[Federal Register Volume 66, Number 113 (Tuesday, June 12, 2001)]
[Notices]
[Pages 31700-31719]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-14755]


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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 May 21, 2001 through June 1, 2001. The last 
biweekly notice was published on May 30, 2001 (66 FR 29349).

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 31701]]

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 and 
Directives Branch, Division of Administrative 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, located at One White Flint North, 11555 Rockville Pike (first 
floor), Rockville, Maryland 20852. The filing of requests for a hearing 
and petitions for leave to intervene is discussed below.
    By July 13, 2001, 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, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland 20852. Publicly available records will be accessible 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: Rulemaking and 
Adjudications Branch, or may be delivered to the Commission's Public 
Document Room, located at One White Flint North, 11555 Rockville Pike 
(first floor), Rockville, Maryland 20852, 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, located at One White Flint 
North, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management Systems (ADAMS) Public Electronic

[[Page 31702]]

Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by email to <a href="/cdn-cgi/l/email-protection#a1d1c5d3e1cfd3c28fc6ced7"><span class="__cf_email__" data-cfemail="f4849086b49a8697da939b82">[email protected]</span></a>.

AmerGen Energy Company, LLC,. et al., Docket No. 50-219, Oyster 
Creek Nuclear Generating Station, Ocean County, New Jersey

    Date of amendment request: April 6, 2001.
    Description of amendment request: The existing Oyster Creek 
Technical Specification (TS) Section 4.7.B.5 requires capacity testing 
of the Station Batteries and the Diesel Generator Starting Batteries at 
least once per 24 months during a plant shutdown. The proposed 
amendment request will allow the 24-month capacity test for the Diesel 
Generator Starting Batteries to be performed during plant shutdowns or 
during the 24-month on-line Diesel Generator inspection (TS 4.7.A.3). 
The proposed revision to Section 4.7.B.5.b also reflects this change in 
specified frequency.
    Additionally, TS 4.7.A.5 is revised to delete the statement that 
the battery capacity test need not be performed if the installed 
batteries were replaced during the previous Diesel Generator on-line 
biennial inspection. This exception is no longer necessary because the 
battery capacity testing is not restricted to refueling outages based 
on the proposed change to Section 4.7.B.5.
    TS 4.7.B.5.a is revised to delete the phrase ``* * * to be 
considered operable'' because all of the specified surveillances 
constitute operability criteria. The title of Section 4.7.B is revised 
to identify applicability to the Diesel Generating Starting Batteries. 
These additional proposed revisions are considered administrative 
changes, which clarify the existing TS.
    TS 4.7 Bases is also revised to reflect the above specification 
changes. Section 4.7 Bases contained on page 4.7-3 are being relocated 
to Bases page 4.7-4. This relocation of the Bases is a purely 
administrative change.
    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 change to allow the batteries to be tested during the 24-
month Diesel Generator inspection outage does not increase the 
probability of occurrence of an accident previously evaluated. No 
change is being made to equipment, equipment operation, or equipment 
requirements. If a Diesel Generator battery were to fail during the 
24-month inspection, the availability of the Diesel Generator will 
not be affected because the Diesel Generator will already be out of 
service for the inspection. The change will allow the Diesel 
Generator out of service time during refueling outages to be reduced 
or eliminated, thereby reducing risk.
    The change to allow the batteries to be tested during the 24-
month Diesel Generator inspection outage does not increase the 
consequences of an accident previously evaluated. No change is being 
made to equipment, equipment operation, or equipment requirements. 
If a Diesel Generator battery were to fail during the 24-month 
inspection, the consequences of the battery failing are not 
affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change to allow the batteries to be tested during the 24-
month Diesel Generator inspection outage does not create the 
possibility of a new or different kind of accident from any 
previously evaluated. Moving the testing will not create a new 
possible failure type, it will only move the detection of a battery 
failure from the refueling outage to the 24-month Diesel Generator 
inspection outage.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The change to allow the batteries to be tested during the 24-
month Diesel Generator inspection outage does not reduce a margin of 
safety. Since the Diesel Generator will already be out of service 
for the 24-month inspection, the margin of safety for the Diesel 
Generator 24-month inspection outage will not be affected. The 
change will allow the Diesel Generator out of service time during 
the refueling outage to be reduced or eliminated, thereby increasing 
the margin of safety during the refueling outage.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: January 23, 2001.
    Description of amendment request: The proposed amendment revises 
the requirements for containment integrity associated with the 
personnel and emergency air locks and other penetrations during fuel 
movement and refueling operations to allow these penetrations to remain 
open. One door in each of the emergency and personnel air locks must be 
capable of being closed and each penetration providing direct access 
from the containment atmosphere to the outside atmosphere shall be 
capable of being closed by an isolation valve, blind flange, or manual 
valve. The supporting revised design basis fuel handling accident 
inside containment analysis will also incorporate alternative source 
term methodology in accordance with Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.67 and Regulatory Guide 1.183, 
``Alternative Radiological Source Terms For Evaluating Design Basis 
Accidents At Nuclear Power Reactors,'' July 2000. Technical 
Specification (TS) 3.8.7 is also revised to provide equivalent 
isolation methods for other penetrations consistent with Babcock & 
Wilcox Owner's Group (BWOG) Standard Technical Specifications (STSs), 
Section 3.9.3.c.1, NUREG-1430, April 1995. TS 3.8.11 is added to 
specify the requirement to maintain at least 23 feet of water over the 
top of the reactor vessel flange and the actions required if this level 
is not maintained. TS Bases 3.8 is revised to provide a description of 
the plant conditions under which the personnel and emergency air locks 
and other penetrations including those consistent with the BWOG STSs, 
Section 3.9., may be open during fuel movement, and the administrative 
controls that would be in place. The surveillance requirements of TS 
4.4.1.3 are also revised to identify the exception allowed by TS 3.8.6 
under which both doors of the personnel and emergency air locks can be 
open.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 31703]]

consideration, which is presented below:

