[Federal Register Volume 65, Number 134 (Wednesday, July 12, 2000)]
[Notices]
[Pages 43040-43058]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-17625]


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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 June 17, 2000, through June 30, 2000. The 
last biweekly notice was published on June 28, 2000 (65 FR 39956).

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.

[[Page 43041]]

However, should circumstances change during the notice period such that 
failure to act in a timely way would result, for example, in derating 
or shutdown of the facility, the Commission may issue the license 
amendment before the expiration of the 30-day notice period, provided 
that its final determination is that the amendment involves no 
significant hazards consideration. The final determination will 
consider all public and State comments received before action is taken. 
Should the Commission take this action, it will publish in the Federal 
Register a notice of issuance and provide for opportunity for a hearing 
after issuance. The Commission expects that the need to take this 
action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By August 11, 2000, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC, and electronically from 
the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. 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).

[[Page 43042]]

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

    Date of amendment request: June 21, 2000. This request supplements 
an earlier application dated October 29, 1999, submitted by GPU 
Nuclear, Inc., which has since been adopted by AmerGen Energy Company, 
LLC.
    Description of amendment request: The proposed amendment revises 
the Technical Specifications (TSs) to include: (1) The addition of 
operating limits for make-up tank (MUT) level and pressure; (2) the 
addition of surveillance requirements for the MUT pressure instrument 
channel; and (3) revision of the calibration frequency for the MUT 
level instrument channel from ``Not to exceed 24 months'' to 
``Refueling interval (once per 24 months)'' along with other 
instruments (high pressure and low pressure injection (LPI) flow 
instruments and the borated water storage tank (BWST) level instrument) 
in the same table as appropriate. Associated Bases changes are also 
proposed. Minor editorial changes (such as updates to the Table of 
Contents and others) are also proposed. This revision to the original 
submittal reflects changes to proposed TS Figure 3.3-1 and adds an 
additional instrument to those for which a surveillance calibration 
frequency extension is requested.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed changes do not represent a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The changes included in this LCA [license change application] 
impose new requirements for MU/HPI [make-up/high pressure injection] 
system operation and testing and extension of calibration 
frequencies for the MUT level, HPI flow and LPI flow instruments and 
BWST level instrument. These changes could not result in initiation 
of any accident previously evaluated. Therefore, the probability of 
an accident could not be affected by changes to the MU/HPI and Decay 
Heat Removal (DHR) systems.
    As described in the list of benefits for operation with MU/HPI 
cross-connect valves open, listed in section III.B above [section 
III.B, pages 5-6 of 14, of the June 21, 2000 supplement], the 
purpose of changing the operation of the MU/HPI system was to 
preclude the possibility of HPI pump damage. The addition of 
surveillance requirements for the MUT pressure instrument and the 
addition of LCO [limiting condition for operation] limits on MUT 
level and pressure along with appropriate action statements and 
required action times will ensure that gas entrainment of the MUT 
does not occur. The proposed change in instrument calibration 
frequencies will continue to maintain the required accuracy of the 
MUT level, HPI flow, LPI flow, and BWST level instruments.
    Minor editorial changes are included in this request to improve 
clarity and readability of the T.S. [technical specifications] and 
could not adversely affect plant operation.
    Therefore, the proposed changes will not adversely impact the 
reliability of the MU/HPI system and could not represent a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This LCA does not involve the addition of any new hardware. 
Along with minor editorial changes, the requested changes involve 
MU/HPI system operation and changes in instrument calibration 
frequency which have been reviewed in accordance with NRC guidance. 
Changes to MU/HPI System operation can only affect RCS [reactor 
coolant system] coolant inventory changes during operation and the 
ability to provide protection in the event of a Loss of Coolant 
Accident (LOCA). The full spectrum of LOCAs has been evaluated in 
the FSAR [Final Safety Analysis Report]. Therefore, no new accident 
scenarios have been created.
    The additional controls on MUT level and pressure provided by 
this LCA will ensure that a malfunction of a different type, gas 
entrainment of the MU/HPI pumps, will not occur. These limits on MUT 
level and pressure ensure that the initial conditions assumed for 
ECCS operation are maintained. The TS limits maintain the accident 
analysis initial conditions such that no operator action is required 
to avoid gas entrainment during ECCS [emergency core cooling 
[system] operation with the postulated single failure as required by 
the TMI-1 licensing basis (Reference 14) [GPU Nuclear Safety 
Evaluation No. SE-000211-015, Revision 0, ``Operation with MU X-
Connect Valves OPEN''].
    Extension of the calibration frequencies for the HPI level, HPI 
flow, LPI flow, and BWST level will continue to maintain the 
accuracy of these instruments and could not create the potential for 
any new accident that has not been evaluated.
    Minor editorial changes are included in this request to improve 
the clarity and readability of the TS and could not adversely affect 
plant operation.
    Therefore, these [proposed] changes do not create the potential 
for any accident different from those that have been evaluated.
    C. These proposed changes do not involve a significant reduction 
in a margin of safety.
    This LCA includes changes to MU/HPI system operation and testing 
and an extension of the calibration frequency for certain 
instruments. The requested changes will serve to maintain the proper 
system initial conditions to ensure the ability of the MU/HPI system 
to provide protection in the event of a Loss of Coolant Accident 
(LOCA) and maintain the required instrument accuracy for the 
instruments where changes to a refueling interval frequency are 
being requested. NRC guidance for addressing the effect on increased 
surveillance intervals on instrument drift and safety analysis 
assumptions presented in GL [generic letter] 91-04 have been 
addressed in enclosure 1A [of the licensee's June 21, 2000 letter].
    Minor editorial changes are included in this request to improve 
clarity and readability of the TS and could not adversely affect 
plant operation.
    These changes, which are consistent with the TMI-1 licensing and 
design basis requirements, do not result in a degradation of safety 
related equipment, and therefore, do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis (paragraph `B') 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    The licensee concluded that ``these [proposed] changes do not 
create the potential for any accident different from those that have 
been evaluated.'' This conclusion is worded slightly differently than 
the standard in 10 CFR 50.92 (``The proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated'') that the licensee is required to analyze 
against pursuant to 10 CFR 50.91. Nevertheless, the licensee did state 
in its application that ``additional controls on MUT level and pressure 
provided by the proposed changes in this LCA will ensure that a 
malfunction of a different type, gas entrainment of the MU/HPI pumps, 
will not occur.'' These additional controls include a prohibited 
operating region which would require plant shutdown if not corrected.
    The licensee further stated, that the portion of the TS Figure 3.3-
1 related to NPSH [net positive suction head] has been deleted because 
operation of MU/HPI pump below the manufacturer's NPSH limits for a 
short period of time may affect pump performance while the NPSH 
shortfall exists, but would not render the pump inoperable. The 
licensee further stated that existing plant procedures will provide 
NPSH MUT pressure verses level operating limits that will ensure the 
recommended NPSH would be available for the NPSH limiting event, an HPI 
line break small-break LOCA. Based on the above, the staff has 
determined that the proposed changes and additional controls on MUT 
level and pressure would not create the possibility of a new or 
different kind of accident from any previously evaluated.

[[Page 43043]]

    The licensee has determined that the proposed extension of the 
calibration frequencies for the HPI level, HPI flow, LPI flow, and BWST 
level, meets applicable staff guidance related to these proposed 
changes and will continue to maintain the accuracy of these instruments 
and could not, therefore, create the potential for any new accident 
that has not been evaluated. The staff has determined that the proposed 
extension of calibration frequencies would not create the possibility 
of a new or different kind of accident from any previously evaluated.
    The proposed editorial changes are minor in nature, and are 
intended to improve the clarity and readability of the TSs, and would 
not create the possibility of a new or different kind of accident from 
any previously evaluated.
    Based on this review, and the licensee's basis for its 
determination with respect to items ``A'' and ``C'' above, 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: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Marsha Gamberoni.

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

    Date of amendments request: June 6, 2000.
    Description of amendments request: The proposed amendments would 
revise information in Figure 3.5.5-1, ``Minimum Required RWT Volume,'' 
in Technical Specification (TS) 3.5.5, ``Refueling Water Tank (RWT),'' 
of the TSs for the three units. The amendments are administrative 
changes to the figure that would (1) Relocate design bases information 
to the Bases of the TSs, (2) truncate the lower end of the RWT limit 
curve at 210  deg.F, (3) re-title the right-hand ordinate from 
``minimum useful volume required in the RWT'' to ``RWT Volume,'' and 
(4) delete the two footnotes and the references to the footnotes.
    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 in its application, which is presented below:

    Standard 1: Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    This proposed administrative change does not involve any changes 
to the design, operation, or maintenance of any structures[,] 
systems or components. The requirements in TS 3.5.5 for RWT 
operability will not be changed. This proposed amendment [for each 
unit] does not alter, degrade, or prevent actions described or 
assumed in an accident described in the PVNGS UFSAR [Updated Final 
Safety Analysis Report] from being performed. It will not alter any 
assumptions previously made in evaluating radiological consequences 
or, affect any fission product barriers. It does not increase any 
challenges to safety systems as well. Any changes to the information 
relocated to the TS Bases would be controlled under the TS Bases 
Control program, TS 5.5.14, which utilizes the criteria of 10 CFR 
50.59 to determine if prior NRC [Nuclear Regulatory Commission] 
approval is required for any changes. Therefore, this proposed 
amendment [for each unit] would not significantly increase the 
consequences of an accident previously evaluated.
    Standard 2: Does the proposed change create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    This proposed administrative change does not involve any changes 
to the design, operation, or maintenance of any structures[,] 
systems or components. The requirements in TS 3.5.5 for RWT 
operability will not be changed. This proposed amendment [for each 
unit] does not alter, degrade, or prevent actions described or 
assumed in an accident described in the PVNGS UFSAR from being 
performed. Any changes to the information relocated to the TS Bases 
would be controlled under the TS Bases Control program, TS 5.5.14, 
which utilizes the criteria of 10 CFR 50.59 to determine if prior 
NRC approval is required for any changes.
    Therefore, the proposed amendment [for each unit] does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    Standard 3: Does the proposed change involve a significant 
reduction in a margin of safety?
    This proposed administrative change does not involve any changes 
to the design, operation, or maintenance of any structures[,] 
systems or components. The requirements in TS 3.5.5 for RWT 
operability will not be changed. This proposed amendment [for each 
unit] does not alter, degrade, or prevent actions described or 
assumed in an accident. Any changes to the information relocated to 
the TS Bases would be controlled under the TS Bases Control program, 
TS 5.5.14, which utilizes the criteria of 10 CFR 50.59 to determine 
if prior NRC approval is required for any changes. Therefore, the 
proposed change does not involve a significant reduction in any 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendments request: June 6, 2000.
    Description of amendments request: The proposed amendments would 
restrict the emergency diesel generator (DG) acceptance criteria for 
steady-state voltage and frequency in several surveillance requirements 
(SRs) involving DG starts in Technical Specification (TS) 3.8.1, ``AC 
Sources--Operating,'' of the TSs for the three units. The amendments 
would also add a note to each SR that states: ``The steady state 
voltage and frequency limits are analyzed values and have not been 
adjusted for instrument error.'' The restricted acceptance criteria is 
to ensure proper DG operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration in its application, which is presented below:

