical Specification safety limit for the Minimum 
Critical Power Ratio from 1.08 for two loop operation and 1.09 for 
single loop operation to 1.11 and 1.12 respectively. The revised safety 
limits will conservatively bound the current LaSalle Unit 2 operating 
cycle for an anticipated 5 percent power uprate.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes increase the two loop operation Minimum 
Critical Power Ratio (MCPR) Safety Limit from 1.08 to 1.11 and the 
single loop operation MCPR Safety Limit from 1.09 to 1.12. MCPR 
Safety Limits have been established consistent with NRC-approved 
methods to ensure that fuel performance is acceptable. These changes 
do not affect the operability of plant systems, nor do they 
compromise any fuel performance limits. Therefore, the probability 
of an accident will not be changed based on these proposed changes.
    The MCPR Safety Limit is set such that no fuel damage is 
calculated to occur if the limit is not violated. A larger value for 
the MCPR Safety Limit is conservative and bounding for the current 
LaSalle County Station, Unit 2, Cycle 8 core at the current licensed 
power level, because compliance with an MCPR Safety Limit equal to 
or greater than the calculated value will ensure that less than 0.1% 
of the fuel rods experience boiling transition. The MCPR Safety 
Limit does not impact the source term or pathways assumed in 
accidents previously evaluated. Therefore, these proposed changes do 
not increase the consequences of an accident previously evaluated.
    Additionally, operational MCPR limits will be applied that will 
ensure the MCPR Safety Limit is not violated during all modes of 
operation and anticipated operational occurrences in accordance with 
the Core Operating Limits Report (COLR), which will be implemented 
prior to operation at uprated power. The MCPR Safety Limit ensures 
that less than 0.1% of the fuel rods in the core are expected to 
experience boiling transition. Therefore, the probability or 
consequences of an accident will not increase.
    2. Do the proposed changes 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. Changing the MCPR Safety Limit does not alter or add 
any new equipment or change modes of operation. The MCPR Safety 
Limit is established to ensure that 99.9% of the fuel rods avoid 
boiling transition.
    The MCPR Safety Limit is changing for LaSalle County Station, 
Unit 2 to support Cycle 8 operation at uprated power conditions. 
Changing the MCPR Safety Limit does not introduce any physical 
changes to the plant, alter the processes used to operate the plant, 
or change allowable modes of operation. Therefore, no new or 
different kind of accident is created that is different from any 
accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The MCPR Safety Limit provides a margin of safety by ensuring 
that less than 0.1% of the fuel rods are predicted to be in boiling 
transition. The proposed changes increase the two loop operation 
MCPR Safety Limit from 1.08 to 1.11 and the single loop operation 
MCPR Safety Limit from 1.09 to 1.12. A larger value for the MCPR 
Safety Limit is conservative and bounding for the current LaSalle 
County Station, Unit 2 Cycle 8 core at the current licensed power 
level, because compliance with a MCPR Safety Limit equal to or 
greater than what is calculated will ensure that less than 0.1% of 
the fuel rods experience boiling transition. Additionally, the 
proposed changes are being submitted prior to completion of the 
detailed calculations for Cycle 8 power uprate. However, based on 
preliminary calculations, these revised limits are anticipated to 
bound Unit 2 Cycle 8 operation at uprated conditions.

    Therefore, the margin of safety will not be reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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.

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

    Date of amendment request: November 23, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 5.5.11--Ventilation Filter Testing 
Program, which provides the test requirements for charcoal filters, to 
assure compliance with the requirements of American Society for Testing 
and Materials (ASTM) D3803-1989.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 15378]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes will ensure that the Technical 
Specification 5.5.11, Section c, required testing of charcoal 
filters in McGuire ventilation systems designed to meet the guidance 
provided in Regulatory Guide 1.52, Revision 2, are performed as per 
ASTM D3803-1989. This will ensure that these filters are capable of 
performing their design function to maintain offsite and control 
room operator doses within the limits of 10 CFR 100, Subpart A and 
10 CFR 50, Appendix A, GDC [General Design Criteria] 19, following a 
LOCA [Loss-of-Coolant Accident] or a postulated fuel handling 
accident. Consequently, the proposed changes only deal with the 
performance of these systems during an accident and have no impact 
on accident probabilities. In addition, since the proposed changes 
help ensure the capability of the subject ventilation systems to 
perform their design function, there will be no reduction in the 
ability of these systems to minimize the consequences of a 
previously evaluated accident.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed changes only help ensure the performance of the 
subject ventilation systems during an accident and have no impact on 
accident possibilities. No changes are being made to actual plant 
hardware or the way in which the plant is being operated. Therefore, 
no new accident causal mechanisms will be generated. Consequently, 
plant accident analyses will not be affected by these changes.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No. Margin of safety is related to the confidence in the ability 
of the fission product barriers to perform their design functions 
during and following accident conditions. These barriers include the 
fuel cladding, the reactor coolant system, and the containment 
system. The performance of these barriers will not be degraded by 
the proposed changes. In addition, the proposed changes to the 
maximum methyl iodide requirements to accommodate planned changes in 
filter efficiencies will not result in any degradation in the 
capability of the affected charcoal filters to perform their design 
function. As a result of the above, plant safety analyses will not 
be affected by the changes proposed in this LAR [License Amendment 
Request].

    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: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

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

    Date of amendment request: January 6, 2000.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications (TS) 3.3.1--Reactor Trip System (RTS) 
Instrumentation, TS 3.3.2--Engineered Safety Feature Actuation System 
(ESFAS) Instrumentation, TS 3.3.5--Loss of Power Diesel Generator Start 
(LOP) Instrumentation, and TS 3.3.6--Containment Purge and Exhaust 
Isolation (VP) Instrumentation. The proposed revisions will facilitate 
treatment of the applicable RTS, ESFAS, LOP, and VP Instrumentation TS 
Trip Setpoints as nominal values. In addition, proposed changes to the 
applicable TS Bases further define the TS Trip Setpoints as nominal 
values.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes are consistent with the current 
licensing basis for the McGuire Nuclear Station, the setpoint 
methodologies used to develop the Trip Setpoints, the McGuire Safety 
Analyses, and current station calibration procedures and practices. 
The Reactor Trip System and Engineered Safety Features Actuation 
System are not accident initiating systems; they are accident 
mitigating systems. Therefore, these proposed changes will have no 
impact on any accident probabilities. Accident consequences will not 
be affected, as no changes are being made to the plant which will 
involve a reduction in reliability of these systems. Consequently, 
any previous evaluations associated with accidents will not be 
affected by these changes.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. The proposed changes are consistent with the current 
licensing basis for the McGuire Nuclear Station, the setpoint 
methodologies used to develop the Trip Setpoints, the McGuire Safety 
Analyses, and current station calibration procedures and practices. 
No changes are being made to actual plant hardware which will result 
in any new accident causal mechanisms. Also, no changes are being 
made to the way in which the plant is being operated. Therefore, no 
new accident causal mechanisms will be generated. Consequently, 
plant accident analyses will not be affected by these changes.
    3. Does this change involve a significant reduction in a margin 
of safety?
    No. The proposed changes are consistent with the current 
licensing basis for the McGuire Nuclear Station, the setpoint 
methodologies used to develop the Trip Setpoints, the McGuire Safety 
Analyses, and current station calibration procedures and practices. 
Margin of safety is related to the confidence in the ability of the 
fission product barriers to perform their design functions during 
and following accident conditions. These barriers include the fuel 
cladding, the reactor coolant system, and the containment system. 
The performance of these barriers will not be degraded by the 
proposed changes. Consequently, plant safety analyses will not be 
affected by these changes.

    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: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: Richard L. Emch, Jr.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: November 23, 1999, as supplemented by 
letter dated February 24, 2000
    Description of amendment request: The proposed amendment would 
incorporate the use of American Society for Testing and Materials 
(ASTM) D3803-1989, ``Standard Test Method for Nuclear-Grade Activated 
Carbon,'' into the Technical Specifications (TSs). Entergy Operations, 
Inc. (the licensee) is submitting this proposed amendment as a complete 
response to Nuclear Regulatory Commission (NRC) Generic Letter (GL) 99-
02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The 
February 24, 2000, supplement proposes additional changes to the TSs to 
ensure that ventilation system velocity requirements are established in 
accordance with the standards of ASTM D3803-1989. This application was 
previously noticed in the Federal Register on March 8, 2000 (65 FR 
12291).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 15379]]

consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    Deleting portions of applicable ANO-1 [Arkansas Nuclear One, 
Unit 1] TSs that reference system design velocity criteria for 
activated charcoal medium testing requires no physical change to 
plant design. NRC GL 99-02, in support of the ASTM D3803-1989 
standard, requires licensees to utilize charcoal testing methods 
that will ensure the current license basis, as it relates to General 
Design Criterion (GDC) 19, is maintained. The existing criterion 
within the affected ANO-1 TSs is less restrictive than that of ASTM 
D3803-1989 standard and, therefore, is being proposed for deletion. 
The testing of charcoal mediums has no impact on the probability of 
an accident occuring. However, the charcoal mediums do act to reduce 
radioiodines released to the environment during and following an 
accident. Testing the charcoal mediums to a more restrictive 
standard, however, does not increase the consequences of an accident 
since such testing ensures the current analyses remain valid.
    Therefore, the proposed changes do not involve a signficant 
increase in the probability or consequences of any accident 
previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident from any Previously Evaluated

    As stated previously, the proposed changes to the ANO-1 TSs do 
not result in any physical change to plant design, nor does the 
testing of charcoal mediums act to create a new or different 
accident than that previously analyzed. The existing criterion 
within the affected ANO-1 TSs is less restrictive than that of ASTM 
D3803-1989 standard and, therefore, is being proposed for deletion. 
Testing criteria governing the operability of charcoal mediums is 
not considered an accident initiator of new, different, or 
previously analyzed accidents. The charcoal mediums act solely to 
reduce radioiodines released to the environment during and following 
accident scenarios.
    Therefore, the proposed changes do not create the possiblity of 
a new or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    Testing of charcoal mediums to more restrictive criteria acts to 
better ensure that these mediums will perform their design function 
during and following accidents that result in a release of 
radioiodines. No reduction in the margin to safety can be construed 
based on the new testing criteria. The charcoal mediums will 
continue to remove radioiodines as originally designed and approved 
by the NRC during and following accidents involving radioactive 
release.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 27, 2000.
    Description of amendment request: The proposed amendment would 
revise the Arkansas Nuclear One, Unit 2 (ANO-2) technical 
specifications (TS) by providing actions associated with inoperable 
control room emergency ventilation or cooling systems during movement 
of irradiated fuel during shutdown modes of operation, when allowed 
outage times associated with these systems are not met.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated

    The inclusion of additional actions within the ANO-2 TSs 
associated with the control room emergency ventilation and air 
conditioning systems during the handling of irradiated fuel does not 
require any physical modification to plant components or systems. 
Implementing the proposed actions act to ensure the operability of 
the remaining system, eliminate the reliance on automatic actuation 
where applicable, and ensure that any active failure will be readily 
detected. The proposed changes, therefore, act to ensure [that] the 
consequences of a fuel handling accident are mitigated and have no 
impact on the probability [of] a fuel handling accident occurring. 
The proposed actions are in addition to those currently required by 
the ANO-2 TSs and, therefore, are more restrictive.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident from any Previously Evaluated

    The inclusion of additional actions within the ANO-2 TSs 
associated with the control room emergency ventilation and air 
conditioning systems during the handling of irradiated fuel does not 
require any physical modification to plant components or systems. 
Implementing the proposed actions act to ensure the operability of 
the remaining system, eliminate the reliance on automatic actuation 
where applicable, and ensure that any active failure will be readily 
detected. The proposed changes, therefore, are not relevant to 
creating new or different kinds of accidents than those previously 
evaluated. The proposed actions are in addition to those currently 
required by the ANO-2 TSs.
    Therefore, this change does not create the possibility or a new 
or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    The inclusion of additional actions within the ANO-2 TSs 
associated with the control room emergency ventilation and air 
conditioning systems during the handling of irradiated fuel act to 
ensure the operability of the remaining system, eliminate the 
reliance on automatic actuation where applicable, and ensure that 
any active failure will be readily detected. The proposed changes, 
therefore, act to maintain the margin of safety by ensuring the 
operability of redundant equipment that is required to protect 
control room personnel in the event of a fuel handling accident. The 
proposed actions are in addition to those currently required by the 
ANO-2 TSs and, therefore, are more restrictive.
    Therefore, 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.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: February 24, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.4.11 on reactor coolant system 
vent flow verification, TS 4.6.1.1.a on containment penetration closure 
verification (non-automatic), and TS 4.6.3.1.2 on containment isolation 
valve actuation verification. These TS surveillances require testing to 
be performed during Modes 5 and/or 6. The proposed change will 
eliminate unnecessary mode restrictions on these surveillance 
requirements.