    1. Will the operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: The proposed change would allow the personnel and 
emergency air lock doors and other penetrations to remain open 
during fuel loading and refueling operations. These penetrations 
were previously closed during this time period in order to prevent 
the escape of radioactive material in the event of a fuel handling 
accident inside containment (FHA). These penetrations are not 
initiators of any accident. The probability of a FHA is unaffected 
by the position of these penetrations.
    The new FHA analysis utilizing an Alternative Source Term with 
an open containment demonstrates that the maximum doses are well 
within the acceptance criteria specified in 10 CFR 50.67 and 
Regulatory Guide 1.183. In the event of a fuel handling accident, 
actual control room and offsite doses will be less than analyzed 
values because containment integrity will be restored following an 
evacuation of containment. As noted above, with the Alternative 
Source Term implementation, the acceptance criteria are also being 
revised. A direct comparison of the new Alternative Source Term dose 
consequences with the existing licensing basis FHA source term dose 
consequences is not practical due to the significant differences in 
methodology and assumptions.
    However, a comparison of the previous thyroid and whole body 
dose results for the postulated TMI Unit 1 FHA Inside Containment 
documented in the TMI Unit 1UFSAR [updated final safety analysis 
report] Chapter 14 with the new dose results expressed in terms of 
Total Effective Dose Equivalent (TEDE), using the guidance in 
Regulatory Guide 1.183 Footnote 7, indicates that the new doses are 
not significantly higher than the previous dose results. The revised 
Alternative Source Term calculated doses remain well within the 
allowable acceptance criteria.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will the operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    Response: The proposed change does not involve the addition or 
modification of any plant equipment. Also, the proposed change would 
not alter the design or method of operation of the plant beyond the 
standard functional capabilities of the equipment. The proposed 
change involves a change to the Technical Specifications that would 
allow the personnel and emergency air lock doors and other 
penetrations to be open during fuel loading and refueling operations 
within the containment. Having these doors and penetrations open 
does not create the possibility of a new accident. Administrative 
provisions will be made to ensure the capability to close the 
containment in the event of a FHA inside containment.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: This proposed change has the potential for an 
increased postulated accident dose due to a FHA Inside Containment; 
However, the analysis demonstrates that the resultant doses are well 
within the appropriate acceptance criteria. The margin of safety, as 
defined by 10 CFR 50.67 and Regulatory Guide 1.183, has been 
maintained. The offsite and control room doses due to a FHA with an 
open containment have been evaluated with conservative assumptions, 
which ensure the calculation bounds the postulated accident dose. 
Closing at least one door in each of the personnel and emergency air 
locks following the evacuation of the containment and closure of 
other open penetrations would reduce the control room and offsite 
doses in the event of a FHA inside containment and provides 
additional margin to the calculated doses.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: January 29, 2001.
    Description of amendment request: The proposed amendment revises 
the Technical Specifications (TSs) to remove the note from TS 4.5.4.1 
that restricts the applicability of the specified engineered safeguards 
feature (ESF) systems leakage rate limit of 15 gallons per hour to the 
current operating Cycle 13 and establish this value as the permanent TS 
limit. This limit had previously been approved with the issuance of 
Amendment No. 215 on August 24, 1999, for Cycle 13 only. The proposed 
amendment also would implement a full scope alternative source term for 
Three Mile Island Nuclear Station, Unit 1, in accordance with Title 10 
of the Code of Federal Regulations (10 CFR) Section 50.67 and the 
guidance contained in Regulatory Guide 1.183.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed permanent Technical Specification limit on ESF 
Systems leak rate is identical with the existing licensing basis 
value and is conservatively reevaluated for the limiting design 
basis Maximum Hypothetical Accident (MHA) using alternative source 
term methodology. Implementation of the alternative source term in 
accordance with Regulatory Guide 1.183 does not affect the design or 
operation of the facility, and therefore, does not significantly 
increase the probability of an accident previously evaluated. Based 
on the results of this reanalysis, it has been demonstrated that 
with the requested Technical Specification change, the offsite and 
control room dose consequences for this limiting event remain within 
the allowable dose criteria specified in 10 CFR 50.67 and Regulatory 
Guide 1.183.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed permanent Technical Specification limit on ESF leak 
rate and implementation of the alternative source term in accordance 
with Regulatory Guide 1.183 does not affect the design, functional 
performance, or operation of the facility or of any equipment within 
the facility. Modifications supporting the proposed change have been 
evaluated and determined not to create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed change involves implementation of the alternative 
source term in accordance with 10 CFR 50.67 and Regulatory Guide 
1.183, and maintains the current Technical Specification limit on 
ESF Systems leak rate. The reanalysis of the limiting design basis 
MHA has been performed using conservative methodologies as specified 
in Regulatory Guide 1.183. Margin has been maintained to ensure that

[[Page 31704]]

the accident analysis dose consequences bound the postulated event 
scenarios. The calculated offsite and control room dose consequences 
for this limiting event are within the acceptance criteria as 
specified in 10 CFR 50.67 and Regulatory Guide 1.183.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy 
Company, 2301 Market Street, S23-1, Philadelphia, PA 19103.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: May 8, 2001.
    Description of amendment request: The proposed amendment would 
revise the frequency of the Technical Specification (TS) surveillance 
requirement to check the movement of the control rods. Specifically, 
the frequency listed for this requirement in TS Table 4.1-3, 
``Frequencies for Equipment Tests,'' would be changed from ``every 31 
days'' to ``quarterly'' during reactor critical operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability [...] or consequences of an accident previously 
evaluated.
    This change to the frequency of performance of surveillance does 
not result in any hardware changes or nor does it change the 
response of control rods in performing their specified function. 
Therefore the change cannot affect the probability of occurrence of 
previously evaluated accidents.
    The proposed frequency has been determined to be adequate to 
assure the reliability of reactor trip based on the conclusions in 
NUREG 1366 [``Improvements to Technical Specification Surveillance 
Requirements''] and the recommendations of GL [Generic Letter] 93-05 
[``Line-Item Technical Specifications Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation''].
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change does not introduce a new failure mechanism 
or a new or different type of accident than those previously 
evaluated since there are no physical changes being made to the 
facility. Performance of the surveillance on the revised frequency 
will not have an adverse affect on the ability of the control rods 
to perform their intended function. The proposed change does not 
degrade the reliability of systems, structures, or components or 
create a new accident initiator or precursor. Therefore, the change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in [a] margin of 
safety.
    The proposed reduction in surveillance testing reduces the risk 
for causing dropped rods or reactor trips. This results in a slight 
improvement in the margin of safety by decreasing challenges to 
reactor components and safety systems.
    The proposed surveillance frequency, as supported by the 
industry experience described in NUREG-1366, continues to provide 
the required assurance of control rod operability, such that safety 
margins established through the design and facility license, 
including the Technical Specifications, remain unchanged.
    Therefore, operation of the facility in accordance with the 
proposed amendment is expected to result in a slight net improvement 
in [a] margin of safety. Hence the proposed change would not involve 
a significant reduction in [a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Richard P. Correia, Acting.