    Standard 1: Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    No. The proposed change does not significantly increase the 
probability of an accident previously evaluated in the Updated Final 
Safety Analysis Report (UFSAR). The more restrictive steady-state 
voltage and frequency ranges ensure that the equipment being powered 
by the diesel generator will function as required to mitigate an 
accident as described in the UFSAR. The diesel generators are part 
of the systems required to mitigate an accident. Mitigation 
equipment is not a factor in accident initiation and, therefore, the 
probability of an accident previously evaluated will not be 
significantly increased.
    The change to the steady state diesel generator voltage and 
frequency acceptance limits does not increase the probability of a 
diesel generator failure [or a failure of offsite power]. Therefore, 
this change does not increase the probability of a station blackout 
event.
    The consequences of an accident previously evaluated in the 
UFSAR will not be significantly increased. The more restrictive 
change to the diesel generator

[[Page 43044]]

steady-state voltage and frequency acceptance limits ensures that 
the equipment powered by the diesel generators will perform as 
analyzed and mitigate the consequences of any accident described in 
the UFSAR. Therefore, the change in steady-state voltage and 
frequency acceptance limits is within the bounds of previously 
analysis in the UFSAR and does not involve a significant increase in 
the consequences of an accidently previously evaluated.
    Standard 2: Does the proposed change create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    No. The possibility of an accident of a new or different kind 
from any accident previously evaluated has not been created. The 
more restrictive change to the diesel generator steady-state voltage 
and frequency acceptance limits ensures that the equipment powered 
by the diesel generators will perform as analyzed. This equipment 
and the diesel generators mitigate the consequences of an accident. 
Mitigation equipment does not contribute to accident initiation. 
Making existing requirements more restrictive will not alter the 
plant configuration (no new or different type of equipment will be 
installed) or change the methods governing normal plant operation. 
These changes are consistent with the assumptions made in the safety 
analysis. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Standard 3: Does the proposed change involve a significant 
reduction in a margin of safety?
    No. The change to the diesel generator steady-state voltage and 
frequency acceptance limits ensures that the equipment powered by 
the diesel generators will perform as analyzed. This equipment and 
the diesel generators mitigate the consequences of an accident. This 
change maintains the required function of the equipment powered by 
the diesel generators and ensures the required operation of the 
plant and any structures[,] systems, or components is as intended by 
the safety analysis. Therefore, the proposed change does not involve 
a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: June 7, 2000.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) related to the Engineered 
Safety Features Actuation System (ESFAS) Instrumentation found in TS 3/
4.3.1, TS 3/4.3.2, and the associated Bases. Specifically, the proposed 
change would revise surveillance test intervals and allowed outage 
times for ESFAS instrumentation in TS 3/4.3.2. The proposed revision is 
based on WCAP-10271, ``Evaluation of Surveillance Frequencies and Out 
of Service Times for the Reactor Protection Instrumentation System,'' 
its supplements, and the NRC approvals issued in the Safety Evaluation 
Reports (SERs) dated February 21, 1985, and February 22, 1989, and the 
Supplemental SER dated April 30, 1990. In addition, the licensee is 
proposing specific changes to the reactor trip system instrumentation 
in TS 3/4.3.1, which are directly associated with implementing the 
ESFAS relaxations proposed in the submittal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The determination that the results of the proposed changes are 
within all acceptable criteria was established in the SERs prepared 
for WCAP-10271 Supplement 2 and WCAP-10271 Supplement 2, Revision 1 
issued by letters dated February 22, 1989 and April 30, 1990. 
Implementation of the proposed changes is expected to result in an 
acceptable increase in total Engineered Safety Features Actuation 
System yearly unavailability. This increase, which is primarily due 
to less frequent surveillance, results in a small increase in core 
damage frequency (CDF) and public health risk. The values determined 
by the WOG [Westinghouse Owners Group] and presented in the WCAP for 
the increase in CDF were verified by Brookhaven National Laboratory 
(BNL) as part of an audit and sensitivity analyses for the NRC 
staff. Based on the small value of the increase compared to the 
range of uncertainty in the CDF, the increase is considered to be 
acceptable.
    Removal of the requirement to perform the Reactor Trip System 
analog channel operational test on a staggered basis will have a 
negligible impact on the Reactor Trip System unavailability. 
Staggered Testing was initially imposed to address the concerns of 
common cause failures. HNP's [Harris Nuclear Plant's] program to 
evaluate failures for common cause, process parameter signal 
diversity, and normal operational test spacing yield most of the 
benefits of staggered testing.
    The proposed changes do not result in an increase in the 
severity or consequences of an accident previously evaluated. 
Implementation of the proposed changes may affect the probability of 
failure of the RPS [reactor protection system], but does not alter 
the manner in which protection is afforded nor the manner in which 
limiting criteria are established.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve hardware changes and do not 
result in a change in the manner in which the Reactor Protection 
System provides plant protection or the manner in which 
surveillances are performed to demonstrate operability. No change is 
being made which alters the functioning of the Reactor Protection 
System. Rather the likelihood or probability of the Reactor 
Protection System functioning properly is affected as described 
above.
    Therefore the proposed changes do not create the possibility of 
a new or different kind of accident.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system setpoints or limiting conditions for 
operation are determined. The impact of reduced testing, other than 
as addressed above, is to allow a longer time interval over which 
instrument uncertainties (e.g., drift) may act. An evaluation has 
been performed to assure that the plant setpoints properly account 
for these instrument uncertainties over the larger time interval.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety as follows:
    a. Less frequent testing will result in fewer inadvertent 
reactor trips and inadvertent actuations of Engineered Safety 
Features Actuation System components.
    b. Less frequent distraction of the operator and shift 
supervisor to attend to and support instrumentation testing will 
improve the effectiveness of the operating staff in monitoring and 
controlling plant operation.
    The foregoing analysis demonstrates that the proposed amendment 
to HNP TS does not involve a significant increase in the probability 
or consequences of a previously evaluated accident, does not create 
the possibility of a new or different kind of accident, and does not 
involve a significant reduction in a margin of safety.
    Based upon the preceding analysis, CP&L [Carolina Power & Light 
Company] concludes that the proposed amendment does not involve a 
significant hazards consideration.

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

[[Page 43045]]