[[Page 15380]]

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

    Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated
    Current regulation requires the licensee to responsibly plan, 
schedule, and perform testing of station equipment. Furthermore, the 
philosophies of the RSTS [Revised Standard Technical Specifications] 
do not restrict surveillance performance to specific modes of 
operation or other plant conditions. Deletion of the mode 
restrictions will not relinquish licensee responsibility from 
prudent planning, scheduling, and performance of testing activities 
and may provide the licensee lower-risk periods of opportunity for 
test performance. Because of this, the proposed changes are 
considered to be administrative in nature and do not significantly 
affect the plant or personnel safety. Modes in which surveillances 
are performed are not analyzed in association with accident 
probability or the consequences of an accident. The proposed changes 
reduce unnecessary restrictions on the licensee and provide 
consistency with the philosophies of the RSTS.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind 
of Accident from any Previously Evaluated

    The licensee will continue to be accountable for proper and 
prudent planning, scheduling, and performance of surveillance 
activities in the absence of the aforementioned mode restrictions 
proposed for deletion. Therefore, the proposed changes are 
considered to be administrative in nature and do not significantly 
affect the plant or personnel safety. The probability of a new or 
different kind of accident being created remains unchanged since the 
licensee currently is required to properly plan and execute 
surveillance tests, even within specific modes of operation. Other 
activities presently ongoing during the currently specified 
operational modes could result in an unexpected or unforseen 
transient or condition if surveillance testing is not properly 
planned and executed given the other activities in progress and 
current plant conditions. Since the responsibility of the licensee 
in these matters remains unchanged by the proposed changes, the 
possibility of a new or different kind of accident being created 
also remains unchanged.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of 
Safety

    The licensee will continue to be accountable for proper and 
prudent planning, scheduling, and performance of surveillance 
activities in the absence of the aforementioned mode restrictions 
proposed for deletion. Therefore, the proposed changes are 
considered to be administrative in nature and do not significantly 
affect the plant or personnel safety. The margin to safety remains 
unchanged since the licensee currently is required to properly plan 
and execute surveillance tests, even within specific modes of 
operation. Other activities presently ongoing during the currently 
specified operational modes could result in an unexpected or 
unforseen transient or condition if surveillance testing is not 
properly planned and executed given the other activities in progress 
and current plant conditions. Since the responsibility of the 
licensee in these matters remains unchanged by the proposed changes, 
no significant reduction in the margin to safety is evident.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 21, 2000.
    Description of amendment request: Entergy Operations, Inc. requests 
revision of the Grand Gulf Nuclear Station licensing basis and 
Technical Specifications to utilize the alternative accident source 
term described in NUREG-1465.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    This proposed amendment to the Grand Gulf Nuclear Station (GGNS) 
Technical Specifications (TS) revises those specifications affected 
by the implementation of the alternative source term concepts in 
accordance with NUREG-1465. In addition, based on the alternative 
source term, changes are proposed to selected specifications 
associated with handling irradiated fuel in the primary or secondary 
containment and CORE ALTERATIONS. Specifically, the proposal uses a 
new term to describe irradiated fuel that contains sufficient 
fission products to require operability of accident mitigation 
systems to meet the accident analysis assumptions. The alternative 
source term changes affect the definitions and the specifications 
for the Control Room Fresh Air System, MSIV [main steam isolation 
valve] leakage surveillance, Standby Gas Treatment System 
surveillance, and revises a license condition to increase the 
allowable control room inleakage. The specifications affected by the 
relaxation of the shutdown controls include those for the Control 
Room HVAC [heating, ventilation, and air conditioning] system, and 
the electrical AC [alternating current] Sources, DC [direct current] 
Sources and Distribution Systems during shutdown.
    The Commission has provided standards for determining whether a 
no significant hazards consideration exists as stated in 
10CFR50.92(c). A proposed amendment to an operating license involves 
a no significant hazards consideration if operation of the facility 
in accordance with the proposed amendment would not: (1) involve a 
significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a 
new or different kind of accident from any accident previously 
evaluated; or (3) involve a significant reduction in a margin of 
safety.
    Entergy Operations, Inc. has evaluated the no significant 
hazards considerations in its request for a license amendment. In 
accordance with 10CFR50.91(a), Entergy Operations, Inc. is providing 
the analysis of the proposed amendment against the three standards 
in 10CFR50.92(c). A description of the no significant hazards 
considerations determination follows:
    1. The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated.
    The alternative source term does not affect the design or 
operation of the facility; rather, once the occurrence of an 
accident has been postulated the new source term is an input to 
evaluate the consequences. The implementation of the alternative 
source term has been evaluated in revisions to the analyses of the 
limiting design basis accidents at Grand Gulf Nuclear Station. Based 
on the results of these analyses, it has been demonstrated that, 
even with the requested Technical Specification and Operating 
License changes, the dose consequences of these limiting events are 
within the regulatory guidance currently proposed by the NRC 
[Nuclear Regulatory Commission] for use with the alternative source 
term. This guidance is presented in NUREG-1465, in the draft 
rulemaking for 10CFR50.67, and in the associated draft Regulatory 
Guide DG-1081.
    A new term to describe irradiated fuel is used to establish 
operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analysis. Because the equipment affected by the revised operational 
conditions is not considered an initiator to any previously analyzed 
accident, inoperability of the equipment cannot increase the 
probability of

[[Page 15381]]

any previously evaluated accident. The proposed requirements bound 
the conditions of the current design basis fuel handling accident 
analysis which concludes that the radiological consequences are 
within the acceptance criteria of NUREG-0800, Section 15.7.4 and 
General Design Criteria [GDC] 19. As noted above, with the 
alternative source term implementation, the acceptance criteria are 
also being revised. The results of the revised Fuel Handling 
Accident demonstrate that the dose consequences are within the 
currently proposed NRC regulatory guidance. This guidance is 
presented in NUREG-1465, in the draft rulemaking for 10CFR50.67, and 
in the associated draft Regulatory Guide DG-1081.
    Therefore, the proposed changes do not significantly increase 
the probability or consequences of any previously evaluated 
accident.
    2. The proposed changes would not create the possibility of a 
new or different kind of accident from any previous[ly] analyzed.
    The alternative source term does not affect the design, 
functional performance, or operation of the facility or of any 
equipment within the facility. Similarly, it does not affect the 
design or operation of any equipment or systems involved in the 
mitigation of any accidents. The proposed changes to the Technical 
Specifications and the Operating License, while they revise certain 
performance requirements, do not involve any physical modifications 
to the plant. Therefore, the proposed changes associated with the 
alternative source term do not create the possibility of a new or 
different kind of accident from any previous[ly] analyzed.
    The new term to describe irradiated fuel is used to establish 
operational conditions where specific activities represent 
situations where significant radioactive releases can be postulated. 
These operational conditions are consistent with the design basis 
analyses. The relaxation of selected shut down controls has been 
modeled in revised analyses. The proposed changes do not introduce 
any new modes of plant operation and do not involve physical 
modifications to the plant. Therefore, the proposed changes related 
to shutdown controls based on the alternative source term do not 
create the possibility of a new or different kind of accident from 
any previous[ly] analyzed.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The changes above are basically associated with the 
implementation of a new licensing basis for Grand Gulf Nuclear 
Station. Approval of the basis change from the original source term 
in accordance with TID-14844 to the new alternative source term of 
NUREG-1465 is requested by this submittal. The results of the 
accident analyses revised in support of this submittal, and 
considering the requested Technical Specification and Operating 
License changes, are subject to revised acceptance criteria. These 
analyses have been performed using conservative methodologies as 
outlined in the currently proposed regulatory guidance. Safety 
margins and analytical conservatisms have been evaluated and are 
well understood. The analyzed events have been carefully selected 
and margin has been retained to ensure that the analyses adequately 
bound all postulated event scenarios. The dose consequences of these 
limiting events are within the acceptance criteria also found in the 
latest regulatory guidance. This guidance is presented in NUREG-
1465, in the approved rulemaking for 10CFR50.67, and in the 
associated draft Regulatory Guide DG-1081.
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries as well as control 
room, are within the corresponding regulatory limit. In a similar 
way, the results of the existing analyses demonstrated that the dose 
consequences were within the applicable NRC-specified regulatory 
limit. Specifically, the margin of safety for these accidents is 
considered to be that provided by meeting the applicable regulatory 
limit, which, for most events, is conservatively set below the 
10CFR100 limit. With respect to the control room personnel doses, 
the margin of safety is the difference between the 10CFR100 limits 
and the regulatory limit defined by 10CFR50, Appendix A, Criterion 
19 (GDC 19).
    Therefore, because the proposed changes continue to result in 
dose consequences within the applicable regulatory limits, they are 
considered to not result in a significant reduction in a margin of 
safety.
    Based on the above evaluation, operation in accordance with the 
proposed amendment involves no significant hazards considerations.

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

    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston 
and Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-
3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 24, 2000.
    Description of amendment request: Entergy Operations, Inc. 
requests revisions to the Grand Gulf Nuclear Station Technical 
Specifications which specify the minimum useable fuel oil 
inventories to be maintained in the Division 1, 2, and 3 Diesel 
Generator Fuel Oil Storage Tanks.
    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:
    Entergy has evaluated this proposed Technical Specification 
change and has determined that it involves no significant hazards 
consideration. This determination has been performed in accordance 
with the criteria set forth in 10CFR50.92. The following evaluation 
is provided for the three categories of the significant hazards 
consideration standards:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change would require additional fuel oil to be stored in 
each of the Division 1, 2, and 3 Diesel Generator Fuel Oil Storage 
Tanks. The amount of diesel fuel required to be kept in the storage 
tanks, which has been determined by Calculation MC-Q1P75-90190 
Revision 2 and Calculation MC-Q1P81-90188 Revision 2, is well within 
the maximum capacity of the Diesel Generator Fuel Oil Storage Tanks. 
As stated in UFSAR [Updated Final Safety Analysis Report] Section 
9.5.4.3 (Safety Evaluation for the diesel fuel storage subsystem) `* 
* * the tank level will be above the ``seven-day capacity'' required 
level and will be kept as near the top as practical.'' Other fuel 
oil storage subsystem components, such as the transfer pumps, are 
similarly designed, as a minimum, for the storage tanks being filled 
to maximum capacity. The Diesel Generator Fuel Oil Storage Tanks 
continue to meet the original design requirements as described in 
the UFSAR. The proposed change will provide adequate fuel for diesel 
generator operation at the Technical Specification surveillance 
testing capacity for Division 1 and 2 Diesel Generators, 5740 KW, 
and the nameplate rating for Division 3 Diesel Generator, 3300 KW, 
rather than the lower post-LOCA [loss-of-coolant accident] load 
profiles previously assumed. Therefore, increasing the quantity of 
fuel oil required to be maintained, will not increase the 
probability of the diesel generators becoming an initiator for any 
previously evaluated accident. Furthermore, since the proposed 
change increases the fuel oil inventory it should enhance the 
ability of the diesel generators to respond to an accident and as 
such the change does not increase the consequences of any previously 
analyzed accident.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The Diesel Generator Fuel Oil subsystem design and operation 
will not change except for the incorporation of increased fuel oil 
inventory requirements. This proposed increase remains within the 
maximum capacity of the Diesel Generator Fuel Oil Storage Tanks. 
Existing analyses and evaluations, concerning the fuel oil storage 
tanks, are not adversely impacted by this increase in the required 
fuel oil inventory. Other fuel oil storage subsystem components, 
such as the transfer pumps, are similarly designed, as a minimum, 
for the storage tanks being filled to maximum capacity. The 
subsystem continues to meet the original design requirements. The 
proposed increased fuel oil inventory cannot adversely affect any 
other equipment. Therefore, since the proposed change only increases 
the fuel oil