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

    Date of amendment request: May 10, 2001.
    Description of amendment request: The proposed amendment would 
remove Technical Specification (TS) surveillance requirement (SR) 
4.6.A.4 that requires each emergency diesel generator (EDG) to be given 
a thorough inspection at least annually following the manufacturer's 
recommendations. The requirement for the EDG inspection will be 
relocated to the Updated Final Safety Analysis Report and will be in 
accordance with the licensee controlled maintenance program. The 
inspection period required by the maintenance program will also be 
changed to specify that it will be ``in accordance with the 
manufacturer's recommendations.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability [...] or consequences of an accident previously 
evaluated.
    There is no change to the design, function, or capability of the 
EDGs as a result of this change. Hence there is no change in the 
probability of occurrence of an accident previously evaluated.
    The change does not affect the ability of the EDGs to mitigate 
the consequences of any accident previously evaluated; including the 
loss of coolant accident coupled with loss of offsite power. To the 
contrary, this change is structured to enhance the availability and 
reliability of the EDGs by tailoring the actual EDG maintenance 
program to the EDGs' operational history and experience. In 
addition, the surveillance testing requirements of TS Surveillance 
Requirements 4.6.A.1, 2 & 3 have not changed and are adequate to 
verify the operability of the EDG system. And, the Maintenance Rule 
Program at IP2 [Indian Point Unit 2] has established specific 
performance criteria for the EDGs. These performance criteria, and 
requirements to ensure the criteria are met, are not affected by 
this change.
    The deletion of the surveillance requirement and controlling EDG 
maintenance using a licensee-controlled maintenance program does not 
alter or prevent the ability of the EDGs to perform their intended 
functions.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The EDG is not an accident initiator. The proposed change does 
not involve any physical design change or operational change. Thus a 
new failure mode is not introduced. In addition, the proposed change 
has been evaluated to not degrade the reliability of any existing 
system, structure, or component. Therefore, the proposed

[[Page 31705]]

change does not create a new accident initiator or precursor, or 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in [a] margin of 
safety.
    As a result of this change, there are no changes to IP2's design 
or to the IP2 TS safety limits, limiting safety system settings, or 
limiting conditions [for] operation. A single SR is replaced by a 
performance-based maintenance program.
    The substitution of the performance-based maintenance program 
for the prescriptive SR is expected to increase the availability of 
the EDGs because the amount of time the EDGs are out-of-service for 
on-line maintenance will decrease. Reducing the number of plant 
operating hours that the unit is exposed to an out-of-service EDG 
improves rather than reduces the margin of safety. The substitution 
of the performance-based maintenance program for the prescriptive SR 
is expected to improve the reliability of the EDGs by minimizing the 
possibility of adverse results that may result from intrusive 
maintenance activities. The expected reliability improvement 
improves rather than reduces [a] margin of safety.
    The transfer of control of EDG maintenance from the TS to a 
licensee-controlled EDG maintenance program is an administrative 
change. But the change is structured so that maintenance program 
changes must be evaluated using the 10 CFR 50.59 process. Use of the 
10 CFR 50.59 process assures that future changes to the EDG 
maintenance program cannot significantly increase the likelihood of 
a malfunction of the EDGs. And use of the 10 CFR 50.59 process, 
instead of the license amendment process, allows Con Edison to 
optimize EDG maintenance in a timely manner to meet the intent of 10 
CFR 50.65.
    The proposed changes do not adversely affect the EDG's ability 
to function when required to mitigate any accident or licensing 
basis event. Therefore, operation of the facility in accordance with 
the proposed amendment would not involve a significant reduction in 
[a] margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Section Chief: Richard P. Correia, Acting.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: April 11, 2001.
    Description of amendment request: The proposed amendment would 
revise Technical Specification definitions 1.12, ``Core Alteration;'' 
3.9.1, ``Refueling Operations--Boron Concentration;'' 3.9.2, 
``Refueling Operations--Instrumentation;'' and 3.9.11, ``Refueling 
Operations--Water Level--Reactor Vessel.'' The Bases for these 
Technical Specifications would also be modified to reflect the proposed 
changes to these definitions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification changes associated with the 
definition for Core Alteration and LCO [limiting condition for 
operation], applicability, action requirements and surveillance 
requirements of Sections 3.9.1, 3.9.2 and 3.9.11 will not cause an 
accident to occur and will not result in any change in operation of 
the associated accident mitigation equipment. The design basis 
accidents (fuel handling and boron dilution event) remain the same 
postulated events described in the Millstone Unit No. 2 Final Safety
    Analysis Report (FSAR). Therefore, the proposed changes will not 
increase the probability of an accident previously evaluated.
    The proposed LCO and Applicability changes are consistent with 
the design basis accident analyses of record. This will ensure that 
the accident mitigation equipment functions and associated equipment 
are available for accident mitigation as assumed in the associated 
accident analyses. The proposed surveillance requirement changes 
will continue to provide reasonable assurance of equipment 
operability. As a result, the accident assumptions and mitigation 
methods will not be adversely affected by the changes. Therefore, 
the proposed changes will not result in [an] increase in the 
consequences of accident[s] previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the Technical Specifications do not 
impact any system or component that could cause an accident. The 
proposed changes will not alter the plant configuration (no new or 
different type of equipment will be installed) or require any new or 
unusual operator actions. The proposed changes will not alter the 
way any structure, system, or component functions, and will not 
significantly alter the manner in which the plant is operated. There 
will be no adverse effect on plant operation or accident mitigation 
equipment. The response of the plant and the operators following an 
accident will not be different. In addition, the proposed changes do 
not introduce any new failure modes. Therefore, the proposed changes 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed LCO and Applicability changes are consistent with 
the design basis accident analyses of record. The proposed 
surveillance requirement changes will continue to provide assurance 
of equipment operability. The proposed changes do not involve any 
changes in the accident analyses, therefore, the proposed changes do 
not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: April 23, 2001.
    Description of amendment request: The proposed amendment would 
remove the surveillance requirement to perform inspections of the 
Emergency Diesel Generators (EDGs) during shutdown conditions from 
Technical Specifications; although, inspections of the EDGs would 
continue to be performed in accordance with procedures prepared in 
conjunction with the recommendations of the manufacturer.
    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 analysis 
which is based on the representations made by the licensee in the April 
23, 2001 application, is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The Technical Specification change is associated with the 
surveillance requirement to perform inspections of the EDGs during 
shutdown conditions. The proposed change will remove this surveillance 
requirement from Technical Specifications; although, inspections of the 
EDGs will continue to