Secretary, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: May 11, 2000.
    Description of amendment request: The proposed changes would revise 
Technical Specification Surveillance Requirement (SR) 3.6.1.3.8. SR 
3.6.1.3.8 currently requires verification of the actuation capability 
of each excess flow check valve (EFCV) every 24 months. This proposed 
change would relax the SR frequency by allowing a ``representative 
sample'' of reactor instrument line EFCVs to be tested every 24 months, 
such that each reactor instrument line EFCV will be tested at least 
once every 10 years.
     Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided an analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The current SR frequency requires each reactor instrument line 
EFCV to be tested every 24 months. The reactor instrument line EFCVs 
at WNP-2 are designed so that they will not close accidentally 
during normal operation, but will close if a rupture of the 
instrument line is indicated downstream of the valve, and have their 
status indicated in the control room. This proposed change allows a 
reduced number of reactor instrument line EFCVs to be tested every 
24 months. There are no physical plant modifications associated with 
this change. Industry operating experience demonstrates a high 
reliability of these valves. Neither reactor instrument line EFCVs 
nor their failures are capable of initiating previously evaluated 
accidents; therefore; there can be no increase in the probability of 
occurrence of an accident regarding this proposed change.
    Reactor instrument lines connecting to the reactor coolant 
pressure boundary are equipped with EFCVs and also have a flow-
restricting orifice inside containment and upstream of the EFCV. The 
consequences of an unisolable rupture of such an instrument line has 
been previously evaluated in WNP-2 FSAR 15.6.2. The instrument lines 
that penetrate primary containment conform to Regulatory Guide 1.11 
(WNP-2 FSAR 7.1.2.4). Those instrument lines are Seismic Category I 
and terminate in instruments that are Seismic Category I (reference 
WNP-2 FSAR Table 6.2-16 note 27).
    The sequence of events in WNP-2 FSAR Section 15.6.2.2 for a 
reactor instrument line break assumes a continuous discharge of 
reactor water through the instrument line until the reactor vessel 
is cooled and depressurized (5 hours). Although not expected to 
occur as a result of this change, the postulated failure of an EFCV 
to isolate as a result of reduced testing is bounded by this 
previous evaluation. Therefore, there is no increase in the 
previously evaluated consequences of the rupture of an instrument 
line and there is no potential increase in the consequences of an 
accident previously evaluated as a result of this change.
    The containment atmosphere and suppression pool instrument line 
EFCVs are required to remain open to sense containment atmosphere 
and suppression pool level conditions during postulated accidents. 
They are not required to close during an instrument line break 
assumed during normal plant operation nor is their design capable of 
closing during normal plant conditions. These EFCVs do not meet the 
criteria for inclusion in 10 CFR 50.36(c)(3) as they have no active 
safety function and thus relocation of their testing requirements to 
processes controlled under 10 CFR 50.59 cannot affect the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change allows a reduced number of reactor 
instrument line EFCVs to be tested each operating cycle and that the 
testing requirements for containment atmosphere and suppression pool 
instrument line EFCVs be relocated to a process controlled under 10 
CFR 50.59. No other changes in requirements are being proposed. 
Industry operating experience demonstrates the high reliability of 
these valves. The potential failure of a reactor instrument line 
EFCV to isolate by the proposed reduction in test frequency is 
bounded by the previous evaluation of an instrument line rupture. 
This change will not physically alter the plant (no new or different 
type of equipment will be installed). This change will not alter the 
operation of process variables, structures, or components as 
described in the safety analysis. Thus, a new or different kind of 
accident will not be created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The consequences of an unisolable rupture of an instrument line 
has been evaluated in WNP-2 FSAR Section 15.6.2 in accordance with 
the requirements of Regulatory Guide 1.11. That evaluation assumed a 
continuous discharge of reactor water for the duration of the 
detection and cooldown sequence (5 hours). The only margin of safety 
applicable to this proposed change is considered to be that implied 
by this evaluation. Since a continuous discharge was assumed in this 
evaluation, any potential failure of a reactor instrument line EFCV 
to isolate as a result of reduced testing frequency is bounded by 
existing analysis and does not involve a significant reduction in 
the margin of safety.
    There is no accident for which the containment atmosphere or 
suppression pool instrument line EFCVs are designed to actuate to 
the isolation position for mitigation. A postulated break of a 
containment atmosphere or suppression pool instrument line under 
normal operating conditions would not result in a condition that 
would create the ability for these EFCVs to operate because neither 
the containment pressure nor the suppression pool level head would 
be sufficient to result in their actuation. As these EFCVs have no 
active design or safety function, the relocation of testing 
requirements would not involve a significant reduction in the margin 
of safety. A postulated break of any instrument line simultaneous 
with a loss of coolant accident is beyond the design basis for the 
plant.
    Based upon the above, the proposed amendment is judged to 
involve 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: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: May 25, 2000.
    Description of amendment request: The proposed amendment would 
change the action statements for Technical Specification (TS) 3.8.2.2, 
A.C. Distribution--Shutdown, and TS 3.8.2.4, D.C. Distribution--
Shutdown, by replacing the requirement to establish containment 
integrity within 8 hours, with a requirement to immediately suspend 
core alterations, the movement of irradiated fuel assemblies, and any 
operations involving positive reactivity additions. Related changes to 
the associated Bases were also proposed.
    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 existing requirement to establish containment integrity upon 
a loss of a required AC or DC bus in Mode 5 or 6 is not relied upon 
in any ANO-2 [Arkansas Nuclear One, Unit 2] accident analysis. Other 
components that may be rendered inoperable upon the loss of a 
required AC or DC bus are governed by other TSs and associated 
action

[[Page 43046]]

statements. Such functions include core cooling, reactor coolant 
makeup capabilities, the status of containment penetrations and 
openings, and reactor coolant inventory. The TSs that govern these 
functions provide appropriate actions to address the failure at 
hand. The proposed change[s] act to minimize the possibility of a 
fuel handling accident when a required AC or DC bus is inoperable by 
requiring the suspension of the handling of irradiated fuel and core 
alterations. In addition, ANO-2 has demonstrated that the offsite 
dose consequences of a fuel handling accident within the containment 
building remain well within 10 CFR 100 limits without taking credit 
for the containment's fission product control function. Deleting the 
requirement to establish containment integrity is not relevant to 
the initiation of any accident previously evaluated, nor does it 
significantly increase the consequences of any accident previously 
evaluated. Other TS LCOs [limiting conditions for operation] provide 
appropriate actions that address shutdown cooling (SDC), makeup 
capability and inventory, and other important functions. The 
proposed change deletes the requirement to establish containment 
integrity in favor of those actions that act to minimize the 
likelihood of a fuel handling accident or a positive reactivity 
excursion. The proposed change reduces unnecessary actions required 
upon the loss of an AC or DC bus and provide greater consistency 
with the philosophies of the RSTS [Revised Standard Technical 
Specifications].
    Therefore, the proposed change does 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 existing actions associated with shutdown mode AC and DC TS 
sources are not considered accident initiators. The proposed 
revision does not present a physical change to plant systems or 
equipment. Deleting the requirement to establish containment 
integrity in favor of actions that aid in minimizing the likelihood 
of a fuel handling accident or positive reactivity excursion does 
not result in any new or different kind of accident from any 
previously evaluated.
    Therefore, the proposed change does 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 existing requirement to establish containment integrity upon 
a loss of any required AC or DC bus in Modes 5 or 6 acts to limit 
offsite release consequences should an accident occur during the 
period of inoperability. The proposed change acts to address the 
source, that is, aids in minimizing the likelihood of a fuel 
handling accident or an undetected positive reactivity addition 
while in Modes 5 and 6. By suspending all handling of irradiated 
fuel and core alterations, the likelihood of a fuel handling 
accident occurring is minimized. Since the loss of a required AC or 
DC bus could impact plant instrumentation, the suspension of all 
activities involving positive reactivity additions aids in 
preventing the impact of a positive reactivity addition from being 
undetected. Other possible Mode 5 and 6 conditions (loss of 
inventory, loss of shutdown cooling, etc.) are addressed in other 
shutdown mode TSs. In addition, ANO-2 has demonstrated that the 
offsite dose consequences of a fuel handling accident within the 
containment building remain well within 10 CFR 100 limits without 
taking credit for the containment's fission product control 
function. Since the proposed change exchanges accident mitigation 
strategy in favor of accident prevention strategy, no significant 
reduction in the margin to safety is evident.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: May 1, 2000.
    Description of amendment request: The proposed amendment would: (1) 
Revise Technical Specification (TS) requirements regarding the minimum 
number of radiation monitoring instrumentation channels required to be 
operable during movement of fuel within the containment; (2) revise the 
Modes in which the surveillance specified by Table 4.3-3, ``Radiation 
Monitoring Instrumentation Surveillance Requirements,'' Item 2.c.ii is 
required; (3) revise TS 3.9.4, ``Containment Building Penetrations,'' 
to allow both Personnel Air Lock (PAL) doors and other containment 
penetrations to be open during movement of fuel assemblies within 
containment, provided certain conditions are met; (4) revise 
applicability and action statement requirements of TS 3.9.4. to be for 
only during movement of fuel assemblies within containment; (5) revise 
periodicity and applicability of Surveillance Requirement (SR) 4.9.4.1; 
(6) revise SR 4.9.4.2 to verify flow rate of air to the supplemental 
leak collection and release system (SLCRS) rather than verifying the 
flow rate through the system; (7) add two new SRs, 4.9.4.3 and 4.9.4.4, 
for verification and demonstration of SLCRS operability; (8) modify TS 
3/4.9.9 for the containment purge exhaust and isolation system to be 
applicable only during movement of fuel assemblies within containment; 
and, (9) revise associate TS Bases as well as make editorial and format 
changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment involves changes to accident mitigation 
system requirements. These systems are related to controlling the 
release of radioactivity to the environment and are not considered 
to be accident initiators to any previously analyzed accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability of an accident previously evaluated.
    Based on the current technical specification requirements, an 
environmental release due to a fuel handling accident (FHA) 
occurring within containment is precluded by a design which 
automatically isolates the containment following detection of 
radioactivity by redundant containment purge monitors. The proposed 
amendment, which permits containment penetrations to be open during 
movement of fuel assemblies within containment, increases the dose 
at the site boundary and the control room operator dose due to a FHA 
occurring within containment; however, the dose remains within 
acceptable limits. Based on a radiological analysis of a FHA within 
containment with open containment penetrations being filtered by the 
Supplemental Leak Collection and Release System (SLCRS), the 
resultant radiological consequences of this event are well within 
the 10 CFR Part 100.11 limits, as defined by acceptance criteria in 
the Standard Review Plan (SRP) Section 15.7.4. Control room operator 
doses remain less than the 10 CFR Part 50 Appendix A General Design 
Criteria (GDC) 19 limit of 5 rem whole body or its equivalent to any 
part of the body. The proposed changes to LCO 3.9.4 and associated 
surveillance requirements will ensure that SLCRS filtration 
assumptions in the associated radiological analysis are met.
    LCO 3.9.10 titled ``Water Leve--Reactor Vessel'' will continue 
to ensure that at least 23 feet of water is maintained over the fuel 
during fuel movement when the plant is in Mode 6. LCO 3.9.3 titled 
``Decay Time'' will continue to ensure that irradiated fuel is not 
moved in the reactor pressure vessel until at least 150 hours after 
shutdown. These LCOs will continue to ensure that two of the key