[[Page 15382]]

inventory requirements and does not result in any change in the 
response of any equipment to an accident, the proposed change does 
not create the possibility of a new or different kind of accident 
from any previously analyzed accident.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Existing Technical Specification 3.8.3 bases state the Diesel 
Generator Fuel Oil Storage Tank minimum level is sufficient to 
operate the respective Diesel Generator for seven days while 
supplying maximum post-LOCA demands. The proposed change increases 
the quantity of fuel oil required to be maintained in each of the 
Division 1, 2, and 3 Diesel Generator Fuel Oil Storage Tanks. The 
proposed change will provide adequate fuel for diesel generator 
operation at the Technical Specification surveillance testing 
capacity for Division 1 and 2 Diesel Generators, 5740 KW, and the 
nameplate rating for Division 3 Diesel Generator, 3300 KW, rather 
than the lower post-LOCA load profiles previously assumed. The 
amount of diesel fuel required to be kept in the storage tanks, 
which has been determined by Calculation MC-Q1P75-90190 Revision 2 
and Calculation MC-Q1P81-90188 Revision 2, is well within the 
maximum capacity of the Diesel Generator Fuel Oil Storage Tanks. 
Therefore, since the proposed change increases the fuel oil 
inventory it should enhance the ability of the diesel generators to 
respond to an accident and as such the change does not decrease any 
margin of safety previously assumed.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: January 27, 2000.
    Description of amendment request: The proposed amendment would 
allow operation of the facility for a period of up to 12 hours with the 
temperature of the ultimate heat sink (UHS) between 75 and 77 deg.F, 
provided water temperature is verified below 77 deg.F at least once per 
hour. Currently the temperature limit is 75 deg.F and is verified at 
least once per 6 hours when the temperature is above 70 deg.F, or once 
per 24 hours below 70 deg.F.
    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:

    In accordance with 10 CFR 50.92 NNECO [Northeast Nuclear Energy 
Company] has reviewed the proposed change and has concluded that it 
does not involve a significant hazards consideration (SHC). The 
basis for this conclusion is that the three criteria of 10 CFR 
50.92(c) are not compromised. The proposed change does not involve a 
SHC because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will allow plant operation to continue for 
an additional 12 hours with the temperature of the UHS up to 2 deg.F 
above the Technical Specification limit of 75 deg.F. This increase 
in UHS temperature will not affect the normal operation of the plant 
to the extent which would make any accident more likely to occur. In 
addition, there exists adequate margin in the safety systems and 
heat exchangers to assure the safety functions are met at the higher 
temperature. An evaluation has confirmed that safe shutdown will be 
achieved and maintained for a loss of coolant accident (LOCA) with a 
loss of normal power (LNP) and a single active failure with an UHS 
water temperature as high as 77 deg.F.
    The proposed change will have no adverse effect on plant 
operation, or the availability or operation of any accident 
mitigation equipment. The plant response to the design basis 
accidents will not change. In addition, the proposed change can not 
cause an accident. Therefore, there will be no significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change will allow plant operation to continue for 
an additional 12 hours with the temperature of the UHS up to 2 deg.F 
above the Technical Specification limit of 75 deg.F. This will not 
alter the plant configuration (no new or different type of equipment 
will be installed) or require any new or unusual operator actions. 
The proposed change will not alter the way any structure, system or 
component functions and will not significantly alter the manner in 
which the plant is operated. There will be no adverse effect on 
plant operation or accident mitigation equipment. The proposed 
change does not introduce any new failure modes. Also, the response 
of the plant and the operators following these accidents is 
unaffected by the change. In addition, the UHS is not an accident 
initiator. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed change will allow plant operation to continue for 
an additional 12 hours with the temperature of the UHS up to 2 deg.F 
above the Technical Specification limit of 75 deg.F. An evaluation 
has been performed which demonstrates that the safety systems have 
adequate margin to ensure their safety functions can be met with an 
ultimate heat sink water temperature of 77 deg.F. In addition, safe 
shutdown capability has been demonstrated for an UHS water 
temperature as high as 77 deg.F.
    The proposed change will have no adverse effect on plant 
operation or equipment important to safety. The plant response to 
the design basis accidents will not change and the accident 
mitigation equipment will continue to function as assumed in the 
design basis accident analysis. Therefore, there will be no 
significant reduction in a margin of safety.
    The proposed change does not alter the design, function, or 
operation of the equipment involved. The impact of the proposed 
change has been analyzed, and it has been determined it does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated, and does not involve a significant reduction in a margin 
of safety. Therefore, NNECO has concluded the proposed change does 
not involve a SHC.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: May 26, 1999.
    Description of amendment request: The proposed amendments would 
relocate Technical Specification (TS) Surveillance Requirement 
4.1.3.5.b, regarding the performance of channel functional test and 
channel calibration of certain control rod scram accumulator 
instrumentation, to the Updated Final Safety Analysis Report and would 
revise TS 3.1.3.5 to allow an alternate method for verifying whether a 
control rod drive pump is operating.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

[[Page 15383]]

    The first proposed change relocates control rod drive (CRD) 
instrumentation requirements from the TS to the UFSAR and plant 
procedures. The second proposed change adds an alternate method for 
verifying operation of a control rod drive pump in the TS action 
statement.
    Regarding the first proposed change, operability of the 
accumulators is determined by verifying that the pressure in each 
accumulator is greater than or equal to 955 psig. TS 4.1.3.5.a 
requires weekly verification of accumulator pressure. The local 
pressure indicator for each accumulator is the normal means of 
satisfying this surveillance. This proposed change does not affect 
or alter the requirements associated with this instrumentation. If 
the local pressure indicator is not functioning or pressure is less 
than 955 psig, the accumulator will still be declared inoperable.
    Operability of the accumulator pressure or water level alarm and 
indication function provided by the Reactor Manual Control System 
(RMCS) is not critical to the ability to insert control rods 
because:
    (1) The rods can be inserted with normal charging water pressure 
if the accumulator is inoperable;
    (2) A controlled shutdown or scram would occur before the 
accumulator would lose its full capability to insert the control 
rod, if it is found that no control rod drive pumps are operating 
according to existing procedural and TS controls placed on the 
plant; and
    (3) The subject instruments' alarm and indication function are 
part of routine operational monitoring and are not considered in the 
plant safety analysis.
    [Therefore, the removal of the accumulator pressure or level 
indication does not impact the consequences or probability of an 
accident previously evaluated. The operational monitoring of the 
accumulator alarms and indication system affords operating personnel 
the status of system condition and the opportunity to initiate 
appropriate actions if deemed necessary.]
    The second proposed change simply adds an alternate method for 
verifying operation of a control rod drive pump. This check provides 
an equivalent method of verifying that inoperable control rod 
accumulators were not caused by a control rod drive pump trip. In 
addition:
    (1) The assumed control rod reactivity insertion rate is not 
changed;
    (2) The maximum number of inoperable accumulators and control 
rods is not changed;
    (3) The TS actions to be taken in the event that a control rod 
drive pump is not operating remain unchanged; and
    (4) The instrumentation for accumulator leakage a pressure 
detection will continue to be maintained and calibrated.
    A RMCS failure does not change the failure modes or the 
reliability of the control rod function as described and evaluated 
in the UFSAR. The CRD system will continue to be available to safely 
shutdown the plant as described and evaluated in the UFSAR.
    Therefore, these proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Neither the mechanism for initiating nor for carrying out a 
scram is modified by either of these proposed changes. These 
proposed changes do not:
    (1) Create a means by which the scram function could be impeded 
or prevented.
    (2) Involve a physical plant alteration or change the methods 
governing normal plant operation.
    (3) Impose or eliminate any requirements or change the controls 
for maintaining the requirements.
    There are no other malfunctions that need to be considered since 
failure of a significant number of control rods to scram is analyzed 
in Section 15.8 of the UFSAR. This is the bounding analysis for 
multiple control rod malfunctions or severe degradation of control 
rod scram performance. This event is mitigated by safety systems not 
directly related to the CRD system including the scram accumulators.
    Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The first proposed change relocates CRD instrumentation 
requirements from TS to the UFSAR and plant procedures. The proposed 
change will not reduce a margin of safety, because it has no impact 
on any safety analysis. * * * [Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.]
    The second proposed change adds an alternate method for 
verifying operation of a control rod drive pump in the TS action 
statement. This proposed change does not reduce a margin of safety 
because the proposed change does not:
    (1) Affect the maximum allowable control rod scram times,
    (2) Change the maximum allowable number or minimum separation of 
inoperable control rods, or
    (3) Modify any of the instrument setpoints or functions.
    The proposed change will either maintain the present margin of 
safety or increase it, by reducing the need for unnecessary 
challenges to the Reactor Protection System (RPS) and resulting 
plant shutdown, while still maintaining the capability to complete a 
reactor scram.
    Therefore, these proposed TS changes do not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: February 3, 2000.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) by revising the reactor water 
level setpoint for the Anticipated Transient Without Scram 
Recirculation Pump Trip (ATWS-RPT) function and the Alternate Rod 
Injection (ARI) functions (Table 3.2-7).
    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:

    Operation of the FitzPatrick plant in accordance with the 
proposed amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS change deals only with an instrumentation 
setpoint which initiates the ATWS-RPT/ARI function. The system is 
intended to provide a mitigation function during a postulated ATWS 
event and does not provide any other plant control function. 
However, if the ATWS-RPT/ARI system were to fail, the result would 
be a trip of the recirculation pumps, or reactor scram, both of 
which are currently evaluated. The design of the system includes a 
one-out-of-two-twice logic, which ensures that a single failure in 
the system cannot cause or inhibit the ATWS-RPT/ARI function. 
Therefore, the probability of an inadvertent recirculation pump trip 
or inadvertent reactor scram is not changed from the event as 
currently described in the JAFNPP UFSAR [James A. FitzPatrick 
Nuclear Power Plant Updated Final Safety Analysis Report].
    FitzPatrick specific analyses were performed by General Electric 
Company with NRC approved methods for postulated ATWS events 
(Reference 1 [``James A. FitzPatrick Nuclear Power Plant Anticipated 
Transient Without Scram Analysis, for Recirculation Pump Trip 
Setpoint Changes,'' General Electric Company, NEDC-32616P, July 18, 
1996, Previously Docketed with NRC]). The specific events evaluated 
include the Main Steamline Isolation Valve closure event, 
Inadvertent Opening of a Relief Valve, and the Loss of Feedwater. 
For these events, the following acceptance criteria were 
established:
    Peak Reactor Pressure (maximum 1 SRV out of service)-- 1500 psig

[[Page 15384]]