[[Page 31706]]

be performed in accordance with procedures prepared in conjunction with 
the recommendations of the manufacturer.
    Removal of the EDG inspection surveillance requirement from 
Technical Specifications does not verify operability or EDG functions 
assumed in the safety analysis. EDG inspections, which are maintenance 
activities that can be adequately controlled by plant procedures, will 
still be performed in accordance with the recommendations of the 
manufacturer. This will provide continued assurance the EDGs will be 
available when required.
    The proposed Technical Specification change will have no adverse 
effect on plant operation or the operation of accident mitigation 
equipment, and will not impact the availability of accident mitigation 
equipment. The plant response to the design basis accidents will not 
change. In addition, the equipment covered by this specification change 
is not an accident initiator and cannot cause an accident. Therefore, 
the proposed Technical Specification change will not result in an 
increase in the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not impact any system or component which 
could cause an accident. The proposed change will not alter the plant 
configuration (no new or different type of equipment will be installed) 
or require any unusual operator actions. The proposed change will not 
alter the way any structure, system, or component functions, and will 
not alter the manner in which the plant is operated. There will be no 
adverse effect on plant operation or accident mitigation equipment. The 
proposed change does not introduce any new failure modes. Also, the 
response of the plant and the operators following an accident will not 
be different as a result of this change. In addition, the accident 
mitigation equipment affected by the proposed change is not an accident 
initiator. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any previously 
analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed change will have no adverse effect on plant operation 
or equipment important to safety. The plant response to the design 
basis accidents will not change and the accident mitigation equipment 
will continue to function as assumed in the design basis accident 
analysis. Therefore, there will be no reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: May 2, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to not require the moderator 
temperature coefficient (MTC) determination in TS 4.1.1.4.2.c if the 
results of the MTC determinations required in TSs 4.1.1.4.2.a and 
4.1.1.4.2.b are within a certain tolerance of the corresponding design 
values.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Under the proposed change, compliance with the TS[s] is 
maintained by measuring the beginning[-]of[-]cycle [(BOC)] 
temperature coefficients.
    This change does not require a modification to any of the 
assumptions used in the input to the safety analyses. The 
assumptions were based on the current range of MTC allowed by TSs. 
The proposed change does not include a revision to the TS allowed 
range of MTC.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different kind of 
accident from any accident previously evaluated?
    This change does not result in changing plant operation or any 
TS limits. The MTC will continue to be acceptably verified within 
specified limits. As described in the Combustion Engineering topical 
report, if the BOC MTC measurements are within the specified 
tolerance when compared to the design value, then the EOC [end-of-
cycle] value is expected to fall within the design margin.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    This change does not modify the range of allowed temperature 
coefficients. The surveillance program consisting of BOC 
measurements, of plant parameter monitoring, and of explicit EOC 
predictions will ensure that the MTC remains within the range of 
acceptable values.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

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

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

    Date of amendment request: April 27, 2001.
    Description of amendment request: The proposed amendment would 
revise TS 5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' 
which requires the inspection of each reactor coolant pump (RCP) 
flywheel in general conformance with the recommendations of Regulatory 
Position C.4.b of NRC Regulatory Guide (RG) 1.14, Revision 1, ``Reactor 
Coolant Pump Flywheel Integrity,'' dated August 1975. The proposed 
change revises TS 5.5.7 to provide an exception to the recommendations 
of Regulatory Position C.4.b which would allow either a qualified in-
place ultrasonic volumetric examination (UT) over the volume from the 
inner bore of the flywheel to the circle of one-half the outer radius 
or a surface examination (i.e., magnetic particle testing (MT) and/or 
liquid penetrant testing (PT)) of exposed surfaces of the removed 
flywheel to be conducted at approximately 10-year intervals. The 
proposed change is in accordance with the NRC approved Improved 
Standard TS Generic Change Traveler TSTF-237, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 31707]]

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?
    An integral part of the Reactor Coolant System (RCS) in a 
Pressurized Water Reactor (PWR) is the Reactor Coolant Pump (RCP). 
The RCP ensures an adequate cooling flow rate by circulating large 
volumes of the primary coolant water at high temperature and 
pressure through the RCS. Following an assumed loss of power to the 
RCP motor, the flywheel, in conjunction with the impeller and motor 
assembly, provide sufficient rotational inertia to assure adequate 
core cooling flow during RCP coastdown.
    Westinghouse Electric Corporation Topical Report WCAP-14535A, 
``Topical Report on Reactor Coolant Pump Flywheel Inspection 
Elimination,'' dated November 1996, provides the technical basis for 
the elimination of inspection requirements for RCP flywheels for all 
domestic Westinghouse plants. In the Safety Evaluation for WCAP-
14535A, dated September 1996, the NRC stated that the evaluation 
methodology described in WCAP-14535A is appropriate and the criteria 
are in accordance with the design criteria of RG 1.14. RCP flywheel 
inspections have been performed for 20 years with no indications of 
service induced flaws. Flywheel integrity evaluations show a very 
high flaw tolerance for the RCP flywheels. Crack extension over a 
60-year service life is negligible. Structural reliability studies 
have shown that eliminating inspections after 10 years of plant life 
will not significantly change the probability of failure.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed change does not 
alter or prevent the ability of structures, systems, and components 
(SSC) from performing their intended function to mitigate the 
consequences of an initiating event within the acceptance limits 
assumed in the Braidwood and Byron Stations' Updated Final Safety 
Analysis Report (UFSAR). The proposed changes do not affect the 
source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated in the Braidwood and Byron Stations' 
UFSAR.
    Therefore, 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 accident from any accident previously evaluated?
    The proposed change does not modify the design or function of 
the RCP flywheels. Based upon the results of WCAP-14535A, no new 
failure mechanisms will be introduced by the revised RCP Flywheel 
Inspection Program. As presented in WCAP-14535A, detailed stress 
analysis and risk assessments have been performed that indicate that 
there would be no change in the probability of failure for RCP 
flywheels if all inspections were eliminated. Flywheel integrity 
evaluations show that RCP flywheels exhibit a very high tolerance 
for the presence of flaws.
    Therefore, the proposed change does not create the possibility 
of 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?
    There is no significant mechanism for in-service degradation of 
the flywheels since they are isolated from the primary coolant 
environment. Additionally WCAP-14535A analyses have shown there is 
no significant deformation of the flywheels even at maximum 
overspeed conditions. Likewise, the results of RCP flywheel 
inspections performed throughout the industry and at the Braidwood 
Station and the Byron Station identified no indications that would 
affect flywheel integrity.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, 
LaSalle County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 16, 2001.
    Description of amendment request: The proposed amendments would 
change the reference in Technical Specification 5.5.6, ``Pre-Stressed 
Concrete Containment Tendon Surveillance Program,'' from Regulatory 
Guide 1.35, ``Inservice Inspection of Ungrouted Tendons in Prestressed 
Concrete Containments,'' Revision 3, 1989, to a reference to Subsection 
IWL, ``Requirements of Class CC Concrete Components of Light-Water 
Cooled Power Plants,'' of Section XI, ``Inservice Inspection,'' of the 
American Society of Mechanical Engineers (ASME) Boiler and Pressure 
Vessel Code, and to delete the applicability of Surveillance 
Requirement (SR) 3.0.2 to TS Section 5.5.6. SR 3.0.2 allows the 
surveillance to be performed within 1.25 times the interval specified 
in the surveillance's frequency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes to Technical Specifications (TS) Section 
5.5.6, ``Pre-Stressed Concrete Containment Tendon Surveillance 
Program,'' change the reference in TS Section 5.5.6 from Regulatory 
Guide (RG) 1.35, ``Inservice Inspection of Ungrouted Tendons in 
Prestressed Concrete Containments,'' Revision 3, 1989, to a 
reference to Subsection IWL, ``Requirements of Class CC Concrete 
Components of Light-Water Cooled Power Plants,'' of Section XI, 
``Inservice Inspection,'' of the American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, and to 
delete the applicability of Surveillance Requirement (SR) 3.0.2 to 
TS Section 5.5.6. SR 3.0.2 allows the surveillance to be performed 
within 1.25 times the interval specified in the surveillance's 
frequency. The proposed changes do not significantly effect the 
Tendon Surveillance Program, inspection frequencies, and acceptance 
criteria which provide the requirements for the performance of the 
primary containment tendon inspections at LaSalle County Station, 
Unit 1 and Unit 2.
    The performance of a primary containment tendon inspection is 
not a precursor to any accident previously evaluated. Thus, the 
proposed changes to the performance of a primary containment tendon 
inspection do not have any effect on the probability of an accident 
previously evaluated.
    The performance of primary containment tendon inspections does 
provide assurance that the primary containment will perform as 
designed. Thus, the radiological consequences of any accident 
previously evaluated are not increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Does the change create the possibility of a new or different 
kind of an accident from any accident previously evaluated?
    The proposed changes to TS Section 5.5.6, provide assurance that 
the primary containment will perform as designed and do not 
introduce any new modes of primary containment operation of failure 
mechanisms.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    On August 8, 1996, the NRC published a final rule in the Federal 
Register (i.e., 61 Federal Register 41303) to amend 10 CFR 50.55a, 
``Codes and standards,'' to incorporate by reference Subsection IWL 
of Section XI, of the ASME B&PV Code. Subsection IWL of Section XI, 
of the ASME