[[Page 43047]]

assumptions used in the radiological safety analysis are met.
    The radiological consequences of the Core Alteration events 
other than the FHA remain unchanged. These events do not result in 
fuel cladding integrity damage. A radioactive release to the 
environment is not postulated since the activity is contained in the 
fuel rods. Therefore, the affected containment systems are not 
required to mitigate a radioactive release to the environment due to 
a Core Alteration event.
    The proposed revision in the minimum number of the Containment 
Purge Exhaust Radiation Monitoring Instrumentation channels required 
to be operable from one to two, ensures that redundant instrument 
channels are available to detect and initiate isolation of the 
containment purge and exhaust containment penetrations during a FHA 
inside containment.
    The proposed administrative, editorial, and format changes do 
not affect plant safety. The Bases section has been revised as 
necessary to reflect the changes to these Specifications. Bases 
Section 3/4.9.9 will also be revised to remove text pertaining to 
Mode 5 applicability that is not relevant to this specification.
    Therefore, the proposed amendment does not significantly 
increase the consequences of any previously evaluated accident.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment affects a previously evaluated accident; 
e.g., FHA. The proposed amendment does not represent a significant 
change in the configuration or operation of the plant. The proposed 
amendment does not impact Technical Specification requirements for 
systems needed to prevent or mitigate other Core Alteration events. 
The filtered SLCRS that will be utilized to control and filter the 
radioactive release from a FHA occurring within containment is the 
same system (with the exception of the flow path to the filter 
banks) currently relied upon to control and filter the release from 
a FHA in the fuel building. The primary function of SLCRS is to 
ensure that radioactive leakage from the primary containment 
following a Design Basis Accident (DBA) or radioactive release due 
to a fuel building FHA is collected and filtered for iodine removal 
prior to discharge to the atmosphere at an elevated release point 
through a ventilation vent. This system will be relied upon to 
control the releases from open containment penetrations should a FHA 
occur inside of containment until such time that these open 
containment penetrations can be isolated. The proposed amendment 
contains the requirement to maintain the capability to close open 
containment penetrations within 30 minutes following a FHA inside 
containment.
    The filtered SLCRS that will be relied upon to mitigate a FHA 
within containment is classified as Quality Assurance (QA) Category 
I, Safety Class 3 and Seismic Category I as stated in Updated Final 
Safety Analysis Report (UFSAR) Section 6.5.3.2.1 titled ``Design 
Bases.'' As described in UFSAR Section 6.5.1 titled ``Engineered 
Safety Feature Filter Systems,'' filtered SLCRS is considered to be 
an engineered safety features (ESF) filter system used to mitigate 
the consequences of accidents.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Based on the current technical specification requirements, an 
environmental release due to a FHA occurring within containment is 
precluded by a design which automatically isolates the containment 
following detection of radioactivity by redundant containment purge 
monitors. The proposed amendment increases the dose at the site 
boundary and the control room operator dose due to a FHA occurring 
within containment; however, the dose remains within acceptable 
limits. The margin of safety as defined by 10 CFR Part 100 has not 
been significantly reduced.
    The revised radiological analysis based on the proposed 
amendment demonstrates that during a FHA inside containment, the 
projected offsite doses will be well within the applicable 
regulatory limits of 10 CFR Part 100.11 of 300 rem thyroid and 25 
rem whole body, and are less than the more restrictive guidance 
criteria in the SRP Section 15.7.4 of 75 rem thyroid and 6 rem whole 
body. Control room operator doses are less than the 10 CFR Part 50 
Appendix A GDC 19 limit of 5 rem whole body or its equivalent to any 
part of the body. This radiological analysis is based on all 
airborne activity reaching the containment atmosphere, as a result 
of a FHA inside containment, being released to the environment over 
a 2 hour period. The 2 hour release period is based on the guidance 
contained in Regulatory Guide 1.25 titled ``Assumptions Used for 
Evaluating the Potential Radiological Consequences of a Fuel 
Handling Accident in the Fuel Handling and Storage Facility for 
Boiling and Pressurized Water Reactors.'' The proposed amendment 
contains a Bases requirement to maintain the capability to close 
open containment penetrations within 30 minutes following a FHA 
inside containment. Completion of this action will reduce the dose 
consequence of a FHA within containment by terminating the release 
to the environment prior to all airborne activity being released 
from the containment.
    The margin of safety for Core Alteration events other than the 
FHA is not significantly reduced due to this proposed amendment. The 
proposed amendment does not impact Technical Specification 
requirements for systems needed to prevent or mitigate such Core 
Alteration events. These events do not result in fuel cladding 
integrity damage. Therefore, a radioactive release to the 
environment is not postulated since the activity is contained in the 
fuel rods.
    The proposed revision in the minimum number of the Containment 
Purge Exhaust Radiation Monitoring Instrumentation channels required 
to be operable from one to two, ensures that redundant instrument 
channels are available to detect and initiate isolation of the 
containment purge and exhaust containment penetrations during a FHA 
occurring inside containment.
    The proposed changes to SR 4.9.4.1 and SR 4.9.9, to remove 
unnecessary detail on when these surveillances are required to be 
performed, are administrative in nature and do not affect plant 
safety.
    The proposed revision of the words ``through the'' to the words 
``to filtered'' in SR 4.9.4.2.a does not change the LCO 3.9.4 
requirements. This change makes the LCO and surveillance 
requirements consistent. This change is administrative in nature and 
does not affect plant safety.
    The proposed administrative, editorial, and format changes do 
not affect plant safety. The Bases section has been revised as 
necessary to reflect the changes to these Specifications. Bases 
Section 3/4.9.9 will also be revised to remove text pertaining to 
Mode 5 applicability that is not relevant to this specification.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: May 31, 2000.
    Description of amendment request: The proposed amendment would 
revise the Crystal River Unit 3 Improved Technical Specifications (ITS) 
to add an additional Condition and Required Action to ITS 3.3.11, 
``Emergency Feedwater Initiation and Control (EFIC) System 
Instrumentation.'' The Action would require tripping the affected 
reactor coolant pump (RCP) status signals to each of the four EFIC 
channels when one or more RCP status signals or Reactor Coolant Pump 
Power Monitors (RCPPMs) for up to two RCPs become inoperable. This 
action is intended to ensure continued operability of the EFIC RCP 
status function when one or more RCPPMs or their associated RCP status 
signals are inoperable. The amendment also proposes changes to ITS 
Table 3.3.11-1 to properly characterize the configuration of the 
signals from the RCPPMs to EFIC, and to clarify source of the Loss of 
Main Feedwater Pump signals to EFIC. The proposed changes to Table 
3.3.11-1 are intended to provide consistency between Table

[[Page 43048]]

3.3.11-1 and information provided in the ITS Bases for the EFIC System 
Instrumentation Specification.
    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 not involve a significant increase in the probability or 
consequences of an accident previously analyzed. 
    The EFIC system is not an initiator of any design basis 
accident. The EFIC RCP status signal function is intended to ensure 
emergency feedwater is available to automatically raise levels in 
the once through steam generator (OTSG) to the natural circulation 
setpoint in the event of a loss of reactor coolant system (RCS) 
forced flow.
    The proposed license amendment adds clarifying information to 
ITS Table 3.3.11-1, and an additional Required Action to ITS 3.3.11 
that assures continued operability of the RCP status function of the 
EFIC system in the event one or more RCPPMs or their associated RCP 
status signals become inoperable. The design functions of the EFIC 
system and the initial conditions for accidents that require EFIC 
will not be affected by the change. Therefore, the change will not 
increase the probability or consequences of an accident previously 
evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed. 
    The proposed amendment involves no changes to the design or 
operation of the EFIC system. The RCPPMs are part of the design of 
the Emergency Feedwater Initiation and Control (EFIC) System, and 
are assumed to function properly in the accident analysis. The 
proposed amendment will assure that the EFIC system performs as 
assumed in the safety analysis in the event of a loss of RCS forced 
flow. The proposed amendment change will not affect the other EFIC 
functions, and will not create any new plant configurations. 
Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does not involve a significant reduction in the margin of 
safety.
    The proposed amendment adds additional actions to be taken in 
the event one or more RCPPMs or their associated RCP status signals 
become inoperable, and provides clarifying information regarding the 
sources and configuration of signals to EFIC. The proposed amendment 
ensures appropriate actions are taken to restore the operability of 
the EFIC RCP status function in the event that one or more RCP 
status signals to EFIC are lost. Thus, the proposed amendment will 
not result in a reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Section Chief: Richard P. Correia.

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

    Date of amendment request: June 1, 2000.
    Description of amendment request: The proposed amendment would 
revise the Crystal River Unit 3 (CR-3) Improved Technical 
Specifications 3.4.14 to extend the interval for calibration of the 
containment sump monitor from the current 18 months to 24 months. The 
monitor is used to detect and measure reactor coolant system (RCS) 
leakage by monitoring changes in the level of water in the containment 
sump. Extending the interval to 24 months would make it consistent with 
the current CR-3 operating cycle.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. 
    The containment sump monitor is not an initiator of any design 
basis accident. This monitor is used during normal plant operation 
to measure and to trend the rate of change of containment sump fluid 
level.
    The containment sump monitor does not perform any safety 
function as part of mitigating the consequences of a design basis 
accident. Separate safety-related instrumentation is used to 
determine post-accident containment sump and containment flood 
levels and to satisfy the requirements of Regulatory Guide (RG) 1.97 
for post-accident monitoring instrumentation. Additionally, the 
containment sump monitor does not have any associated safety system 
setpoint. The level switch in the instrument circuit is used only 
for automatic pumping of sump fluid using the two containment sump 
pumps.
    A longer interval between calibrations may result in some 
increase in the amount of drift that the containment sump level 
monitor might experience between calibrations. The behavior of 
instrumentation, including considerations such as the amount of 
drift that the instrument might experience between calibrations, is 
not an accident precursor. Thus, changes to instrument maintenance 
such as intervals for performance of calibration, and the behavior 
of instruments including such considerations as the amount of drift, 
do not affect the probability of an accident. The probability of an 
accident previously evaluated is independent of the amount of drift 
that the containment sump level monitor might experience.
    The containment sump monitor is used to detect RCS leakage 
during normal operation and does not have an accident mitigation 
function. Additionally, the ability of the instrument to detect 
small leaks will not be affected by extending the calibration 
interval.
    Based on the above, increasing the interval between calibrations 
does not result in a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    The proposed license amendment involves no changes to the design 
or operation of the containment sump level monitor. Extending the 
interval between calibrations of the containment sump level monitor 
from 18 months to 24 months might result in greater drift of the 
monitor during the period of operation. However, the only function 
of the monitor is to detect changes and trends in the containment 
sump level during normal operation and the amount of drift that the 
monitor has experienced does not affect its ability to measure such 
changes and trends of the containment sump level. Furthermore, 
changes in the behavior of instrumentation, such as the amount of 
drift that the instrument might experience between calibrations, do 
not create the possibility of a new or different kind of accident.
    Because initiation of accidents is independent of 
instrumentation behavior parameters such as drift, extending the 
calibration interval from 18 to 24 months does not create the 
possibility of any new or different kind of accident from any 
previously evaluated.
    3. Involve a significant reduction in a margin of safety?
    The CR-3 operating license, i.e., the Improved Technical 
Specifications, requires that instrumentation to detect leakage of 
reactor coolant system (RCS) inventory be available and operable 
during power operation. The required instrumentation is one 
containment sump monitor and one containment atmosphere 
radioactivity monitor.
    The proposed extension of the containment sump monitor 
calibration interval from 18 to 24 months does not compromise the 
ability of the instrumentation to perform its safety function, i.e., 
early detection of RCS leakage. This is so because the only function 
of the containment sump monitor is to detect changes and trends in 
the containment sump level during normal operation. The proposed 
license amendment makes no changes to either the design or operation 
of the sump monitor. The proposed license amendment makes no changes 
to the license requirements or to the design or operation of the 
containment atmosphere radioactivity monitor.
    Because no changes are made to either the design or operation of 
the sump monitor, the

[[Page 43049]]

sump monitor remains operable with the requested changes, and no 
changes are made to the containment atmosphere radioactivity 
monitor, FPC concludes that the change does not result in a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Section Chief: Richard P. Correia.