    Peak Suppression Pool Temperature-- 190 deg.F
    Fuel Remains Cooled--Coolant Level > TAF [Top of Active Fuel]
    The analyses demonstrate that all criteria were adequately met 
with the proposed TS change implemented, further ensuring no 
increase in the consequences of the postulated events.
    The basis for changing the ARI initiation setpoint on reactor 
level to be consistent with that proposed for the ATWS RPT is 
documented in Reference 2 [JAF-ICD-NBI-;03998, Rev. 0--Alternate Rod 
Insertion Setpoint (an internal FitzPatrick interface document)]. 
The ARI initiation point is not specified in the Technical 
Specification.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed TS change deals only with a reactor water level 
instrumentation setpoint, which initiates the ATWS-RPT/ARI function. 
The existing level transmitters and wiring will be used, and new 
analog trip units will be incorporated which are identical to 
existing low-low reactor water level trip units currently shared 
with HPCI [High Pressure Coolant Injection] and RCIC [Reactor Core 
Isolation Cooling] initiation. These new analog trip units are of a 
different design (General Electric) than those used in the Reactor 
Protection System (Rosemount) and therefore, the diversity 
requirement of 10 CFR 50.62 (c)(3) remain[s] satisfied. This allows 
the HPCI and RCIC setpoints to remain the same while only lowering 
the ATWS-RPT/ARI setpoint. The sensing, logic and actuation of the 
ATWS-RPT/ARI design is not modified. This includes the use of the 
existing one-out-of-two-twice logic, which ensures that a single 
failure in the circuit will not cause or inhibit the ATWS-RPT/ARI 
function. There are no new signals required as input, and the trip 
function is accomplished with the existing RPT breakers and existing 
scram pilot air header solenoid valves. The system does not provide 
input to any other plant function. The plant will not operate in any 
new mode nor are there any new operational requirements as a result 
of the proposed change. Therefore, it is not considered possible for 
the ATWS-RPT/ARI system to fail in any new or different way from 
those events currently evaluated in the JAFNPP UFSAR.
    3. Involve a significant reduction in a margin of safety.
    The ATWS-RPT/ARI function protects the fuel, reactor and 
containment from failure during a postulated ATWS event. The fuel 
cladding barrier is protected via adequate cooling, provided by 
ensuring that the core remains covered throughout the entire event. 
The reactor coolant system boundary is protected by ensuring 
compliance with the ASME [American Society of Mechanical Engineers] 
emergency class pressure limit of 120% of design pressure. The 
containment is protected by ensuring the suppression pool 
temperature limits are met.
    FitzPatrick specific ATWS analyses were performed by postulating 
events that challenge each of these limits (Reference 1). With the 
proposed TS change considered, each of these limits were met without 
a need for any reduction in the margin of safety established in the 
JAFNPP UFSAR for the primary fission product barriers.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Section Chief: Marsha Gamberoni, Acting.

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

    Date of amendment request: February 24, 2000.
    Description of amendment request: The proposed amendment would 
approve a revision to the Hope Creek Generating Station Updated Final 
Safety Analysis Report (UFSAR) to reflect the use of the Mechanical 
Vacuum Pumps (MVPs) to evacuate the condenser during plant startup at 
power levels less than or equal to 5%. These revisions are required to 
make the UFSAR accident analyses associated with a Control Rod Drop 
Accident (CRDA) consistent with actual plant operation. Public Service 
Electric and Gas Company (PSE&G) has performed an engineering 
calculation that demonstrates that there is an increase in the 
radiological consequences of a CRDA coincident with MVP operation. 
Nuclear Regulatory Commission (NRC) approval of the proposed UFSAR 
changes is required, in accordance with Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.59, since these changes involve an 
unreviewed safety question.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The Condenser Air Removal System has no safety-related function 
and its failure does not jeopardize the function of any safety-
related system or component or prevent a safe shutdown of the plant. 
Neither the MVPs, nor other components associated with the Condenser 
Air Removal, Gaseous Radwaste Off-Gas, Process Radiation Monitoring, 
or Turbine Building HVAC [Heating, Ventilation, and Air 
Conditioning] systems or the South Plant Vent are design basis 
accident initiators. The operation of mechanical vacuum pump at 
power levels  5% will not increase the probability of 
occurrence of a main condenser air removal system leak or failure of 
the line leading to the steam jet air ejector (SJAE) near the main 
condenser. Additionally, the design and operation of the condenser 
off-gas system is not impacted. Moreover, MVP operation will not 
increase the probability of occurrence of a CRDA or any other design 
basis accident. Consequently, this proposal does not increase the 
probability of an accident previously evaluated.
    The engineering calculation performed to assess the impact of 
the use of the MVPs demonstrated that the radiological consequences 
of a CRDA coincident with MVP operation increase but remain well 
within the 10CFR100 guidelines and meet SRP [Standard Review Plan] 
Section 15.4.9, Appendix A, acceptance criteria. Additionally, the 
calculation demonstrated that the radiological consequences of a 
CRDA coincident with MVP operation are within the GDC [General 
Design Criterion] 19 guidelines for control room personnel and plant 
operators and remain bounded by the loss of coolant accident 
analysis for on-site personnel. Therefore, although the proposal 
does increase the consequences of a CRDA, the proposal does not 
significantly increase the 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 proposal involves crediting manual action to trip the MVPs; 
however, PSE&G has evaluated this operator action against the 
criteria in NRC Information Notice 97-78 and has concluded that 
adequate controls are in place to ensure that the subject manual 
action is taken. In addition, the proposal does not change monitor 
setpoints, affect equipment qualification, or otherwise create an 
accident initiator not previously considered. Consequently, this 
proposal does not create the possibility of an accident of a 
different type from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The Condenser Air Removal System has no safety-related function. 
Failure of the system does not jeopardize the function of any 
safety-related system or component or prevent a safe shutdown of the 
plant.
    The radiological activity evaluated in this proposal does not 
result in scenarios that could impact 10 CFR 50 Appendix I, 10 CFR 
20, or 40 CFR 190 release criteria. Post-scram shutdown or startup 
condition MVP operation in accordance with plant operating 
procedures will not degrade the original design for the Condenser 
Air Removal System.
    An engineering calculation was prepared that demonstrated that 
the radiological consequences of a CRDA coincident with MVP 
operation remain well within the 10 CFR 100 guidelines and that the 
consequences meet SRP Section 15.4.9, Appendix A, acceptance 
criteria. Additionally, the engineering calculation demonstrated 
that the radiological

[[Page 15385]]

consequences of a CRDA coincident with MVP operation are within GDC 
19 guidelines for control room personnel and plant operators and 
remain bounded by the loss of coolant accident analysis for on-site 
personnel.
    Since no design bases are degraded, the Technical Specifications 
operating limits, that provide sufficient operating range such that 
the acceptance limits are not exceeded during plant operations and 
analyzed transients, are not [ ] affected. Since the acceptance 
limits are not exceeded, implementation of this proposal does not 
reduce the margin of safety as described in the basis for any 
Technical Specifications.

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

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

    Date of amendment requests: February 23, 2000 (PCN 508).
    Description of amendment requests: The amendment application is a 
request to allow an option regarding the methodology for measuring the 
reactivity worth of control element assembly (CEA) groups for San 
Onofre Nuclear Generating Station (SONGS) Units 2 and 3 during low-
power physics testing following a refueling. The proposed option 
involves measuring the worth of approximately three-fourths of the 
full-length CEA groups each refueling cycle rather than the present 
methodology, which measures the worth of all full-length CEA groups 
each refueling cycle. Measured CEA groups would be rotated such that 
each full-length group would be measured at least every other 
refueling. The licensee has determined this change to involve an 
unreviewed safety question.
    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 proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    No. The proposed option to the Low Power Physics Test (LPPT) 
program will involve performance of rod worth measurements of 
typically six of eight full-length control element assembly (CEA) 
groups each refueling, rather than performance of rod worth 
measurements of all eight CEA groups each refueling. Thus, the LPPT 
option will result in a reduction in the number of plant 
manipulations required for LPPT. Inverse Boron Worth (IBW) is not 
required in the proposed LPPT program option, but it may be 
determined during the performance of a boration or dilution, which 
is already a part of the present LPPT program. The manipulations 
which will be performed are a subset of the evolutions which are 
performed in the existing test sequence. Therefore, the LPPT testing 
option does not carry any increased risk of any accident evaluated 
in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR). 
Since the number and duration of manipulations are reduced, there 
would actually be a small reduction in accident potential.
    The proposed test program option will not compromise the 
technical objectives of the LPPT program. Fuel fabrication, core and 
reactor internals reassembly, CEA worths, mechanical integrity and 
reliability, performance of core physics design codes and 
consistency with design and safety analysis expectations will remain 
validated with the same effectiveness as is achieved in the current 
program. In addition, the reduced duration of operation in the LPPT 
Special Test Exception of the Technical Specifications has a 
positive impact on nuclear safety.
    Therefore, the proposed LPPT program option does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed test program option will eliminate CEA exchange 
measurements and determine CEA worth by dilution/boration 
measurements. Measurement of CEA worth by the dilution/boration 
methods achieves typically higher quality results than the CEA 
Exchange method.
    The proposed LPPT program option does not include the 
requirement to measure inverse boron worth. However, a measured 
initial critical boron concentration and measured CEA group worths 
that match predicted values within acceptance criteria are 
sufficient to verify adequate core physics modeling without a 
separate IBW measurement.
    Since the proposed test sequence option continues to ensure that 
core operation and reactivity control are consistent with design 
expectations, the proposed LPPT option will not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Therefore, the proposed LPPT program option does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Does the amendment request create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    No. The proposed LPPT program option does not create any plant 
condition or manipulation which is materially different from those 
of the existing program. Furthermore, the number of manipulations 
and duration of Special Test Exceptions are significantly reduced. 
The proposed LPPT program option relies entirely on conventional 
boration and dilution rod worth measurement test methods which have 
been industry standards. The methodology used to measure IBW, if 
performed, does not introduce any new evolutions during LPPT and 
cannot create a new or different type of accident.
    Therefore, the proposed LPPT program option does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Does this amendment request involve a significant reduction 
in a margin of safety?
    No. The proposed LPPT program option fully achieves objectives 
of the reload test program by validating fuel fabrication, core 
reassembly, CEA worths, mechanical integrity and reliability, 
performance of physics design codes and consistency with design and 
safety analysis expectations with the same effectiveness as is 
achieved in the current program. As a result, all assumptions made 
in support of UFSAR Chapter 15 Safety Analyses regarding CEA 
performance remain valid.
    The effectiveness of the SONGS 2 & 3 Reload Test program, 
including LPPT and Power Ascension Testing, has been evaluated and 
shown to be uncompromised by the proposed LPPT option. Specific 
testing requirements imposed by the Nuclear Regulatory Commission 
are captured in Technical Specification Surveillance Requirements. 
The proposed LPPT program option is fully compliant with existing 
Technical Specification Surveillance Requirements and validates the 
core physics models regarding core performance, reactivity control 
and proper core reassembly to an extent equivalent to that of the 
present program.
    The proposed LPPT program option is also consistent with the 
recently modified ANSI/ANS 19.6.1-1997 standard for Pressurized 
Water Reactor reload testing, with the exception of the requirement 
and methodology to determine IBW. The ANSI/ANS standard was 
developed with participation from industry and NRC representatives 
and represents an expert panel assessment of what is appropriate for 
an LPPT program. A measured initial critical boron concentration and 
measured CEA group worths that match predicted values within 
acceptance criteria are sufficient to verify adequate core physics 
modeling, and infer that the IBW value is within standard acceptance 
critieria, without a separate IBW measurement.
    Therefore, the proposed LPPT program option does not involve a 
significant reduction in a margin of safety.

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

[[Page 15386]]

    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, 
California 91770.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: August 24, 1999, as supplemented on 
December 29, 1999.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) 3.3.2 ``Engineered Safety 
Feature Actuation System (ESFAS) Instrumentation'' to relax the slave 
relay test frequency from quarterly to a refueling frequency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The results of WCAP-13878 demonstrate that slave relays are 
highly reliable. WCAP-13878 also provides guidance to assure that 
slave relays remain highly reliable. The aging assessment concludes 
that the age/temperature-related degradation of all ND relays, and 
NE relays produced after 1992, is sufficiently slow such that a 
refueling frequency surveillance interval will not significantly 
increase the probability of slave relay failures. Finally, the 
evaluation of the auxiliary relays actuated during slave relay 
testing has concluded that based on the tests of the auxiliary 
relays performed during other equipment testing, reasonable 
assurance is provided that failures will be identified if the 
associated slave relays are tested on a refueling frequency.
    Therefore, the proposed changes do 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 alter the performance of the ESFAS 
mitigation systems assumed in the plant safety analysis. Changing 
the interval for periodically verifying ESFAS slave relays (assuring 
equipment operability) will not create any new accident initiators 
or scenarios.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated for VEGP.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not affect the total ESFAS response 
assumed in the safety analysis since the reliability of the slave 
relays will not be significantly affected by the decreased 
surveillance frequency.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: Richard L. Emch, Jr.