[[Page 31708]]

B&PV Code, provides rules for the inservice inspection and repair of 
the reinforced concrete and post tensioning systems of Class CC 
components. LaSalle County Station, Unit 1 and Unit 2, primary 
containments are Class CC components. The amended 10 CFR 50.55a 
required incorporation of Subsection IWL of Section XI, of the ASME 
B&PV Code, into inspection programs by September 9, 2001. We have 
developed an inspection program to implement Subsection IWL of 
Section XI, of the ASME B&PV Code. The proposed TS changes support 
this program.
    The revised Tendon Surveillance Program, inspection frequencies, 
and acceptance criteria developed to implement Subsection IWL of 
Section XI, of the ASME B&PV Code, as required by 10 CFR 50.55a, 
provide acceptable requirements to perform inspections of the 
tendons in the LaSalle County Station, Unit 1 and Unit 2, primary 
containments. Thus, the proposed change to TS Section 5.5.6 will 
continue to ensure the integrity of the Unit 1 and Unit 2 primary 
containment tendons as required by the current TS.
    Thus, 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 19 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Vice President, 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: Anthony J. Mendiola.

Exelon Generation Company, LLC, PSEG Nuclear LLC, and Atlantic City 
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for Amendments: April 3, 2001.
    Description of amendment request: The proposed amendment would 
revise the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, 
technical specifications (TSs) in accordance with Technical 
Specification Task Force (TSTF) item TSTF-258, Revision 4. This TSTF 
has been previously reviewed and approved by the NRC as generically 
applicable to nuclear plants with improved standard TSs, such as PBAPS. 
The proposed amendment revises TS Section 5.0, ``Administrative 
Controls,'' to delete details of staffing requirements, eliminate 
specific details for working hour limits, clarify requirements for the 
Shift Technical Advisor position, add regulatory definitions for Senior 
Reactor Operators and Reactor Operators, revise the Radioactive 
Effluents Control Program to be consistent with the intent of Title 10 
of the Code of Federal Regulations (10 CFR) Part 20, delete periodic 
reporting requirements for main stream relief valve openings, and 
revise radiological control requirements for radiation areas to be 
consistent with those specified in 10 CFR 20.1601(c).
    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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS changes are administrative in nature and do not 
impact the operation, physical configuration, or function of plant 
equipment or systems. The changes do not impact the initiators or 
assumptions of analyzed events, nor do they impact mitigation of 
accidents or transient events. Therefore, these proposed changes do 
not increase the probability of occurrence 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 TS changes are administrative in nature and 
do not alter plant configuration, require that new equipment be 
installed, alter assumptions made about accidents previously 
evaluated, or impact the operation or function of plant equipment. 
The proposed changes do not introduce any new modes of plant 
operation or make any changes to system setpoints. Therefore, these 
proposed changes do not create the possibility of a new or different 
kind of accident than previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed TS changes are administrative in nature and do not 
involve physical changes to plant structures, systems, or components 
(SSCs), or the manner in which these SSCs are operated, maintained, 
modified, tested, or inspected. The proposed changes do not involve 
a change to any safety limits, limiting safety system settings, 
limiting conditions for operation, or design parameters for any SSC. 
The proposed changes do not impact any safety analysis assumptions 
and do not involve a change in initial conditions, system response 
times, or other parameters affecting any accident analysis. 
Therefore, these changes do no involve any reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for Licensee: Mr. Edward Cullen, Vice President and 
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, 
Kennett Square, PA 19348.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: April 4, 2001.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) Section 1.7, Definitions--
Reportable Events, and TS 6.6, Reportable Event Action, from the Davis-
Besse Nuclear Power Station Operating License, and revise TS 6.5.3, 
Technical Review and Control--Activities, and TS Bases 4.0.3, 
Applicability. These changes are being proposed to delete TS 
requirements already required by Title 10 of the Code of Federal 
Regulations Part 50 (10 CFR 50), update the TS Bases to reflect recent 
changes made to 10 CFR 50.73, revise the approval authorizations for 
procedures, plant modifications, tests and experiments, and reflect 
recent changes made to 10 CFR 50.59.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit No. 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions or assumptions are affected by the proposed changes to 
delete Technical Specification (TS) 1.7, Definitions--Reportable 
Event, and TS 6.6, Reportable Event Action, from the Davis-Besse 
Nuclear Power Station (DBNPS) Operating License; and revise TS Bases 
4.0.3, Applicability. Reportable Events are addressed by 10 CFR 
50.73 and it is not necessary for the TS to include items already 
required by federal regulation. The proposed changes to TS Bases 
4.0.3 would make these Bases consistent with the recent revision to 
10 CFR 50.73. The proposed changes to the TS Index reflect the 
deletion of TS 1.7 and TS 6.6, Reportable Event Action, and are 
administrative changes.
    The proposed changes to TS 6.5.3, Technical Review and Control--
Activities, provide for the approval of activities affecting nuclear 
safety by personnel authorized by procedure. These changes continue 
to implement the DBNPS Quality Assurance

[[Page 31709]]

Program commitments. Qualification requirements for individuals 
performing reviews of activities affecting nuclear safety are not 
affected. Accordingly, there is no increase in the probability of an 
accident.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected by the proposed changes. The proposed 
changes do not alter the source term, containment isolation, or 
allowable releases. The proposed changes, therefore, will not 
increase the radiological consequences of a previously evaluated 
accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes. The proposed changes do not alter any existing accident 
scenarios, or involve a modification or change in operation of any 
plant systems, structures, or components.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes are administrative in-nature and do not 
reduce or adversely affect the capabilities of any plant structures, 
systems or components to perform their nuclear safety functions.