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 amendment requests: June 8, 2000.
    Description of amendment requests: The proposed amendments would 
approve changes to the Updated Final Safety Analysis Report (UFSAR) to 
allow the use of probabilistic risk assessment (PRA) techniques in 
evaluating the need for tornado-generated missile barriers.
    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 change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    The possibility of a tornado reaching the Donald C. Cook Nuclear 
Plant (CNP) site is a design basis event considered in the UFSAR. 
The proposed change does not affect the probability that a tornado 
will reach the CNP site. However, the change affects the probability 
assumed in the current licensing basis that missiles generated by 
the winds of a tornado might strike certain plant systems or 
components.
    No other accident scenarios, new initiators, or event precursors 
are affected or introduced by this change. There are a limited 
number of safety-related components that could potentially be struck 
by a tornado-generated missile. The total (aggregate) probability of 
exceeding 10 CFR 100 guidelines resulting from tornado missile 
strikes remains below the acceptance criterion ensuring overall 
plant safety. Thus, the proposed change does not constitute a 
significant increase in the probability of occurrence of an 
accident.
    This change does not result in an increase in the quantity of 
radioactive materials potentially available for release to the 
environment in the event of an accident. The principle barriers to 
the release of radioactive materials are not modified or affected by 
this change. No new release pathways are created. Thus, the proposed 
change does not significantly affect potential offsite dose 
consequences.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not significantly increased.
    (2) Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The possibility of a tornado reaching CNP site is a design basis 
event considered in the UFSAR. This change recognizes the 
acceptability of performing tornado missile probability calculations 
in accordance with established regulatory guidance. The change, 
therefore, deals with an established design basis event (the 
tornado). The change does not affect or create new accident 
initiators or precursors. Therefore, the change does not contribute 
to the possibility of a new or different kind of accident from those 
previously analyzed.
    Therefore, the change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    (3) Does the change involve a significant reduction in a margin 
of safety?
    The existing licensing basis for CNP, with respect to the design 
basis event of a tornado reaching the plant, generating missiles, 
and directing them toward safety-related systems and components, is 
to provide positive missile protection for every required SSC 
[System, Structure, and Component] or area. This change recognizes 
the extremely low probability, below an established acceptance 
limit, that a limited subset of SSCs, and areas could be struck. 
This change from ``protecting all required systems, structures, and 
components'' to an ``extremely low probability of exceeding 10 CFR 
100 guidelines as a result of tornado-generated missiles,'' does not 
constitute a significant decrease in the margin of safety due to the 
extremely low probability.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

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 amendment requests: May 15, 2000.
    Description of amendment requests: The proposed amendments would 
change Technical Specification 3.7.B.6 one time only to explicitly 
allow de-energizing Motor Control Center (MCC) 1T1 and MCC 1T2. The 
proposed change would allow either MCC 1T1 or MCC 1T2, one at a time, 
to be out of service for up to 72 hours provided the redundant MCC, its 
associated 480 Volt bus is verified operable, and the diesel generator 
and safeguards equipment associated with the redundant MCC are 
operable. The reason for the change is to install transfer switches for 
MCC 1T1 and MCC 1T2 for personnel protection, and to increase the 
allowed outage time for the MCC's to ensure sufficient time to install 
the transfer switches. This would prevent a dual unit shutdown to 
install each transfer switch under current Technical Specification 
3.7.B.6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment will not involve a significant 
increase in the probability or consequences of accident previously 
evaluated.
    The proposed changes do not involve any systems, structures or 
components whose failure would initiate an accident, thus, this 
change does not affect the probability of an accident.
    The proposed changes extend the allowed out of service time for 
MCC 1T1 and MCC 1T2. The proposed changes would be applied only in 
support of a one-time modification to install transfer switches for 
the affected MCC's. The proposed changes do not extend the allowed 
out of service time for any components, supplied by these MCC's, 
that are relied on to mitigate the consequences of an accident. 
Thus, this change does not significantly increase the expected 
consequences of an accident.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not change the way any systems, 
structures or components are operated. Nor does the proposed change 
introduce any new failure modes.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident.
    (3) The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes do not extend the allowed out of service times 
for any safety related components powered by the affected MCC's. 
Further, the proposed changes only allow one train (one of the affected 
MCC's) to be out of

[[Page 43050]]

service and only if the opposite train MCC, its supporting sources and 
supplied safeguards equipment is verified operable. Thus, the proposed 
changes do not substantially impact the ability of operators to protect 
the fuel cladding, reactor coolant system or containment.
    Therefore, the proposed changes will not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment requests: May 12, 2000.
    Description of amendment requests: The proposed amendment would 
allow the design upgrade of the refueling water purification (RWP) 
system from design class II/non-seismic category 1 to design class I/
seismic category 1 for purposes of permitting the cleanup of the 
refueling water storage tank (RWST) water while the RWST is required to 
be operable. This license amendment request (LAR) also proposes to 
allow the crediting of operator action to isolate a manual code 
boundary valve connected to the RWST following a seismic event or 
safety injection. It is desired to take suction from the RWST through 
an existing tank drain line to facilitate RWST recirculation through a 
non-seismically qualified reverse osmosis system while the RWST is 
required to be operable. This reverse osmosis system will be used to 
remove silica from the RWST water.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The upgrade of the refueling water purification (RWP) system 
piping will allow connection of the RWP system to the refueling 
water storage tank (RWST) while the RWST is required to be operable. 
The installation and use of a reverse osmosis (RO) system will allow 
removal of silica from the RWST while the RWST is required to be 
operable. The upgrade to the RWP system piping and use of the RO 
system does not involve any changes or create any new interfaces 
with the reactor coolant system or main steam system piping. 
Operation of the RWST is required to mitigate a loss-of-coolant and 
main steam line break accident, therefore, the connection of the RWP 
system to the RWST and use of the RO system would not affect the 
probability of these accidents occurring.
    Neither the RWP system nor the RO system are credited for safe 
shutdown of the plant or accident mitigation. The upgrade to the RWP 
system piping to seismic category I will prevent seismically induced 
failure of the RWP system piping and thus prevent a loss of RWST 
inventory while the RWP system is connected to the RWST. The RWST 
can perform its safety function with an active failure in the RWP 
system in the short term phase of an accident while the RWP system 
is connected to the RWST. The RWST can perform its safety function 
with an active or passive failure in the RO system in the short term 
phase of an accident. Since the RWST inventory is not credited in 
the long term phase of an accident, active and passive failures in 
the RWP or RO system in the long term phase of an accident need not 
be considered.
    Continuous operation of the RWP pump during a design basis event 
will not reduce the RWST water inventory nor the emergency core 
cooling system (ECCS) pump suction supply. The increase in RWST 
discharge flow due to an operating RWP pump will not adversely 
impact the required net positive suction head of the operating ECCS 
pumps.
    A combination of design and administrative controls ensure that 
both the RWP and RO systems maintain RWST boron concentration and 
tank volume requirements whenever the contents of the RWST are 
processed through these systems. Potential boron dilution or volume 
losses of the RWST inventory during tank processing through the RWP 
system is prevented by administratively maintaining closed all 
manual boundary valves within the RWP system while the RWP system is 
used to clarify RWST contents. Prior to RO system operation, the 
RWST volume margin will be verified to be adequate to compensate for 
postulated RO system line losses and process losses through the RO 
system reject waste stream. The waste stream losses will be 
monitored throughout RO system operation. The RO system is designed 
to maintain a high boron recovery rate, which will be verified 
through testing prior to initial installation. Potential boron 
dilution during each batch operation of the RO system is prevented 
through verifying RWST boron margin prior to RO system operation and 
monitoring the RO system boron recovery rate by grab samples taken 
of the system inlet and outlet after each batch operation. Following 
each batch operation of the RO system, RWST mixing and sampling will 
be performed to verify the RWST boron concentration, and boron 
additions to the RWST will be made accordingly. Since the RWST will 
continue to perform its safety function, overall system performance 
is not affected, assumptions previously made in evaluating the 
consequences of the accident are not altered, and the consequences 
of the accident are not increased.
    Therefore, the changes will not increase the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The upgrade of the RWP system piping to seismic category I will 
prevent seismically induced failure of the RWP piping. An active RWP 
pump failure will not result in a loss of the RWST safety function. 
An active or passive failure in the RO system will not result in 
loss of the RWST safety function. Adequate RWST volume and boron 
margin will be verified prior to RO system operation, the RO system 
boron recovery rate will be monitored by grab samples taken of the 
system inlet and outlet after each batch operation, a flow limiting 
device will limit the maximum potential RWST inventory loss rate to 
a low value, and operator action can be taken within 1 hour to 
isolate the RO system from the RWST. The upgrade to the RWP system 
and use of the RO system do not impact any other systems and thus 
cannot create a new failure mode in another system which could 
potentially create a new type of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Neither the RWP system nor the RO systems are credited for safe 
shutdown of the plant or accident mitigation. The upgrade to the RWP 
system piping to seismic category I will prevent seismically induced 
failure of the RWP system piping and prevent loss of RWST inventory 
due to a seismic event while the RWP system is connected to the 
RWST. The RWST can perform its safety function with an active 
failure in the RWP system in the short term phase of an accident 
while the RWP system is connected to the RWST. The RWST can perform 
its safety function with an active or passive failure in the RO 
system in the short term phase of an accident. Since the RWST 
inventory is not credited in the long term phase of an accident, 
active and passive failures in the RWP or RO system in the long term 
need not be considered.
    Adequate RWST volume and boron margin will be verified prior to 
RO system operation, a flow limiting device will limit the maximum 
inventory loss rate to a low value, and operator action can be taken 
within 1 hour to isolate the RO system from the RWST. The RO system 
waste stream losses will be monitored throughout RO system 
operation.
    Potential boron dilution of the RWST inventory during tank 
processing through the RWP system is prevented by administratively 
maintaining closed all manual boundary valves within the RWP system 
while the RWP system is connected to the RWST. The