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, 
Unit 2, Matagorda County, Texas

    Date of amendment request: February 21, 2000.
    Description of amendment request: STP Nuclear Operating Company 
proposes to amend the South Texas Project (STP), Unit 2 technical 
specifications (TS) so that steam generator tube eddy-current 
inspection indications of less than or equal to 3.0 volts can be left 
in service if found at intersections of tube hot-leg tube-support-
plates C through M (3.0-volt alternate repair criteria). The new 
alternate repair criteria would apply only until the Unit 2 Model E 
steam generators are replaced during the outage currently scheduled to 
commence in fall of 2002. STP Nuclear Operating Company also proposes 
to amend the STP, Unit 2 TS to make an editorial correction to Note 1 
and Note 2, on page 3/4-16a to align the notes with the preceding 
paragraph. STP Nuclear Operating Company also provided, for information 
only, changes to the Bases for TS 3/4.4.5 to provide the structural 
margins and Westinghouse topical report references used as the bases 
for the use of the 3.0-volt alternate repair criteria.
    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:

    In accordance with the criteria set forth in 10 CFR 50.92, the 
STP Nuclear Operating Company (STPNOC) has evaluated these proposed 
Technical Specification changes and determined they do not represent 
a significant hazards consideration. Conformance of the proposed 
amendment to the standards for a determination of no significant 
hazard as defined by the criteria set forth in 10 CFR 50.92 is shown 
in the following discussions addressed to each criterion:
    (1) Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    During the limiting design-basis steam-line-break (SLB) event, 
South Texas Project (STP) Unit 2 steam generator tube burst criteria 
are inherently satisfied for marginally degraded (primarily axially-
oriented ODSCC [outer diameter stress corrosion cracking]) tube 
spans at certain tube support plate (TSP) intersections.
    Steam generator tubes pass through holes drilled in the TSP. The 
inside diameter (ID) of the drilled holes closely approximates the 
outside diameter (OD) of the tubes. Generally, the TSP precludes 
those tube spans within the drilled holes from deforming beyond the 
diameters of the drilled holes, thus, precluding tube burst in the 
restrained regions. However, design basis SLB events may vertically 
displace a TSP, removing its support from the tube spans passing 
through it. For TSP C through M, maximum displacement during a 
postulated SLB event is less than 0.15 inch. Because TSP C through M 
remain essentially stationary during all conditions, tube spans 
included within the drilled holes are restrained during the limiting 
SLB event. Thus, the tube burst margin for intersections of tube 
hot-legs and TSP C through M is independent of voltage related 
growth rates and the proposed 3-volt ARC [alternate repair criteria] 
is compliant with RG [Regulatory Guide] 1.121 [Bases for Plugging 
Degraded PWR Steam Generator Tubes] criteria.
    Given a TSP displacement of  0.15 inch, tube hot-leg spans 
enclosed within TSP C through M have a negligible tube burst 
probability of less than 10-\10\ for a single tube. This 
is eight orders of magnitude less than the 10-\2\ 
probability-of-burst criterion specified by GL [Generic Letter] 95-
05 [Voltage-Based Repair Criteria for Westinghouse Steam Generator 
Tubes Affected by Outside Diameter Stress Corrosion Cracking] and 
represents negligible axial tube burst probabilities for tube hot-
leg spans intersecting TSP C through M. Thus, repair limits to 
preclude burst are not needed and tube repair limits may be based 
primarily on limiting leakage to acceptable levels during accident 
conditions.
    Cracks that include cellular corrosion may yield to axial loads, 
resulting in tensile tearing of the tube at that location. A tensile 
load requirement to prevent this establishes a structural limit for 
the tube expansion based plugging criterion. In order to establish a 
lower bound for the structural limit, tensile tests were used to 
measure the force required to separate a tube that exhibits cellular 
corrosion. Additionally, pulled tubes with cellular and/or inter-
granular attack (IGA) tube wall degradation were evaluated and the 
tensile strength of the tube conservatively calculated from the 
remaining non-corroded cross-section of the tube. This calculation 
assumes that the degraded portions contribute nothing to the axial 
load carrying ability of the tube. Data from these tests shows that 
circumferential cracks exhibiting

[[Page 15387]]

bobbin-coil-probe-indication-voltages greater than 35 volts require 
tube-pressure-differentials well above the operating limit of 3-
times-normal differential pressure in order to produce 
circumferential ruptures (i.e., axial separation at the plane of the 
crack). This proposal specifies a structural limit of 17 volts 
(safety factor of 2) to ensure conservative results for repairs at 
intersections of tubes with TSP C through M.
    GL 95-05 states that licensees must perform SLB leak rate and 
tube burst probability analyses before returning to power from 
outages during which they perform steam generator inspections. 
Licensees must include the results in a report to the NRC within 90 
days after restart. If an analysis reveals that leak-rate or burst-
probability exceeds limits, the licensee must report it to the NRC 
and assess the safety significance of this finding. Model E steam 
generator SLB leak rates are calculated for indications found at 
intersections of tube hot-legs and TSP. Both SLB leak rate and tube 
burst probability are calculated for tube hot-leg intersections with 
FDB [flow distribution baffles], hot-leg intersections with TSP N 
through R, and indications found at intersections of tube cold-legs 
with any TSP.
    It has been established that the design basis main SLB outside 
of containment and upstream of the MSIV [main steam isolation 
valves] produces the limiting radiological consequence from any tube 
leakage that may be postulated to exist at the initiation of an 
accident. With use of 3-volt ARC, STPNOC [STP Nuclear Operating 
Company] will calculate the maximum primary-to-secondary leakage for 
the last day of the coming steam generator service-cycle and use 
this value to calculate the radiological consequence of the limiting 
SLB event. This methodology will ensure that site boundary doses for 
this accident remain within an acceptable fraction of the 10 CFR 100 
guidelines and that doses to the control room operators remain 
within GDC 19 [10 CFR Part 50, Appendix A, General Design Criterion] 
limits.
    Based on the above, STPNOC concludes that operation of South 
Texas Project Unit 2 in accordance with the proposed license 
amendment 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?
    Use of the proposed steam generator tube 3-volt ARC does not 
significantly change circumstances or conclusions assumed by the 
plant design basis. Application of the 3-volt ARC does not 
significantly increase the probability of either single or multiple 
tube ruptures. Steam generator tube integrity remains adequate for 
all plant operating conditions.
    STPNOC has confirmed that the allowed post-accident primary-to-
secondary leakage rate for SLB events results in the limiting 
offsite and control room doses for South Texas Project Unit 2. A 
projected SLB leak rate of 15.4 gpm is calculated to produce doses 
90% of the currently licensed South Texas Project Unit 2 dose limits 
(Reference 2 [STPNOC letter dated July 15, 1998, NOC-AE-000228, 
Response to NRC Request for Additional Information related to STP 
Unit 2 Amendment No. 83]). STPNOC TS impose a normal leak rate limit 
of 150 gpd (0.1 gpm) per steam generator to minimize the potential 
for excessive leakage during all plant conditions. The 150 gpd limit 
provides added margin to accommodate contingent leakage should a 
stress corrosion crack grow at a greater than expected rate or 
extend outside the TSP. Leakage trending consistent with EPRI Report 
TR-04788, ``PWR Primary-to-Secondary Leak Guidelines'' has been 
established for South Texas Project Unit 2.
    Since steam generator tube integrity will meet GL 95-05 
requirements and be confirmed through in-service inspection and 
primary-to-secondary leakage monitoring, the proposed license 
amendment does not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    (3) Does this change involve a significant reduction in a margin 
of safety?
    RG 1.121 describes a method for meeting GDC 14, 15, 31, and 32 
by reducing the probability or consequences of steam-generator tube-
rupture through application of criteria for removing degraded tubes 
from service. These criteria set limits of degradation for steam 
generator tubing through in-service inspection. Analyses show that 
tube integrity will remain consistent with the criteria of 
Regulatory Guide 1.121 after implementation of the proposed 3-volt 
ARC. Even under the worst case ODSCC occurrence at TSP elevations, 
3-volt ARC will not cause or significantly increase [the] 
probability of a steam-generator tube-rupture event.
    In addressing combined LOCA [loss-of-coolant accident] + SSE 
[safe-shutdown earthquake] effects on steam generator components as 
required by GDC 2, analysis has shown that tube collapse may occur 
in certain regions of the steam generators of some plants. This 
collapse is caused by TSP plastic deformation in the region of the 
TSP wedge supports. Plastic deformation occurs when TSP experience 
large lateral loads concentrated at wedge support points on the 
periphery of a TSP undergoing combined loading effects of a LOCA 
rarefaction wave and SSE. Deformation impinges on TSP apertures 
through which tubes pass, deflecting tube walls inward. The 
resulting pressure differential across deformed tube walls may cause 
some tubes to collapse.
    There are two issues associated with steam generator tube 
collapse. First, collapse of steam generator tubing reduces RCS 
[reactor coolant system] flow. RCS flow reduction increases 
resistance to heat flow from the core during a LOCA, increasing Peak 
Clad Temperature (PCT). Second, partial through-wall tube-cracks 
could become full through-wall tube-cracks during tube deformation 
or collapse. Tubes in regions affected by this phenomenon are 
usually excluded from evaluation under 3-volt ARC. STP Model E steam 
generator design does not produce this plastic deformation, thus is 
not subject to tube collapse. No STP Unit 2 tubes are excluded, for 
this reason, from application of the proposed 3-volt ARC.
    End of Cycle (EOC) distribution of crack indications at affected 
TSP elevations will be confirmed to allow no more than the 
acceptable primary-to-secondary leakage rate during all plant 
conditions and not adversely affect radiological dose consequences. 
For the limiting SLB event, STPNOC will calculate leak rates as 
free-span leakage for ODSCC indications at tube and TSP 
intersections. The calculations will use GL 95-05 leak rate methods 
with an additional component for potentially overpressurized 
indications [discussed in detail in the Safety Evaluation section of 
the licensee's February 21, 2000, application under the heading 
``SLB Leak Rate and Tube Burst Probability Considerations''].
    Inspections conducted in accordance with RG 1.83, Rev. 1 [In 
Service Inspection of Pressurized Water Reactor Steam Generator 
Tubes], using 3-volt ARC for intersections of tube hot-legs with TSP 
C through M, and using 1-volt ARC at remaining hot-leg and cold-leg 
intersections will be supplemented by:
    (1) enhanced eddy current inspection procedures to achieve 
consistency in voltage normalization,
    (2) eddy current inspection of 100% of tubes found, using 
inspection of a 20% tube sample, to have ODSCC at intersections with 
TSP, and
    (3) a required RPC [rotating pancake coil] inspection of the 
larger indications to confirm that the principal degradation 
mechanism continues to be ODSCC.
    Plugging steam generator tubes reduces RCS flow margin. As 
previously noted, increasing repair limits for indications found at 
TSP intersections will reduce the number of tubes that must be 
plugged. Thus, 3-volt ARC will conserve RCS flow margin, preserving 
operational and safety benefits that would otherwise be reduced by 
unnecessary plugging.
    Therefore, the proposed license amendment does not result in a 
significant increase in dose consequences represented in the current 
licensing basis, and does not involve a significant reduction in 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. The staff also reviewed the proposed editorial change for no 
significant hazards consideration. The proposed editorial correction 
does not affect the design or operation of the facility and satisfies 
the three standards of 10 CFR 50.92(c). Therefore, the NRC staff 
proposes to determine that the request for amendments involves no 
significant hazards consideration.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

[[Page 15388]]