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

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: May 15, 2001.
    Description of amendment requests: The proposed amendments would 
replace the current Technical Specification (TS) requirement to 
establish containment integrity within 8 hours if less than the 
specified minimum complement of A.C. or D.C. busses and equipment is 
operable in Modes 5 and 6. The proposed TS would require immediate 
suspension of operations involving core alterations, positive 
reactivity changes, and movement of irradiated fuel assemblies, and 
immediately initiate actions to restore the required busses and 
equipment to operable status, and to immediately declare the associated 
required residual heat removal loop(s) inoperable. The current Action 
requirement presents a scheduling and administrative burden during 
outages and extended shutdowns. In the addition, the proposed amendment 
would add options to the TS to allow containment penetration closure 
methods that are equivalent to those that are currently required during 
core alterations or movement of irradiated fuel in containment, and 
allow unisolation of some penetrations under administrative control. 
The additional options will allow flexibility in scheduling outage 
activities.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The determination that the criteria set forth in 10 CFR 50.92 
are met for this amendment request is indicated below.
    1. Does the change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?

Probability of Occurrence of an Accident Previously Evaluated

    The proposed changes to Action statements for T/S 3/4.8.2.2 and 
T/S 3/4.8.2.4 will eliminate current compensatory requirements that 
can only mitigate the consequences of accidents. The current 
requirements will be replaced with requirements that include 
measures to reduce the likelihood of accidents and assist in 
responding to malfunctions. The proposed requirements to immediately 
suspend operations involving core alterations, positive reactivity 
changes, and movement of irradiated fuel assemblies provide 
assurance that the applicable accidents, fuel handling and shutdown 
dilution accidents, will not occur by requiring cessation of 
activities that may cause them. The proposed requirements to 
immediately initiate actions to restore the required busses and 
equipment to operable status and to immediately declare associated 
required RHR loop(s) inoperable provide assurance operators can take 
timely corrective action for malfunctions that may lead to a 
dilution accident, and will take appropriate corrective actions for 
RHR malfunctions. Therefore, there is no adverse effect on accident 
initiators or precursors.
    The proposed change to the Applicability requirements for T/S 3/
4.8.2.2 and T/S 
3/4.8.2.4 expands the conditions under which the T/S are invoked. 
The proposed change will assure that the electrical power is 
available for mitigation of a fuel handling accident, regardless of 
the operational mode of the plant. The proposed change only involves 
accident mitigation capabilities and does not affect any accident 
initiators or precursors.
    The proposed changes to the LCO for T/S 3/4.9.4 will provide 
additional options for assuring closure of containment penetrations 
during core alterations or movement of irradiated fuel in 
containment. Containment closure provides only mitigation for the 
consequences of a fuel handling accident and does not affect the 
initiators or precursors of the accident.
    The proposed change to the Surveillance requirements for T/S 3/
4.9.4 allows the LCO to define the penetration status that is to be 
periodically verified. The effect of the proposed Surveillance 
change is bounded by the effect of the proposed LCO change as 
described above. Therefore, the proposed Surveillance change does 
not adversely affect any accident initiators or precursors.

Consequences of an Accident Previously Evaluated

    The proposed changes to the Action requirements for T/S 3/
4.8.2.2 and T/S 
3/4.8.2.4 provide assurance that fuel handling and dilution 
accidents will not occur and that timely and appropriate responses 
can and will be taken for malfunctions, thereby reducing the 
likelihood that radioactive material will be released.
    The proposed change to the Applicability requirements for Unit 1 
T/S 3/4.8.2.2 and T/S 3/4.8.2.4 provides assurance that electrical 
power is available for mitigation of a fuel handling accident (FHA), 
regardless of the operational mode of the plant. Since the current 
Applicability requirement only provides this assurance in Modes 5 
and 6, the proposed change will not increase the consequences of the 
accident.
    The additional options provided by the proposed changes to the 
LCO for T/S 3/4.9.4 will mitigate the consequences of a fuel 
handling accident in containment as effectively as those specified 
by the current LCO. Additionally, the consequences of a FHA in 
containment determined by the accident analyses will not increase 
since the analyses do not credit mitigation by closure of 
containment penetrations.
    The proposed change to the Surveillance requirements for T/S 3/
4.9.4 only reflects the change proposed for the LCO. The effect of 
the proposed Surveillance change is bounded by the effect of the 
proposed LCO change as described above. Therefore, the proposed 
Surveillance change does not adversely affect the consequences of an 
accident.
    The proposed changes to the Bases for the above identified T/S 
only provide explanatory information regarding the intent of the 
specifications and how they are to be implemented. The proposed 
Bases changes do not alter requirements of the associated T/S. 
Therefore, the effect of the Bases changes on accident initiators 
and precursors and on the consequences of an accident is bounded by 
the effect of the associated Action or LCO change as described 
above. The format changes do not alter any requirements.
    Therefore, the probability of occurrence or the consequences of 
accidents previously evaluated are not increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes to Action statements for T/S 3/4.8.2.2 and 
T/S 3/4.8.2.4 to eliminate requirements to establish containment 
integrity does not affect existing, or create new, accident 
initiators or precursors because only existing passive