[[Page 43051]]

RO system is designed to maintain a high boron recovery rate, which 
will be verified through testing prior to initial installation. 
Potential boron dilution during each batch operation of the RO 
system is prevented through verifying RWST boron margin prior to RO 
system operation and monitoring the RO system boron recovery rate by 
grab samples taken of the system inlet and outlet after each batch 
operation. Following each batch operation of the RO system, RWST 
mixing and sampling will be performed to verify the RWST boron 
concentration, and boron additions to the RWST will be made 
accordingly. These measures will ensure the TS minimum RWST boron 
concentration is available to mitigate the short term consequences 
of a small break LOCA, large break LOCA, or main steam line break 
accident.
    Therefore, the change does not involve a significant reduction 
in a margin of safety as defined in the basis for any technical 
specification.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon 
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California

    Date of amendment requests: June 2, 2000.
    Description of amendment requests: The proposed amendment would 
revise Technical Specification (TS) 3.5.2, ``ECCS--Operating,'' Action 
A, to change the allowed completion time for repair or replacement of 
the centrifugal charging pump (CCP) 2-1 during Cycle 10 of Unit 2 from 
72 hours to 7 days. In response to high CCP 2-1 vibration, planning has 
been done for replacing the CCP 2-1 discharge head and bearing housing 
or to change out the entire CCP 2-1. The 72-hour allowed completion 
time is not sufficient to accomplish such emergent repairs on an 
inoperable CCP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The emergency core cooling system (ECCS) and the centrifugal 
charging pumps (CCPs) are designed to respond to mitigate the 
consequences of an accident. They are not an accident initiator, and 
as such cannot increase the probability of an accident.
    The loss of both CCPs, due to an inoperable CCP 2-1 and a single 
failure of CCP 2-2, could increase the consequences of an accident. 
A PRA was performed to evaluate the increased consequences. The 
worst case risk increment due to the increased completion time for 
CCP 2-1 and the maximum allowed out of service time is 2.5 percent. 
This is a non-significant risk increase for core damage frequency 
(CDF). Also, there is no noticeable increase in the large early 
release frequency as a result of this request.
    Allowing 7 days to complete the repairs and post-maintenance 
testing of CCP 2-1 is acceptable since the ECCS system remains 
capable of performing its intended function of providing at least 
the minimum flow assumed in the accident analyses. During the 
extended maintenance and test period, appropriate compensatory 
measures will be implemented to restrict high risk activity. The 
consequences of accidents, which rely on the ECCS system, will not 
be significantly affected.
    Therefore, the proposed changes will not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no new failure modes or mechanisms created due to 
plant operation for an extended period to perform repairs and post-
maintenance testing of CCP 2-1. Extended operation with an 
inoperable CCP does not involve any modification in the operational 
limits or physical design of the systems. There are no new accident 
precursors generated due to the extended allowed completion time.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Plant operation for 7 days with an inoperable CCP 2-1 does not 
adversely affect the margin of safety. During the extended allowable 
completion time the ECCS system maintains the ability to perform its 
safety function of providing at least the minimum flow assumed in 
the accident analyses. During the extended maintenance and test 
period, appropriate compensatory measures will be implemented to 
restrict high risk activity.
    Therefore, the change does not involve a significant reduction 
in a margin of safety as defined in the basis for any Technical 
Specification.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: May 15, 2000.
    Description of amendment request: The proposed change is to LGS 
Unit 1 Technical Specifications (TSs) Figure 3.4.6.1-1, ``Minimum 
Reactor Vessel Metal Temperature vs. Reactor Vessel Pressure,'' and 
associated changes to TS Bases Section 3/4.4.6. The proposed change 
revises the pressure-temperature (P-T) limits by revising the heatup, 
cooldown and inservice test limitations for the Reactor Pressure Vessel 
(RPV) of Unit 1 from 12 effective full power years (EFPY) to a maximum 
of 32 EFPY. The proposed change also eliminates the requirement to 
maintain reactor coolant system within a narrow temperature band less 
that 212  deg.F during pressure testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    1. The proposed TS changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    There are no physical changes to the plant being introduced by the 
proposed changes to the P-T curves. The proposed changes do not modify 
the reactor coolant pressure boundary, i.e., there are no changes in 
operating pressure, materials or seismic loading. The proposed changes 
do not adversely affect the integrity of the reactor coolant pressure 
boundary such that its function in the control of radiological 
consequences is affected. The proposed P-T curves were generated in 
accordance with the fracture toughness requirements of 10 CFR Part 50, 
Appendix G, and American Society of Mechanical Engineers (ASME) Boiler 
and Pressure Vessel (B&PV) Code, Section XI, Appendix G, in conjunction 
with ASME Code Cases N-640 and N-588. The proposed P-T curves were 
established in compliance with the methodology used to calculate the

[[Page 43052]]

predicted irradiation effects on vessel beltline materials. Usage of 
these procedures provides compliance with the intent of 10 CFR Part 50, 
Appendix G, and provides margins of safety that ensure that failure of 
the reactor vessel will not occur. The proposed P-T curves prohibit 
operational conditions in which brittle fracture of reactor vessel 
materials is possible. Consequently, the primary coolant pressure 
boundary integrity will be maintained. Therefore, the 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.
    The proposed changes to the P-T curves were generated in accordance 
with the fracture toughness requirements of 10 CFR Part 50, Appendix G, 
and ASME B&PV Code, Section XI, Appendix G, in conjunction with ASME 
Code Cases N-640 and N-588. Compliance with the proposed P-T curves 
will ensure that conditions in which brittle fracture of primary 
coolant pressure boundary materials are possible will be avoided. No 
new modes of operation are introduced by the proposed changes. The 
proposed changes will not create any failure mode not bounded by 
previously evaluated accidents. Further, the proposed changes to the P-
T curves do not affect any activities or equipment, and are not assumed 
in any safety analysis to initiate any accident sequence. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes reflect an update of the P-T curves to extend 
the reactor pressure vessel operating limit to 32 Effective Power Years 
(EFPY). The revised curves are based on the latest ASME guidance. These 
proposed changes maintain the relative margin of safety commensurate 
with that which existed at the time that the ASME B&PV Code, Section 
XI, Appendix G, was approved in 1974. The revised pressure-temperature 
limits, although less restrictive than the current limits, were 
established in accordance with current regulations and the latest ASME 
Code information. Because operation will be within these limits, the 
reactor vessel materials will continue to behave in a non-brittle 
manner, thus preserving the original safety design bases. No plant 
safety limits, set points, or design parameters are adversely affected 
by the proposed TS changes. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.
    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: 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.

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

    Date of amendment request: May 15, 2000.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) requirements to test the remaining 
diesel generators when (1) One of the two independent off-site power 
sources is inoperable as delineated in TS 3/4.8.1, Action a, and (2) a 
diesel generator is inoperable for other than preventative maintenance 
reasons as delineated in TS 3/4.8.1, Action b.
    The proposed change also (1) Expands the diesel generator loading 
band for the monthly, six-month, and the two hour loaded pre-requisite 
requirement for the hot restart test in accordance with the guidance of 
Regulatory Guide 1.9, ``Selection, Design, Qualification, and Testing 
of Emergency Diesel Generator Units Used as Class 1E Onsite Electric 
Power Systems at Nuclear Power Plants,'' Rev. 3, 1993; and (2) corrects 
an administrative error in a note associated with TS 3.8.1.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The emergency diesel generator system is not an accident 
initiator. Eliminating the requirement to demonstrate that the 
operable diesel generators function properly, when there is no 
evidence that the inoperability of the affected diesel generator is 
the result of a potential common mode failure, will not increase the 
probability or the consequences of previously evaluated accidents, 
which rely upon emergency power supplies.
    Eliminating the testing of the diesel generators whenever a 
single off-site power source is inoperable does not establish 
operability of the remaining off-site power source. Operability is 
determined by the performance of surveillance 4.8.1.1.1.1.a.
    Elimination of unnecessary starts (challenges) to the diesel 
generators will result in increased equipment reliability and hence 
improved overall reliability for emergency onsite power supplies, as 
follows:
    (A) Reduce the overall engine degradation resulting from wear 
and tear of testing and reduce the probability of failure due to 
engine degradation, and,
    (B) Minimize the number of entries into an equipment 
configuration where a potential challenge to the safety function 
exists during the period of the tests.
    Expanding the band from 2500-2600 KW to 2330-2600 KW to 
accommodate instrument inaccuracy does not change any design 
parameter. The diesel generator will still be fully loaded (90% to 
100% of continuous rating) in accordance with Reg. Guide 1.9, Rev. 
3, Section 2.2.2. The full capability of the diesel generator to 
carry its load will continue to be demonstrated during the 24 
endurance run, which is unaffected by this request.
    The proposed change to the note in TS 3.8.1.2 is a correction of 
an administrative oversight (renumbering of a surveillance 
requirement) and does not change the surveillance content or intent.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    Eliminating the requirement to demonstrate that the operable 
diesel generators function properly affects testing requirements 
only and does not alter the physical configuration of the plant, 
replace or modify existing equipment, affect operating practices or 
create any new or different accident precursors.
    Similarly, expanding the band from 2500-2600 KW to 2330-2600 KW 
to accommodate instrument inaccuracy does not change the manner in 
which the diesel generator is operated, or introduces any new or 
different failure from any previously evaluated.
    The proposed change to the note in TS 3.8.1.2 is a correction of 
an administrative oversight (renumbering of a surveillance 
requirement) and does not change the surveillance content or intent.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
analyzed.
    3. Does not involve a significant reduction in a margin of 
safety.
    Eliminating the testing of the diesel generators whenever a 
single off-site power source is inoperable does not establish 
operability of the remaining off-site power source. Operability of 
the remaining off-site power source is determined by the performance 
of surveillance 4.8.1.1.1.1.a. The normally performed monthly 
surveillance ensures the diesel will be available to perform their 
safety function.
    Eliminating the requirement to demonstrate that the operable 
diesel