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia

    Date of amendment request: November 29, 1999.
    Description of amendment request: The proposed changes will modify 
the Technical Specifications (TS) in Section 3.23 for the Main Control 
Room and Emergency Switchgear Room Ventilation and Air Conditioning 
Systems; TS Surveillance Requirement Sections 4.20, Basis 4.20.A.7, and 
4.20.B.4 for the Control Room Air Filtration System; and TS 
Surveillance Requirement Sections 4.12.A.6, 4.12.A.7, 4.12.A.8, 
4.12.B.7, and 4.12 Basis for the Auxiliary Ventilation Exhaust Filter 
Trains. The proposed changes will revise the above Surveillance 
Requirements for the laboratory testing of the carbon samples for 
methyl iodide removal efficiency to be consistent with American Society 
for Testing and Materials (ASTM) Standard D3803-1989, ``Standard Test 
Method for Nuclear-Graded Activated Carbon,'' with qualification, as 
the laboratory testing standard for both new and used charcoal 
adsorbent used in the ventilation system.
    Basis for proposed no significant hazards consideration 
determination: In 10 CFR 50.92, three criteria are provided to 
determine whether a proposed license amendment involves a significant 
hazards consideration. No significant hazards consideration is involved 
if operation of the facility 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. 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:

    Virginia Electric and Power Company has reviewed the 
requirements of 10 CFR 50.92 as they relate to the proposed changes 
for Surry Units 1 and 2 and determined that a significant hazards 
consideration is not involved. The proposed Technical Specification 
changes adopt the nuclear-grade charcoal testing requirements of 
ASTM D3803-1989, with qualification, for methyl iodide removal 
efficiency and the requirements of ASTM D3803-1979, with 
qualification, for elemental iodine removal efficiency. The method 
of testing nuclear-grade activated charcoal does not affect the 
design or operation of the plant. The changes also do not involve 
any physical modification to the plant or result in a change in a 
method of system operation. The adoption of the 1989 edition of ASTM 
D3803 for methyl iodide testing conforms with approved guidance for 
testing of nuclear-grade activated charcoal. This provides assurance 
that testing of ventilation systems is being performed with a 
suitable standard to ensure that charcoal adsorbers are capable of 
performing their required safety function and that the regulatory 
requirements regarding onsite and offsite dose consequences continue 
to be satisfied. The changes do not create an unreviewed safety 
question.
    (a) The proposed changes modify surveillance testing 
requirements and do not affect plant systems or operation and 
therefore do not increase the probability or the consequences of an 
accident previously evaluated. The proposed surveillance 
requirements adopt ASTM D3803-1989, with qualification, as the 
laboratory method for testing samples of the charcoal adsorber for 
methyl iodide removal efficiency in response to NRC's Generic Letter 
99-02. This method of testing charcoal adsorbers has been approved 
by the NRC as an acceptable method for determining methyl iodide 
removal efficiency. Since the charcoal adsorbers are used to 
mitigate the consequences of an accident, the more accurate the 
test, the better assurance we have that we remain within our 
accident analysis assumptions. Testing of the charcoal adsorbers' 
efficiency for removing elemental iodine is performed in accordance 
with the 1979 version of ASTM D3803 since the 1989 version does not 
address elemental iodine removal efficiencies. The laboratory test 
acceptance criteria contain a safety factor to ensure that the 
efficiency assumed in the accident analysis is still valid at the 
end of the operating cycle. There is no change in the method of 
plant operation or system design.
    (b) The proposed changes modify surveillance testing 
requirements and do not impact plant systems or operations and 
therefore do not create the possibility of an accident or 
malfunction of a different type than evaluated previously. The 
proposed surveillance requirements adopt ASTM D3803-1989, with 
qualification, as the laboratory method for testing samples of the 
charcoal adsorber for methyl iodide removal efficiency. This change 
is in response to NRC's request in Generic Letter 99-02. Testing of 
the charcoal adsorbers' efficiency for removing elemental iodine is 
performed in accordance with the 1979 version of ASTM D3803 since 
the 1989 version does not address elemental iodine removal 
efficiencies. There is no change in the method of plant operation or 
system design. There are no new or different accident scenarios, 
transient precursors, nor failure mechanisms that will be 
introduced.
    (c) The proposed changes modify surveillance test requirements 
and do not impact plant systems or operations and therefore do not 
significantly reduce the margin of safety. The revised surveillance 
requirements adopt ASTM D3803-1989, with qualification, as the 
laboratory method for testing samples of the charcoal adsorber for 
methyl iodide removal efficiency. The 1989 edition of this standard 
imposes very stringent requirements for establishing the capability 
of new and used activated carbon to remove methyl iodide from air 
and gas streams. The results of this test provide a more 
conservative estimate of the performance of nuclear-graded activated 
carbon used in nuclear power plant HVAC [heating, ventilation, and 
air conditioning] systems for the removal of methyl iodide. Testing 
of the charcoal adsorbers' efficiency for removing elemental iodine 
is performed in accordance with the 1979 version of ASTM D3803 since 
the 1989 version does not address elemental iodine removal 
efficiencies. The laboratory test acceptance criteria contain a 
safety factor to ensure that the efficiency assumed in the accident 
analysis is still valid at the end of the operating cycle.

    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: Mr. Donald P. Irwin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Section Chief: Richard L. Emch, Jr.

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.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: February 18, 2000.
    Brief description of amendment request: The amendment changes 
current Technical Specification (TS) 4.9a.2 and improved TS 3.7.5 and 
its associated bases to remove requirements associated with the backup 
steam supply to turbine-driven auxiliary feedwater pump P-8B.

[[Page 15389]]

    Date of publication of individual notice in Federal Register: March 
1, 2000 (65 FR 11089)
    Expiration date of individual notice: Comment period expired March 
14, 2000; Notice period expires March 31, 2000.

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

    Date of application for amendments: February 25, 2000.
    Brief description of amendments: The amendment revises Technical 
Specification Table 3.3.2-1, ``Engineered Safety Feature Actuation 
System Instrumentation'' to provide a one-time exception, until the 
next time the turbine is removed from service, from the requirement to 
perform response time testing for the solenoid valve 1-FSV-47-027.
    Date of publication of individual notice in the Federal Register: 
March 2, 2000.
    Expiration date of individual notice: March 16, 2000.

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

    Date of application for amendment: January 14, 2000, as 
supplemented by letter dated February 17, 2000 (ULNRC-04172 and -
04187).
    Brief description of amendment request: The amendment would revise 
several sections of the improved Technical Specification (ITSs) to 
correct 14 editorial errors made in either (1) the application dated 
May 15, 1997, (and supplementary letters) for the ITSs, or (2) the 
certified copy of the ITSs that was submitted in the licensee's letters 
of May 27 and 28, 1999. The ITSs were issued as Amendment No. 133 by 
the staff in its letter of May 28, 1999, and will be implemented by the 
licensee to replace the current TSs by April 30, 2000.
    Date of publication of individual notice in Federal Register: 
February 25, 2000 (65 FR 10118).
    Expiration date of individual notice: March 27, 2000.

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).

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

    Date of application for amendment: March 1, 1999.
    Brief description of amendment: The amendment approves changes to 
the Updated Safety Analysis Report concerning design requirements for 
physical protection from tornado missiles.
    Date of issuance: February 29, 2000.
    Effective date: February 29, 2000.
    Amendment No.: 124.
    Facility Operating License No. NPF-62: The amendment allows a 
change to the Updated Safety Analyis Report concerning tornado missile 
protection.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19558).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 29, 2000.
    No significant hazards consideration comments received: No.

AmerGen Energy Co., LLC, Docket No. 50-289, Three Mile Island Nuclear 
Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: June 4, 1999, as supplemented 
December 13, 1999.
    Brief description of amendment: The amendment modified the limiting 
conditions for operation in the Technical Specifications (TSs) under 
which a reduction in the number of means of decay heat removal (DHR) 
capability may occur by deleting two of these conditions. The amendment 
also makes related Bases changes and clarifies the DHR requirements for 
redundancy.
    Date of issuance: February 28, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 220.
    Facility Operating License No. DPR-50. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35207). The December 13, 1999, letter withdrew a Bases change of the 
June 4, 1999, application and did not change the initial proposed no 
significant hazards consideration determination or expand the amendment 
beyond the scope of the initial notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2000.
    No significant hazards consideration comments received: No.

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of application for amendment: May 26, 1999.
    Brief description of amendment: The amendment authorized changes to 
Chapters 5 and 14 of the Updated Final Safety Analysis Report (UFSAR). 
The changes reflect the use of an Electric Power Research Institute-
developed Conservative Deterministic Failure Margin methodology for 
seismic analysis of the portions of the nonsafety-related auxiliary 
steam line piping located in the Auxiliary, Control, and Fuel Handling 
buildings at TMI-1.
    Date of issuance: March 10, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 221.
    Facility Operating License No. DPR-50. Amendment authorizes changes 
to the UFSAR.

[[Page 15390]]

    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35207). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 10, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 8, 1999, as supplemented 
July 20 and November 24, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications to increase the storage capacity of spent fuel 
in the fuel storage pools by allowing credit for soluble boron and 
decay time in the safety analysis, and to increase the maximum radially 
averaged fuel enrichment from 4.3 weight percent to 4.8 weight percent.
    Date of issuance: March 2, 2000.
    Effective date: March 2, 2000.
    Amendment Nos.: Unit 1-125, Unit 2-125, Unit 3-125.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 20, 1999 (64 
FR 50835). The July 20 and November 24, 1999, letters provided 
additional clarifying information that was within the scope of the 
original application and Federal Register notice and did not change the 
staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 2, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: September 28, 1999.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) in response to your submittal dated September 28, 
1999. The amendment revises TS 2.1.1.2, ``Reactor Core Safety Limits,'' 
and TS 5.6.5, ``Core Operating Limits Report,'' by removing safety 
limit restrictions which are no longer applicable.
    Date of issuance: March 1, 2000.
    Effective date: March 1, 2000.
    Amendment No.: 207.
    Facility Operating License No. DPR-71: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59797).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 2000.
    No significant hazards consideration comments received: No.

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.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) in response to the licensee's submittal dated 
September 28, 1999. The amendment revises TS 2.1.1.2, ``Reactor Core 
Safety Limits,'' by changing the Minimum Critical Power Ratio.
    Date of issuance: March 1, 2000.
    Effective date: March 1, 2000.
    Amendment No.: 208.
    Facility Operating License No. DPR-71: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70080).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 23, 1999, as supplemented 
on October 21, 1999, and December 15, 1999.
    Brief description of amendments: The amendments approved the 
installation of new Boral high density spent fuel storage racks at 
Byron and Braidwood stations. The amendments also approved an increase 
in the spent fuel pool storage capacity from 2,870 assemblies to 2,984 
assemblies at each station.
    Date of issuance: March 1, 2000.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 112 and 105.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32280). The October 21 and December 15, 1999, supplements did not 
change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 1, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 12, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 2.2, ``Limiting Safety System Settings,'' and TS 3/
4.1.A, ``Reactor Protection System,'' to remove an anticipatory reactor 
scram signal, the turbine electro-hydraulic control (EHC) low oil 
pressure trip, from the reactor protection system trip function 
requirements.
    Date of issuance: January 28, 2000.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 193 & 189.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67331).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 16, 1999.
    Brief description of amendments: The amendments change Technical 
Specification Table 4.1.A-1, ``Reactor Protection System 
Instrumentation Surveillance Requirements,'' to modify the surveillance 
requirements for Functional Unit 3, ``Reactor Vessel Steam Dome 
Pressure--High,'' to reflect replacement of the pressure switches with 
analog trip units.
    Date of issuance: January 28, 2000.
    Effective date: Immediately, to be implemented before startup from 
Refueling Outage 16 for Unit 1 and before startup from Refueling Outage 
15 for Unit 2.
    Amendment Nos.: 194 & 190.