[[Page 31710]]

accident mitigation features are involved. Implementation of the 
proposed new requirements to suspend operations involving core 
alterations, positive reactivity changes, and movement of irradiated 
fuel assemblies does not affect existing, or create new, accident 
initiators or precursors because these activities do not require the 
operation of existing equipment in a new or different manner, or 
involve the operation of new or different equipment. Implementation 
of the proposed new requirements to initiate actions to restore the 
required busses and equipment to operable status and to declare 
associated required RHR loop(s) inoperable does not affect or create 
new accident initiators or precursors because these activities are 
currently required by existing procedures and other T/S.
    The proposed change to the Applicability requirements for T/S 3/
4.8.2.2 and T/S 3/4.8.2.4 does not affect or create new accident 
initiators or precursors because it only expands the conditions 
under which the T/S are invoked.
    The proposed changes to the LCO for T/S 3/4.9.4 to provide 
additional options for assuring closure of containment penetrations 
during core alterations or movement of irradiated fuel in 
containment does not affect or create new accident initiators or 
precursors because the changes involve only containment penetrations 
which are passive accident mitigation measures.
    The proposed change to the Surveillance requirements for T/S 3/
4.9.4 allows the LCO to define the penetration status that is to be 
periodically verified. The effect of the proposed Surveillance 
change is bounded by the effect of the proposed LCO change as 
described above. Therefore, the proposed Surveillance change does 
not affect or create new accident initiators or precursors.
    The proposed changes to the Bases for the above identified T/S 
only provide explanatory information regarding the intent of the 
specifications and how they are to be implemented. The proposed 
Bases changes do not alter requirements of the associated T/S. 
Therefore, the effect of the Bases changes on accident initiators or 
precursors is bounded by the effect of the associated Action or LCO 
change as described above. The format changes do not alter any 
requirements.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margins of safety pertinent to the proposed changes to 
Action statements for T/S 3/4.8.2.2 and T/S 3/4.8.2.4 are those 
associated with a FHA, a shutdown dilution event, and a RHR system 
malfunction. The applicable margin of safety for a FHA is that 
defined by the off site dose analyses for the accident. Since the 
analyses do not credit mitigation by the containment, the margin of 
safety is unaffected. The applicable margin of safety for a shutdown 
dilution event is the time available for operators to take action to 
preclude violating shutdown margin requirements. The proposed new 
Action requirements to immediately suspend operations involving 
positive reactivity changes, and to immediately initiate actions to 
restore the required electrical busses and equipment to operable 
status, would not decrease the margin of safety for a shutdown 
dilution event. The applicable margin of safety for a RHR system 
malfunction is the time available for operators to take action to 
restore decay heat removal capabilities. The proposed new actions 
requirements to immediately initiate actions to restore the required 
electrical busses and equipment to operable status and to 
immediately declare associated required RHR loop(s) inoperable would 
not decrease the margin of safety for a RHR system malfunction.
    The margin of safety pertinent to the proposed changes to LCO 
for T/S 3/4.9.4 is that associated with a FHA. The applicable margin 
of safety for a FHA is that defined by the off site dose analyses 
for the accident. Since the analyses do not credit mitigation by the 
containment, the margin of safety is unaffected.
    There is no margin of safety pertinent to the proposed changes 
to associated Applicability requirements, Surveillance requirements, 
and Bases for the above identified T/S. The format changes do not 
alter any requirements.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In summary, based upon the above evaluation, [Indiana Michigan 
Power Company (I&M)] has concluded that the proposed amendment 
involves no significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, 
Buchanan, MI 49107.
    NRC Section Chief: Claudia M. Craig.

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

    Date of amendment request: February 28, 2001.
    Description of amendment request: The proposed amendment would 
change the Technical Specification (TS) to incorporate laboratory 
testing recommendations of Generic Letter 99-02, ``Laboratory Testing 
of Nuclear-Grade Activated Charcoal,'' June 3, 1999.
    The proposed charcoal testing changes and explicit reference to 
American Society for Testing and Materials (ASTM) D3803-1989 nuclear-
grade activated charcoal test protocol do not affect engineered safety 
feature (ESF) ventilation system operation or performance, reliability, 
actuation setpoints, or accident mitigation capabilities. The proposed 
changes also do not affect the operation and performance of any other 
equipment important to safety at Cooper Nuclear Station (CNS).
    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 evaluated.
    The proposed charcoal testing changes and explicit reference to 
ASTM D3803-1989 nuclear-grade activated charcoal test protocol do 
not affect ESF ventilation system operation or performance, 
reliability, actuation setpoints, or accident mitigation 
capabilities. The proposed changes also do not affect the operation 
and performance of any other equipment important to safety at CNS. 
ASTM D3803-1989 is a more accurate and demanding test which ensures 
that the charcoal filter efficiencies assumed in the CNS accident 
dose analysis are maintained. The proposed changes involve ESF 
ventilation system charcoal testing only and do not affect accident 
initiators. Therefore the proposed changes do not significantly 
increase the probability or consequences of an accident previously 
evaluated in the Updated Safety Analysis Report (USAR), as revised 
by the Design Basis Accident (DBA) radiological assessment 
calculational methodology revisions submitted to the U. S. Nuclear 
Regulatory Commission (NRC) under Reference 2.
    2. Does not create the possibility for a new or different kind 
of accident from any accident previously evaluated.
    The charcoal testing changes, and explicit reference to ASTM 
D3803-1989 nuclear-grade activated charcoal test protocol, do not 
affect ESF ventilation system operation or performance, or the 
operation and performance of any other equipment important to safety 
at CNS. The proposed changes clarify and explicitly identify the 
testing of the ESF ventilation system charcoal samples. No new or 
different accident scenarios, transient precursors, failure 
mechanisms, plant operating modes, or limiting single failures are 
introduced as a result of these changes. Therefore, the possibility 
of a new or different kind of accident from that previously 
evaluated in the USAR, as revised by the DBA radiological assessment 
calculational methodology revision submitted to the NRC under 
Reference 2, is not created by this change.
    3. Does not create a significant reduction in the margin of 
safety.
    The required performance of the ESF ventilation systems 
following a DBA is not impacted by utilizing a more demanding 
protocol for charcoal testing. Thus, the margin of safety assumed in 
the CNS accident analysis, as revised by the DBA radiological 
assessment calculational methodology revision submitted to the NRC

[[Page 31711]]

under Reference 2, is maintained. Revising the TS to clarify 
charcoal testing methodology and explicitly referencing the charcoal 
absorber testing being performed does not affect ESF ventilation 
system performance or operation, or the operation and performance of 
any other equipment important to safety at CNS. Therefore, these 
changes do not result in a significant reduction in the margin of 
safety.