[[Page 43053]]

generators function properly, when there is no evidence that the 
inoperability of the affected diesel generator is the result of a 
potential common mode failure, does not reduce the margin of safety. 
If the evaluation is inconclusive or determines that a cause of 
inoperability for a diesel generator is a potential common mode 
failure then operability testing will be conducted for the remaining 
operable diesels. This action will assure that the initial 
assumption of two independent power supplies, utilized in the 
accident analysis, remain valid.
    The proposed changes do not adversely affect the ability of the 
diesels to operate when called upon. Rather, these changes should 
result in improved overall reliability of the diesels and therefore 
the margin of safety is preserved for those events in which there is 
a dependence upon on-site AC power supplies.
    Expanding the band from 2500-2600 KW to 2330-2600 KW to 
accommodate instrument inaccuracy does not introduce any new or 
different failure from any previously evaluated or changes the 
manner in which the diesel generator is operated. Expanding the band 
does not change any instrumentation set point, or changes to the 
auto loading sequence of the diesel. The capability of the diesel to 
be loaded to its manufactured maximum ratings will continue to be 
demonstrated during the performance of the diesel endurance run, 
which is unaffected by this request.
    The proposed change to the note in TS 3.8.1.2 is a correction of 
an administrative oversight (renumbering of a surveillance 
requirement) and does not change the surveillance content or intent.
    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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.
    TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Units 1 and 2, Somervell County, Texas
    Date of amendment request: May 25, 2000.
    Brief description of amendment: The proposed amendment would revise 
the Comanche Peak Steam Electric Station, Units 1 and 2, Technical 
Specifications, Limiting Condition for Operation (LCO) 3.9.4, 
``Containment Penetrations,'' to allow certain containment penetrations 
to be open during refueling activities under appropriate administrative 
controls.
    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 change involve a significant increase in 
the probability or consequences of an accident previously evaluated? 

    Response: No.
    The proposed change to Technical Specification (TS) 3.9.4, 
``Containment Penetrations,'' would allow certain containment 
penetration flow paths to be open during core alterations and 
movement of irradiated fuel within containment under specific 
administrative controls. The fuel handling accident [(FHA)] 
radiological analysis does not take credit for containment isolation 
or filtration. Therefore, the time required to close any open 
penetrations is not relevant to the confirmatory radiological 
analysis dose calculations and the proposed change does not involve 
a significant increase in the consequences of an accident previously 
evaluated. The proposed administrative controls for containment 
penetrations are conservative even though not required by the 
accident analysis.
    The status of the penetration flow paths during refueling 
operations has no affect on the probability of the occurrence of any 
accident previously evaluated. The proposed revision does not alter 
any plant equipment or operating practices in such a manner that the 
probability of an accident is increased. Because the FHA outside 
containment remains the limiting accident and the probability of a 
accident is not affected by the status of the penetration flow 
paths, the proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated? 
    Response: No.
    The open containment penetration flow paths are not accident 
initiators and do not represent a significant change in the 
configuration of the plant. The proposed allowance to open the 
containment penetrations during refueling operations will not 
adversely affect plant safety functions or equipment operating 
practices such that a new or different accident could be created. 
Therefore, the proposed revision will not create a new or different 
kind of accident from any accident previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety? 
    Response: No.
    Technical Specification LCO 3.9.4 closure requirements for 
containment penetrations ensure that the consequences of a 
postulated FHA inside containment during core alterations or fuel 
handling activities are minimized. The LCO establishes containment 
closure requirements, which limit the potential escape paths for 
fission products by ensuring that there is at least one integral 
barrier to the release of radioactive material. The proposed change 
to allow the containment penetration flow paths to be open during 
refueling operations under administrative controls does not 
significantly affect the expected dose consequences of a FHA because 
the limiting FHA is not changed. The proposed administrative 
controls provide assurance that prompt closure of the penetration 
flow paths will be accomplished in the event of a FHA inside 
containment thus minimizing the transmission of radioactive material 
from the containment to the outside environment. Under the proposed 
TS change, the provisions to promptly isolate open penetration flow 
paths provide assurance that the offsite dose consequences of a FHA 
inside containment will be minimized. Therefore, the proposed change 
to the Technical Specifications does not involve a significant 
reduction in the margin of safety.

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

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

    Date of amendment request: June 23, 2000.
    Description of amendment request: The proposed amendment would 
revise Limiting Condition for Operation (LCO) 3.9.4, ``Containment 
Penetrations,'' of the technical specifications (TS) to allow certain 
containment penetrations to be open during refueling operations under 
administrative controls. The amendment would (1) Revise the note in the 
LCO for containment penetrations that may be open under administrative 
controls, deleting the reference to penetrations P-63 and P-98, and (2) 
delete the exception for penetrations P-63 and P-98 in Surveillance 
Requirement (SR) 3.9.4.1. In addition, there would be format and 
editorial corrections to TS 3.8.3, ``Diesel Fuel Oil, Lube Oil, and 
Start Air,'' and TS 5.2.2.b, ``Administrative Controls,'' to remove 
errors in the conversion to improved TSs issued March 31, 1999, in 
Amendment No. 123. There are also changes to the TS Bases for the 
proposed changes to LCO 3.9.4 and SR 3.9.4.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 43054]]

issue of no significant hazards consideration, which is presented 
below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. 
    The status of the penetration flow paths during refueling 
operations has no [effect] on the probability of the occurrence of 
any accident previously evaluated. The proposed revision does not 
alter any plant equipment or operating practices in such a manner 
that the probability of an accident is increased. Since the 
consequences of a FHA [fuel handling accident] inside containment 
with open penetration flow paths are bounded by the current analysis 
described in the USAR [updated safety analysis report for Wolf 
Creek] and the probability of an accident is not affected by the 
status of the penetration flow paths, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed changes to correct editorial/format errors involve 
corrections to the technical specifications that are associated with 
the original conversion application and supplements or the certified 
copy of the improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements in the technical 
specifications.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The open containment penetration flow paths are not accident 
initiators and do not represent a significant change in the 
configuration of the plant. The proposed allowance to open the 
containment penetrations during refueling operations will not 
adversely affect plant safety functions or equipment operating 
practices such that a new or different accident could be created.
    The proposed changes to correct editorial/format errors involve 
corrections to the technical specifications that are associated with 
the original conversion application and supplements or the certified 
copy of the improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements [in] the 
technical specifications.
    Therefore, the proposed revision will not create a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Technical Specification LCO 3.9.4 closure requirements for 
containment penetrations ensure that the consequences of a 
postulated FHA inside containment during core alterations or fuel 
handling activities are minimized. The LCO establishes containment 
closure requirements, which limit the potential escape paths for 
fission products by ensuring that there is at least one integral 
barrier to the release of radioactive material. The proposed change 
to allow the containment penetration flow paths to be open during 
refueling operations under administrative controls does not 
significantly affect the expected dose consequences of a FHA because 
the limiting FHA is not changed. The proposed administrative 
controls provide assurance that prompt closure of the penetration 
flow paths will be accomplished in the event of a FHA inside 
containment thus minimizing the transmission of radioactive material 
from the containment to the outside environment. Under the proposed 
TS change, the provisions to promptly isolate open penetration flow 
paths provide assurance that the offsite dose consequences of a FHA 
inside containment will be minimized.
    The proposed changes to correct editorial/format errors involve 
corrections to the technical specifications that are associated with 
the original conversion application and supplements or the certified 
copy of the improved Technical Specifications. As such, these 
changes are considered as administrative changes and do not modify, 
add, delete, or relocate any technical requirements in the technical 
specifications.
    Therefore, the proposed changes to the Technical Specifications 
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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Section Chief: Stephen Dembek.

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.

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

    Date of amendment request: May 22, 2000.
    Description of amendment request: The proposed amendment would add 
new Technical Specifications (TSs) 3.7.2.a(ii) and 3.7.2.h to address 
voltage on the 230 kV (kilovolt) grid as a precondition of criticality 
and to provide a time limit for when the 230 kV grid voltage is found 
to be insufficient to support Loss-of-Coolant Accident (LOCA) 
electrical loading during power operation. The application also 
requests various minor editorial changes. The Bases have also been 
changed to reflect the addition of the two new TS and to provide 
clarification of the components to which surveillance is applicable.
    Date of publication of individual notice in Federal Register: June 
2, 2000 (65 FR 35404).
    Expiration date of individual notice: July 3, 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.

[[Page 43055]]

    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. 50-289, Three Mile Island 
Nuclear Station, Unit 1, Dauphin County, Pennsylvania

    Date of application for amendment: May 4, 2000, as supplemented May 
9, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 4.12.1.3, for the control building automatic 
isolation and recirculation dampers to remove the individual damper 
component tag numbers. The surveillance requirements do not change. The 
associated Bases is also changed to reflect the applicable section of 
the Updated Final Safety Analysis Report.
    Date of issuance: June 29, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 223.
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 22, 2000 (65 FR 
32132).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 29, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 19, 1999, as 
supplemented April 21, 2000.
    Brief description of amendments: The amendments approved changes in 
the Updated Final Safety Analysis Report (UFSAR) that constitute an 
unreviewed safety question as described in 10 CFR 50.59. These changes 
increase the probability of occurrence of a malfunction. These changes 
were not previously evaluated in the UFSAR, specifically, Section 
5.3.1, ``External Missiles'' of the UFSAR did not address the 
probability of a missile from Unit 1 turbine-generator striking: (1) 
The refueling water tanks, (2) the No. 11 fuel oil storage tank, and 
(3) the plant equipment through various roof slabs or through non-
missile-proof openings in the missile-proofing walls. The UFSAR only 
discusses a turbine missile strikingthe containment, control room, 
switchgear room, and waste processing area. The amendment authorizes 
the licensee to revise the turbine missile analysis to include the 
additional targets.
    Date of issuance: June 19, 2000.
    Effective date: As of the date of issuance to be implemented by 
December 31, 2000.
    Amendment Nos.: 236 and 210.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70079).
    The April 21, 2000, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated June 19, 2000.
    No significant hazards consideration comments received: No.