[[Page 15391]]

    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70082).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 28, 2000.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: June 2, 1999, as supplemented 
August 25, 1999.
    Brief description of amendment: The amendment allows for the 
relocation of the Quality Assurance related administrative controls to 
the Quality Assurance Program Description in accordance with NRC 
Administrative Letter 95-06, ``Relocation of Technical Specification 
Administrative Controls Related to Quality Assurance.''
    Date of issuance: February 25, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 206.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 3, 1999 (64 FR 
59799).
    The August 25, 1999, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 25, 2000.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: February 29, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.7.D.1 to correct an editorial error, TS 6.2.2 to 
change the senior reactor operator license requirement for the 
Operations Manager, and TS 6.3.1 to modify the qualification 
requirement for the Operations Manager.
    Date of issuance: February 29, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 207.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17023).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 29, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: June 24, 1999, as supplemented 
by letter dated November 24, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications by revising the minimum reactor coolant system 
(RCS) flow rate limit, the reactor coolant average temperature, and the 
pressurizer pressure limits, and by restricting operation to a RCS flow 
deficit of no more than one percent.
    Date of issuance: March 1, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--184; Unit 2--176.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43770).
    The November 24,1999, letter provided clarifying information that 
did not change the scope of the June 24, 1999, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 1, 2000.
    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: June 24, 1999, as supplemented 
by letter dated November 24, 1999.
    Brief description of amendments: The amendments revise the minimum 
reactor coolant system (RCS) flow rate limit, reduce the reactor 
coolant average temperature and pressurizer pressure limits, restrict 
operation to a RCS flow deficit of no more than one percent, and change 
the low RCS flow reactor trip setpoint.
    Date of issuance: March 2, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--191; Unit 2--172.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43772).
    The November 24, 1999, supplemental letter did not expand the scope 
of the application initially noticed or change the proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 2, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 13, 1999.
    Brief description of amendment: The amendment removes footnote (d) 
from Function 5, ``RHR [residual heat removal] SDC [shut down cooling] 
System Isolation'' of Technical Specification (TS) Table 3.3.6.1-1, 
``Primary Containment Isolation Instrumentation.'' 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 trip system valve, RHR-V-8, is no longer transferred 
to the alternate remote shutdown panel and is now required during Modes 
1, 2 and 3. Therefore, footnote (d) is no longer needed. Footnote (e) 
is relettered as footnote (d) for consistency.
    Date of issuance: March 9, 2000.
    Effective date: March 9, 2000, to be implemented within 30 days of 
issuance.
    Amendment No.: 161.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70082).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2000.
    No significant hazards consideration comments received: No.

[[Page 15392]]

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: December 16, 1999.
    Brief description of amendment: The amendment authorizes the 
licensee to revise fuel handling accident (FHA) dose calculations for 
three scenarios described in the River Bend Station, Unit 1, Updated 
Safety Analysis Report. The first is an FHA in the fuel building, 
assumed to occur 24 hours post-shutdown. A second FHA analysis was 
prepared to support Amendment 35 to RBS Technical Specifications (TS) 
which assumed an FHA occurs in the primary containment 80 hours post-
shutdown during local leakage rate testing (LLRT). A third analysis was 
prepared in support of Amendment 85 to the River Bend Station Technical 
Specifications which assumed the containment is open at 11 days. These 
analyses are being updated to account for several changes that were 
determined by the licensee to involve an unreviewed safety question in 
accordance with Title 10 of the Code of Federal Regulations, Section 
50.59(a)(2)(i).
    Date of issuance: March 2, 2000.
    Effective date: The license amendment is effective as of its date 
of issuance and shall be implemented in the next periodic update to the 
USAR in accordance with 10 CFR 50.71(e). Implementation of the 
amendment is the incorporation into the USAR update, the changes to the 
description of the facility as described in the licensee's application 
dated December 16, 1999, and evaluated in the staff's Safety Evaluation 
attached to this amendment.
    Amendment No.: 110.
    Facility Operating License No. NPF-47: The amendment authorized 
changes to the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4272).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 2000.
    No significant hazards consideration comments received: No.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Units 1 and 2, Pope County, Arkansas

    Date of amendment request: September 17, 1999.
    Brief description of amendments: The amendments modify TS 3.25.2, 
``Radioactive Gas Storage Tanks,'' at Arkansas Nuclear One, Unit 1 
(ANO-1) and TS 3/4.11.2, ``Gas Storage Tanks,'' at Arkansas Nuclear 
One, Unit 2 (ANO-2). This change will reduce the limiting condition for 
operation for the maximum quantity of stored radioactivity per tank 
from 300,000 curies of noble gases as Xenon-133 (Xe-133) equivalent to 
78,782 curies of noble gases as Xe-133 equivalent at ANO-1, and 82,400 
curies of noble gases as Xe-133 equivalent at ANO-2.
    Date of issuance: February 18, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: ANO-1--204; ANO-2--211.
    Facility Operating License Nos. DPR-51 and NPF-6: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 12, 2000 (65 FR 
1921).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 18, 2000.
    No significant hazards consideration comments received: No.

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
    Brief description of amendment: The amendment replaces the existing 
reference to the Asea Brown Boveri-Combustion Engineering, Inc. small 
break loss-of-coolant accident emergency core cooling system 
performance evaluation model with the revised model described in the 
topical report CENPD-137, Supplement 2, P-A, April 1998.
    Date of issuance: March 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment No.: 158.
    Facility Operating License No. NPF-38: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70085).
    The July 29, October 28, and November 11, 1999, letters provided 
additional information that did not change the scope of the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2000.
    No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412, 
Beaver Valley Power Station, Unit No. 2, Shippingport, Pennsylvania

    Date of application for amendments: March 16, 1999.
    Brief description of amendments: This amendment revised TS 3/
4.7.1.3 and associated Bases for the Primary Plant Demineralized Water 
(PPDW) system to clarify that the minimum specified volume of water in 
the PPDW Storage Tank is a usable volume. Additionally, the minimum 
usable volume of water in the PPDW Storage Tank is increased, and a 
clarifying footnote that the specified value is an analysis value is 
added. Finally, several editorial and administrative changes, such as 
revision of action statement wording, addition of license number to TS 
page, and addition of clarifying information to the TS Bases regarding 
analysis assumptions are made.
    Date of issuance: February 28, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 106.
    Facility Operating License No. NPF-73: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999, (64 FR 
19556).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 27, 1999.
    Brief description of amendments: The amendments relocate the 
seismic monitoring instrumentation requirements contained in Technical 
Specification (TS) 3/4.3.3.3 to the Licensing Requirements Manual (LRM) 
based on the guidance provided in Generic Letter 95-10, ``Relocation of 
Selected Technical Specifications Requirements Related to 
Instrumentation.'' The Bases section for Specification 3/4.3.3.3 is 
also relocated to the LRM. The appropriate Index pages, Table Index 
page (Unit No. 1 only), TS pages and Bases pages are revised to reflect 
the removal of the seismic monitoring instrumentation specification 
from the TSs. An additional TS page is added to reflect that TS Number 
3/4.3.3.4 is not used.

[[Page 15393]]

This additional page also denotes the number of the following page. 
Finally, the Bases section is modified to denote that TS Number 3/
4.3.3.4 is not used.
    Date of issuance: February 28, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 228 and 107.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35203).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: May 27, 1999.
    Brief description of amendments: The amendments (1) revised the 
frequency for performing the CHANNEL FUNCTIONAL TEST of the manual 
initiation functional units specified in the Beaver Valley Power 
Station, Unit Nos. 1 and 2, Engineered Safety Features Actuation System 
(ESFAS) Instrumentation Technical Specifications (TSs) from monthly, 
with an accompanying footnote which allows the manual initiation to be 
tested on a refueling interval, to each refueling interval; (2) revise 
footnotes associated with TS ESFAS tables; (3) revise associated TS 
Bases.
    Date of issuance: February 28, 2000.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 229 and 108.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35205).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 21, 1999, as supplemented by 
submittals dated December 1, 1999, and January 28, 2000.
    Brief description of amendment: This amendment revises the 
Technical Specifications to expand the present spent fuel storage 
capability by 289 storage locations by allowing the use of spent fuel 
racks in the cask pit area adjacent to the spent fuel pool.
    Date of issuance: February 29, 2000.
    Effective date: February 29, 2000.
    Amendment No.: 237.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 8, 1999 (64 FR 
36933).
    The supplemental information contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 29, 2000.
    No significant hazards consideration comments received: No.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: August 18, 1999.
    Brief description of amendment: This amendment decreases the 
surveillance frequency, listed in the updated Final Safety Analysis 
Report (UFSAR), for cycling steam valves in the turbine overspeed 
protection system from monthly to quarterly.
    Date of Issuance: February 28, 2000.
    Effective Date: As of the date of its issuance, to be incorporated 
into the UFSAR at the time of its next update.
    Amendment No.: 108.
    Facility Operating License No. NPF-16: Amendment revised the UFSAR.
    Date of initial notice in Federal Register: September 22, 1999 (64 
FR 51345).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 2000.
    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: December 1, 1999, as 
supplemented December 15, 1999.
    Breif description of amendments: The amendments revised License 
Condition 3.L for Turkey Point, Units 3 and 4, Operating Licenses DPR-
31 and DPR-41 to reflect the December 1, 1999, date of the last 
revision to the Physical Security Plan. Also, the phrase ``Turkey Point 
Plant, Units 3 and 4 Security Plan'' was revised to ``Turkey Point 
Physical Security Plan.''
    Date of issuance: February 28, 2000.
    Effective date: February 28, 2000.
    Amendment Nos.: 204 and 198.
    Facility Operating License Nos. DPR-31 and DPR-41: Amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73092).
    The Commssion's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 28, 2000.
    No significant hazards consideration comments received: No.

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 amendment revises the Technical 
Specifications, Appendix B, ``Environmental Protection Plan (Non-
Radiological)'' to incorporate the reasonable and prudent measures, and 
the terms and conditions, of the Incidental Take Statement in the 
Biological Opinion issued by the National Marine Fisheries Service.
    Date of issuance: February 29, 2000.
    Effective date: February 29, 2000.
    Amendment No.: 190.
    Facility Operating License No. DPR-31: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70090).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 29, 2000.
    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: December 22, 1999.
    Brief description of amendments: The amendments delete Technical 
Specification 5.4.2, ``Reactor Coolant System Volume,'' regarding the 
reactor coolant system (RCS) volume information. Information concerning 
the RCS volume is included in the D. C. Cook Updated Final Safety 
Analyses Report (UFSAR), and any changes to the information are 
controlled in accordance with 10 CFR 50.59.