Consolidated Edison Company of New York, Inc., Docket Nos. 50-003 and 
50-247, Indian Point Nuclear Generating Station, Units 1 and 2, 
Buchanan, New York.

    Date of amendment request: February 14, 2000.
    Brief description of amendments: The amendments would eliminate 
from Environmental Technical Specifications Section 5.4.1, Routine 
Reports, the discussion regarding Section 4.2. Specifically, the 
proposed change seeks to delete the reference to and discussion about 
Section 4.2, which was deleted as part of Amendment No. 90 to Operating 
License No. DPR-26.
    Date of issuance: June 8, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 47 to DPR-5, and 210 to DPR-26.
    Facility Operating License Nos. DPR-5 and DPR-26: The amendments 
revised the Environmental Technical Specifications.
    Date of initial notice in Federal Register: April 5, 2000 (65 FR 
17912). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 8, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: April 13, 2000, as supplemented 
by letter dated May 30, 2000.
    Brief description of amendments: The amendments revise the 
Technical Specifications and associated Bases pages to accommodate the 
use of Mark-B11 fuel with M5 cladding.
    Date of Issuance: June 21, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 313, 313, and 313.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in  Federal Register: May 17, 2000 (65 FR 
31356).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 21, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: July 29, 1999.
    Brief description of amendment: The amendment revised Surveillance 
Requirement 3.5.2.2. The change requires maintaining a higher level in 
the condensate storage tanks.
    Date of issuance: June 20, 2000.
    Effective date: June 20, 2000, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 165.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 25, 1999 (64 FR 
46431).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 20, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 1, 1999, as 
supplemented by letter dated May 10, 2000.
    Brief description of amendment: This amendment revised the 
frequency of performing Technical Specification Surveillance 
Requirement (SR) 3.6.1.7.4, verification that each containment spray 
nozzle is unobstructed. The frequency

[[Page 43056]]

for performing SR 3.6.1.7.4 has been changed from once every 10 years 
to conditions following maintenance which could result in nozzle 
blockage.
    Date of issuance: June 29, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days.
    Amendment No.: 113.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 15, 1999 (64 
FR 70088).
    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 June 29, 2000.
    No significant hazards consideration comments received: No.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear 
Generating Station, Ocean County, New Jersey

    Date of application for amendment: July 7, 1999.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to change the component surveillance frequencies 
for the following TSs to indicate a frequency of once per 3 months: 
Core Spray System TS 4.4.A.1 and 4.4.A.2, Containment Cooling System TS 
4.4.C.1, Emergency Service Water System TS 4.4.D.1, Fire Protection 
System TS 4.4.F (isolation valves only), and Pressure Suppression 
Chamber--Drywell Vacuum Breakers TS 4.5.F.5.a. The TSs currently 
stipulate a component surveillance frequency of once per month. Also, 
the amendment revised TS pages 4.4-1 and 4.4-2 to incorporate editorial 
format changes and TS page 4.4-3 to accommodate the expanded text.
    Date of Issuance: June 26, 2000.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 210.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications. Date of initial notice in Federal Register: 
October 20, 1999 (64 FR 56531).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated June 26, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 29, 1999.
    Description of amendment request: The amendment relocates 
Surveillance Requirement 4.8.1.1.2f.1 which requires inspection of the 
Emergency Diesel Generator (EDGs) at least once per 18 months in 
accordance with procedures prepared in conjunction with its 
manufacturer from the Technical Specifications to the Seabrook Station 
Technical Requirements Manual.
    Date of issuance: June 16, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 90 days.
    Amendment No.: 71.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4281).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: November 30, 1999, as supplemented on 
April 28, 2000.
    Description of amendment request: The amendment revises the 
Technical Specifications (TSs) by: (1) Inclusion of a new 
Administrative Control TS 6.7.6i for establishing, implementing, and 
maintaining a Diesel Fuel Oil Testing Program for testing new and 
stored fuel oil; (2) relocation of current surveillance requirement 
(SR) 4.8.1.1.2d and SR 4.8.1.1.2e.1, containing SRs for fuel oil 
sampling and testing, to the Diesel Fuel Oil Testing Program in the 
Seabrook Station Technical Requirements (SSTR) Manual; (3) revision of 
SR 4.8.1.1.2d to reference the Diesel Fuel Oil Testing Program as a 
surveillance requirement; (4) inclusion of additional surveillance 
requirements to SR 4.8.1.2 for checking and removing accumulated water 
from the day and storage fuel oil tanks, verifying new and stored fuel 
oil properties and visually inspecting diesel generator exhaust leakage 
when the plant remains in Modes 5 and 6 of operation; (5) relocation to 
the Diesel Fuel Oil Testing Program SR 4.8.1.12h for cleaning diesel 
fuel storage tanks at a 10-year frequency to the SSTR Manual; and (6) 
revision of TS Bases 3/4.8.1 by adding a statement that the exceptions 
to the certain Regulatory Guides are specified in the plant's Updated 
Final Safety Analysis Report.
    Date of issuance: June 27, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 73.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in  Federal Register: May 17, 2000 (65 FR 
31358).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 27, 2000.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: April 14, 2000.
    Description of amendment request: This amendment revises the 
Technical Specifications by relocating Sections 3/4.9.5, 
``Communications'', 3/4.9.6, ``Refueling Machine'', and 3/4.9.7, 
``Crane Travel--Spent Fuel Storage Areas'' to the Seabrook Station 
Technical Requirement Manual.
    Date of issuance: June 23, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 72.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 17, 2000 (65 FR 
31358).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 23, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: March 16, 2000, as supplemented 
by letters dated April 11, April 19, June 2, and June 9, 2000.
    Brief description of amendments: The amendments revise several 
sections of the improved Technical Specification (ITS) to correct 19 
editorial errors made in either (1) the application dated June 2, 1997, 
(and supplemental letters) for the ITSs, or (2) the certified copy of 
the ITSs that was submitted in the licensee's letters of May 19 and 27, 
1999. The proposed amendment would

[[Page 43057]]

also revise 10 instances of incorrect incorporation of the CTS into the 
ITS. One of the proposed editorial errors and one of the incorrect 
incorporations of the CTS will be addressed in a future letter. The 
ITSs were issued as License Amendments 135 and 135 dated May 28, 1999.
    Date of issuance: June 21, 2000.
    Effective date: June 21, 2000, to be implemented by June 30, 2000.
    Amendment Nos.: Unit 1-142; Unit 2-142
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 19, 2000 (65 FR 
21032).
    The April 19, June 2, and June 9, 2000, supplemental letters 
provided additional clarifying information, did not expand the scope of 
the application as originally noticed, and did not change the staff's 
original proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 21, 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: December 27, 1999, as 
supplemented April 11, 2000.
    Brief description of amendment: This amendment revises Technical 
Specifications (TSs) 4.6.2.2.b, ``Suppression Pool Spray,'' and 
4.6.2.3.b, ``Suppression Pool Cooling,'' to modify the acceptance 
criteria associated with flow rate testing of the Residual Heat Removal 
system pumps.
    Date of issuance: June 16, 2000.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 128.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: January 26, 2000 (65 FR 
4289)
    The April 11, 2000, supplement provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 16, 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: November 5, 1999, as 
supplemented December 3, 1999.
    Brief description of amendment: The amendment revises the 
applicability for the reactor power distribution limits and Average 
Power Range Monitor gain adjustments. The applicability is revised to 
operation at  25% Rated Thermal Power.
    Date of Issuance: June 21, 2000.
    Effective date: As of its date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 188.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 29, 1999 (64 
FR 73102)
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated June 21, 2000.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: January 19, 2000.
    Brief description of amendments: These amendments revise Technical 
Specification 15.4.4-II.A to clarify that a different primary 
containment tendon may be designated a control tendon providing that 
the new control tendon has not previously been physically changed 
(e.g., retensioned).
    Date of issuance: June 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment Nos.: 196 and 201.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 8, 2000 (65 FR 
12295).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 27, 2000.
    No significant hazards consideration comments received: No.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: October 27, 1998, as 
supplemented on February 23, 2000.
    Brief description of amendment: The amendment revises the plugging 
limits specified in TS 4.2.b, ``Steam Generator Tubes,'' for the 
Westinghouse hybrid-expansion-joint sleeve and the Westinghouse laser-
welded sleeve. The proposed amendment also revises the list of 
applicable references specified in TS 4.2.b.
    Date of issuance: June 27, 2000.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 148.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64126). The February 23, 2000, supplement is within the scope of the 
original notice and does not change the proposed no significant hazards 
consideration finding.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 27, 2000.
    No significant hazards consideration comments received: No.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power 
Station, Franklin County, Massachusetts

    Date of application for amendment: March 17, 1999, as supplemented 
April 23, July 21, and November 2, 1999, and March 6, 2000.
    Brief description of amendment: The amendment revises Technical 
Specification Section 6.0, Administrative Controls, by consolidating 
management positions and modifying review and audit functions.
    Date of issuance: June 20, 2000.
    Effective date: June 20, 2000.
    Amendment No.: 154.
    Facility Operating License No. DPR-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17033) The April 23, July 21, and November 2, 1999, and March 6, 2000, 
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 June 20, 2000.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 5th day of July 2000.

[[Page 43058]]

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-17625 Filed 7-11-00; 8:45 am]
BILLING CODE 7590-01-P