[[Page 15394]]

    Date of issuance: March 1, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 241 and 222.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 13, 2000 (65 FR 
2199).
    The Commssion's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 1, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: October 6, 1999, as supplemented 
February 9, 2000.
    Brief description of amendment: The amendment addresses the 
following changes to the Technical Specifications: (1) provisions for 
implementation of 10 CFR Part 50, Appendix J, Option B, (Technical 
Specification Task Force (TSTF) Change 52, Revision 2) (2) extension of 
the required surveillance interval for the containment air lock 
interlock mechanism from 18 to 24 months (TSTF Change 17, Revision 1), 
(3) clarification of the valve types requiring isolation time testing 
(TSTF Change 46, Revision 1), and (4) provisions for use of 
administrative means for verification of isolation devices that are 
locked, sealed or otherwise secured (TSTF Change 269, Revision 2).
    Date of issuance: March 3, 2000.
    Effective date: March 3, 2000, to be implemented within 30 days.
    Amendment No.: 180.
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73092). The February 9, 2000, supplement provided clarifying 
information that was within the scope of the October 6, 1999, 
application and the staff's original Federal Register notice and did 
not change the staff's initial proposed no significant hazards 
consideration determination.
    The Commssion's related evaluation of the amendment is contained in 
a Safety Evaluation dated March 3, 2000.
    No significant hazards consideration comments received: No.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station, Unit 1, Oswego County, New York

    Date of application for amendment: August 26, 1999, as supplemented 
December 17, 1999.
    Brief description of amendment: The amendment changes Technical 
Specification 3.2.3, ``Coolant Chemistry,'' to support the 
implementation of noble metal chemical addition.
    Date of issuance: March 8, 2000.
    Effective date: As of the date of issuance to be implemented before 
the licensee first performs the noble metal chemical addition.
    Amendment No.: 169.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 22, 1999 (64 
FR 51347).
    The licensee's supplemental letter dated December 17, 1999, did not 
change the Commission's finding of no significant hazards 
consideration.
    The Commssion's related evaluation of the amendment is contained in 
a Safety Evaluation dated March 8, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 25, 1999, as 
supplemented on February 2 and 7, 2000.
    Brief description of amendment: The amended Technical 
Specifications permit use of the already-installed Oscillation Power 
Range Monitor system.
    Date of issuance: March 2, 2000.
    Effective date: As of the date of issuance to be implemented before 
activation of the Oscillation Power Range Monitor System, but no later 
than August 31, 2000.
    Amendment No.: 92.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 1, 1999 (64 FR 
67336).
    The February 2 and 7, 2000, letters provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: April 19, 1999, as supplemented 
August 25, October 14, November 3, December 20, 1999, and February 29, 
2000.
    Brief description of amendments: The amendment replaces the current 
Technical Specifications for fuel storage pool water lever, crane 
operability, and crane travel with a spent fuel cask with new Technical 
Specifications to reflect the permanently defueled status of the plant.
    Date of Issuance: March 7, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment No.: 107.
    Facility Operating License No. DPR-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999, (64 FR 
35208).
    The August 25, October 14, November 3, December 20, 1999, and 
February 29, 2000, letters provided clarifying information that did not 
change the scope of the original application and proposed no hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2000.
    No significant hazards consideration comments received: No.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-336 and 50-
423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London 
County, Connecticut

    Date of application for amendment: November 23, 1999.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) 4.0.5, ``Limiting Conditions for Operation and 
Surveillance Requirements'' by adding a biennial or 2-year surveillance 
interval and incorporating a required frequency for performing 
inservice testing activities of once per 731 days.
    Date of issuance: March 8, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 241 and 178.
    Facility Operating License Nos. DPR-65 and NPF-49: Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4286).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 8, 2000.

[[Page 15395]]

    No significant hazards consideration comments received: No.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: March 2, 1998, supplemented on 
January 21, 2000.
    Brief description of amendments: The amendments change the second 
paragraph of Technical Specification 3.8.D, ``Spent Fuel Pool Special 
Ventilation System,'' to clarify restrictions on movement of loads in 
the spent fuel pool enclosure with one train of spent fuel pool special 
ventilation system inoperable.
    Date of issuance: February 17, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 147 and 138.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 20, 1998 (63 FR 
27763).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 2000.
    No significant hazards consideration comments received: No.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie 
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, 
Minnesota

    Date of application for amendments: November 6, 1996, supplemented 
April 10 and October 1, 1997, and March 4, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification Section 5.0, ``DESIGN FEATURES,'' by relocating certain 
portions of the design features information to the Updated Safety 
Analysis Report, consistent with NUREG-1431, ``Standard Technical 
Specifications, Westinghouse Plants,'' Revision 1.
    Date of issuance: February 29, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 148 and 139.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4338).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 29, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 26, 1999.
    Brief description of amendment: This amendment raises the 
condensate storage tank (CST) low level setpoint and the corresponding 
allowable value in Technical Specification Tables 3.3.3-2 and 3.3.5-2. 
The subject setpoint is associated with the automatic transfer of the 
High Pressure Coolant Injection (HPCI) and Reactor Core Isolation 
Cooling (RCIC) pump suctions from the CST to the suppression pool in 
the event of low CST level. These changes are being made to address 
concerns regarding potential vortexing in the HPCI and RCIC suction 
flowpaths.
    Date of issuance: March 6, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 124.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 22, 1999 (64 
FR 51348).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 6, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 29, 1999, as supplemented 
November 30, 1999.
    Brief description of amendments: The amendments revise Technical 
Specifications Surveillance Requirement 4.6.1.1 to clarify when 
verification of primary containment integrity may be performed by 
administrative means and to change the surveillance interval for 
verification of manual valves and blind flanges inside of containment.
    Date of issuance: February 29, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 227 and 208.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 22, 1999 (64 
FR 51349).
    The November 30, 1999, letter provided clarifying information that 
did not change the staff's initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 29, 2000.
    No significant hazards consideration comments received: No.

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: April 11, 1996 (PCN 460), as 
supplemented April 6, 1998, and March 22 and July 29, 1999.
    Brief description of amendments: The amendments revise Technical 
Specification 3.6.3, ``Containment Isolation Valves,'' to specify that 
the completion time for required action for certain containment 
isolation valves be in accordance with the applicable limiting 
condition for operation pertaining to the engineered safety features 
system in which they are installed.
    Date of issuance: March 9, 2000.
    Effective date: March 9, 2000, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 2-165; Unit 3-156.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 19, 2000 (65 FR 
2993), as corrected January 26, 2000 (65 FR 4265).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 9, 2000.
    No significant hazards consideration comments received: No.

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: December 13, 1999, as 
supplemented February 24, 2000 (PCN-507).
    Brief description of amendments: The amendments revise the license 
expiration dates for San Onofre Unit 2 to February 16, 2022, and for 
San Onofre Unit 3 to November 15, 2022, thus extending the units' 
periods of operation to the full 40-year design-basis lifetime.
    Date of issuance: March 9, 2000.
    Effective date: March 9, 2000, to be implemented within 30 days of 
issuance.

[[Page 15396]]

    Amendment Nos.: Unit 2--166; Unit 3--157.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Operating Licenses.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73098).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 9, 2000.
    No significant hazards consideration comments received: No

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

    Date of amendment request: September 30, 1998, as supplemented May 
14 and October 21, 1999.
    Brief description of amendments: The amendments revise the South 
Texas Project, Units 1 and 2, offsite dose licensing bases to account 
for (1) operation of the existing steam generators at reduced feedwater 
inlet temperatures and (2) operation with the new replacement steam 
generators, also at a reduced feedwater temperature. The changes 
revised calculated offsite doses for four existing Updated Final Safety 
Analysis Report (UFSAR) Chapter 15 accidents and added a discussion in 
Chapter 15 of the radiological analysis for the voltage-based criteria 
for steam generator tubes.
    Date of issuance: March 2, 2000.
    Effective date: March 2, 2000, to be implemented within 30 days.
    Amendment Nos.: Unit 1--124; Unit 2--112
    Facility Operating License Nos. NPF-76 and NPF-80: Amendments 
authorize revisions to the UFSAR.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64124).
    The May 14 and October 21, 1999, supplemental letters provided 
clarifying information that was within the scope of the original 
Federal Register notice and did not change the staff's initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 2, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: August 30, 1999, as 
supplemented January 13, 2000.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) to delete the necessity for time response 
testing various instrument transmitters based on historical records 
indicating satisfactory time responses in the past.
    Date of issuance: February 29, 2000.
    Effective date: February 29, 2000.
    Amendment Nos.: Unit 1--251; Unit 2--242.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: October 6, 1999 (64 FR 
54381). The supplemental letter of January 13, 2000, did not expand the 
scope of the initial amendment request or change the NRC staff's 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 29, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: October 14, 1999 as 
supplemented February 23 and March 2, 2000.
    Brief description of amendments: Revise Section 4.4 of the 
Technical Specification (TS) surveillance testing requirements and 
their associated Bases to incorporate an alternate repair criteria for 
axial primary water stress corrosion cracking at dented tube support 
plate intersections.
    Date of issuance: March 8, 2000.
    Effective date: March 8, 2000.
    Amendment Nos.: Unit 1--252; Unit 2--243.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73100). The supplemental letters dated February 23, and March 2, 
2000, did not expand the scope of the original amendment request or 
change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 8, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 25, 1999, as supplemented 
December 17, 1999.
    Brief description of amendment: The amendment revises the main 
steam safety valve Technical Specification (TS) Section 3.7.1 to 
provide a new requirement to reduce the power range neutron flux-high 
reactor trip setpoints when two or more main steam safety valves 
(MSSVs) per steam generator are inoperable.
    Date of issuance: March 7, 2000.
    Effective date: March 7, 2000.
    Amendment No.: 19.
    Facility Operating License No. NPF-90: Amendment revises the TSs.
    Date of initial notice in Federal Register: August 11, 1999 (64 FR 
43781). The letter dated December 17, 1999 provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 7, 2000.
    No significant hazards consideration comments received: No

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

    Date of amendment request: February 11, 1999, as supplemented by 
letters dated September 3 and December 20, 1999.
    Brief description of amendments: The amendments change the 
Technical Specifications to authorize an increase in the allowable 
spent fuel storage capacity and the crediting of soluble boron, in the 
spent fuel pool, for spent fuel reactivity control.
    Date of issuance: February 24, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 74.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1999 (64 FR 
25522).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 24, 2000.
    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: January 20, 2000.
    Brief description of amendment: The amendment redefines the 
functional testing criteria for the noble gas activity

[[Page 15397]]

monitor instrumentation in the Augmented Off-Gas system.
    Date of Issuance: March 6, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 184.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 2000 (65 FR 
4999).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 6, 2000.
    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: February 11, 2000.
    Brief description of amendment: The amendment deletes the 
requirement to exercise the main steam isolation valves (MSIVs) twice 
weekly by partial closure and subsequent re-opening. Testing of the 
MSIVs to demonstrate their safety function will continue to be 
performed on a quarterly basis in accordance with the Vermont Yankee 
Inservice Testing program, Technical Specifications (TSs), and 
applicable provisions of Section XI of the ASME Boiler and Pressure 
Vessel Code. The TS change is issued as a follow-up amendment to NOED 
00-06-01, which was orally granted on February 10, 2000.
    Date of Issuance: March 9, 2000
    Effective date: As of the date of issuance, and shall be 
implemented prior to March 25, 2000.
    Amendment No.: 185
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
considerations: Yes (65 FR 8749) February 22, 2000. That notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by March 23, 2000, but indicated that if the Commission makes a 
final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 9, 2000.
    No significant hazards consideration comments received: No

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

    Date of application for amendments: May 6, 1999, as supplemented 
June 22 and December 16, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specifications Sections 3.3.1.1; 4.3.1.1.1; 4.3.1.1.2; 
4.3.1.1.3; 3.3.2.1; 4.3.2.1.1; 4.3.2.1.2; 4.3.2.1.3; 3/4.3.1; 3/4.3.2 
and 6.8.4.9 and Tables 3.3-1; 4.3-1; 3.3-3 and 4.3-2 for Unit 1, and 
Sections 3.3.1.1; 4.3.1.1.1; 4.3.1.1.2; 4.3.1.1.3; 3.3.2.1; 4.3.2.1.1; 
4.3.2.1.2; 4.3.2.1.3; 3/4.3.1; 3/4.3.2 and 6.8.4.9 and Tables 3.3-1; 
4.3-1; 3.3-3 and 4.3-2 for Unit 2, to revise the surveillance frequency 
for the Reactor Trip System (RTS) and the Engineered Safety Features 
Actuation System (ESFAS) analog instrumentation channels. In addition, 
the allowed outage time and action times for the RTS and ESFAS analog 
instrumentation and the actuation logic are being modified.
    Date of issuance: March 9, 2000
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 221 and 202.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: June 16, 1999 (64 FR 
32291). The letters of June 22 and December 16, 1999, contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 9, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: December 15, 1999.
    Brief description of amendment: The amendment modified the improved 
technical specifications (ITS) that were issued in Amendment No. 123 on 
March 31, 1999, and implemented on December 18, 1999. The changes 
expand the region of acceptable reactor coolant pump (RCP) seal 
injection flow to each RCP in Figure 3.5.5-1 and provides 10 editorial 
changes to the ITS.
    Date of issuance: March 1, 2000.
    Effective date: March 1, 2000, to be implemented within 60 days of 
the date of issuance.
    Amendment No.: 132.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4292).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 15th day of March 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-6913 Filed 3-21-00; 8:45 am]
BILLING CODE 7590-01-P