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

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of amendment request: May 25, 2001.
    Description of amendment request: The proposed amendment would 
change the Kewaunee Nuclear Power Plant Technical Specification 4.2 to 
remove the steam generator tube alternate repair criteria, because 
these alternate repair criteria, as approved, are not compatible with 
the replacement steam generators scheduled to be installed in the fall 
of 2001. In addition, the proposed amendment would make administrative 
changes revising the phrasing of text without altering technical 
content.
    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.
    Changing the technical specification within limits of the 
bounding accident analyses cannot change the probability of an 
accident previously evaluated or the currently licensed radiological 
consequence predicted by the analyses of record. Removal of an 
allowance for alternate repair criteria defaults to the more 
conservative repair criteria of plugging degraded tubes. Thus, 
nothing in this proposal will cause an increase in the probability 
or consequence of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Removal of alternate repair criteria from [Technical 
Specification] TS leaves in its place the more conservative, more 
restrictive criteria for plugging degraded steam generator tubes. 
Plugging degraded steam generator tubes is a currently licensed 
repair methodology for [Kewaunee Nuclear Power Plant] KNPP, is 
consistent with current plant design bases, and does not adversely 
affect any fission product barrier, nor does it alter the safety 
function of safety significant systems, structures and components or 
their roles in accident prevention or mitigation. Currently, 
licensed design basis accident and transient analyses of record 
bound the effect of plugging tubes. Thus, this proposal does not 
create the possibility of a new or different kind of accident.
    (3) Involve a significant reduction in the margin of safety.
    The proposed change does not alter the manner in which Safety 
Limits, Limiting Safety System Setpoints, or Limiting Conditions for 
Operation are determined. It places TS 4.2 in a more conservative 
configuration than that previously approved for use by the [Nuclear 
Regulatory Commission] NRC. It conforms to plant design bases, is 
consistent with current safety analyses, and limits actual plant 
operation within analyzed and licensed boundaries. Removal of 
reference to use of alternate repair criteria from TS 4.2 and its 
Bases leaves existing and more conservative criteria in place. Thus, 
changes proposed by this request 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: Claudia M. Craig.

Nuclear Management Company, LLC, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: May 18, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to delete a redundant requirement 
for valving out a control rod drive, revise control rod accumulator 
operability requirements, add the option to hydraulically isolate 
control rod drives, and correct an inconsistency in core monitoring 
describing when source range monitors are required to be operable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Deleting the paragraph which specifies one specific pattern of 
control rod inoperability does not degrade the safe operation of the 
plant as inoperable control rods must still be analyzed to meet 
shutdown requirements.
    Revising the operability requirements for control rod 
accumulators from ``a nine-rod square array'' to: ``provided that no 
other control rod within two control rod cells in any direction has 
a:'' is a clarification. No technical requirements are changed, 
therefore, the probability or consequences of previous evaluations 
of accidents have not been affected. This change will assure 
conformance with the Banked Position Withdrawal Sequence (BPWS) 
analysis documented in General Electric (GE) report NEDO-21231. No 
changes in plant equipment will occur.
    The proposed change adds the option to hydraulically isolate the 
drive to prevent inadvertent drive withdrawal and not consider the 
accumulator inoperable. This provides a method of isolating a 
control rod drive with an inoperable accumulator in addition to 
electrical isolation when the control rod is fully inserted. A 
statement on when an inoperable accumulator is allowed is being 
relocated so that it also applies during refueling. Since in 
refueling, the plant is already shutdown, the accumulators are not 
required. As such, this change does not increase the probability or 
consequences of an accident previously evaluated.
    A qualifier is being added that source range monitors (SRMs) 
only need to be functionally tested when there are more than two 
fuel assemblies present in any reactor quadrant. Criticality is not 
considered possible with two or less fuel bundles in each quadrant 
and adjacent to an SRM. Since this change will only allow bypassing 
SRM functional checks when two fuel bundles or less are present in 
each quadrant, this change cannot result in an inadvertent 
criticality. This proposed change would reduce surveillance testing 
to that time when the instrument is required to be operable and 
provide consistency between specifications.
    The proposed Technical Specification changes do not introduce 
new equipment or new equipment operating modes, nor do the proposed 
changes alter existing system relationships. The proposed amendment 
does not introduce new failure modes. Based on the above 
justification, the proposed amendment will have no impact on the 
probability or consequences of an accident.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    This change does not degrade the safe operation of the plant as 
inoperable control rods must still be analyzed to meet existing 
shutdown reactivity requirements. It will assure conformance with 
the Banked Position Withdrawal Sequence analysis documented in 
General Electric report NEDO-21231. No changes in plant equipment 
will occur.
    Adding hydraulic isolation will not create the possibility of a 
new or different kind of

[[Page 31712]]

accident from any accident previously analyzed.
    Since this change will only allow bypassing SRM functional 
checks with two fuel bundles or less present in each quadrant, this 
change cannot result in an inadvertent criticality.
    The proposed Technical Specification changes do not introduce 
new equipment or new equipment operating modes, nor do the proposed 
changes alter existing system relationships. The proposed amendment 
does not introduce new failure modes. Based on the above 
justification, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    Revising the control rod operability requirement does not 
degrade the safe operation of the plant.
    Hydraulic isolation provides a method of isolating the drive in 
addition to the current electrical isolation. Both methods disarm 
the control rod drive and preclude the possibility of inadvertent 
drive withdrawal during subsequent operations. Adding applicability 
during refueling has little impact on safety as the drive is 
required to be fully inserted prior to isolation. As such, they do 
not involve a significant reduction in the margin of safety.
    Since this change will only allow bypassing SRM functional 
checks with two fuel bundles or less present in each quadrant, this 
change cannot result in an inadvertent criticality.
    Based on the above justification, the proposed Technical 
Specification 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: Jay E. 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 4, 2001.
    Description of amendment requests: The proposed amendments delete 
requirements from the Technical Specifications (and, as applicable, 
other elements of the licensing bases) to maintain a Post Accident 
Sampling System (PASS). Licensees were generally required to implement 
PASS upgrades as described in NUREG-0737, ``Clarification of TMI [Three 
Mile Island] Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by Order for many facilities and were added to or 
included in the technical specifications (TS) for nuclear power 
reactors currently licensed to operate. Lessons learned and 
improvements implemented over the last 20 years have shown that the 
information obtained from PASS can be readily obtained through other 
means or is of little use in the assessment and mitigation of accident 
conditions.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 11, 2000 (65 FR 49271) on possible 
amendments to eliminate PASS, including a model safety evaluation and 
model no significant hazards consideration (NSHC) determination, using 
the consolidated line item improvement process. The NRC staff 
subsequently issued a notice of availability of the models for 
referencing in license amendment applications in the Federal Register 
on October 31, 2000 (65 FR 65018). The licensee affirmed the 
applicability of the following NSHC determination in its application 
dated May 4, 2001.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident From Any Previously Evaluated

    The elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radionuclides 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact

[[Page 31713]]

to the margin of safety. Methodologies that are not reliant on PASS 
are designed to provide rapid assessment of current reactor core 
conditions and the direction of degradation while effectively 
responding to the event in order to mitigate the consequences of the 
accident. The use of a PASS is redundant and does not provide quick 
recognition of core events or rapid response to events in progress. 
The intent of the requirements established as a result of the TMI-2 
accident can be adequately met without reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    The NRC staff proposes to determine that the amendment request 
involves 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.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna 
Steam Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: August 8, 2000.
    Description of amendment request: The proposed change would add a 
new condition and associated required actions to Technical 
Specification (TS) 3.6.1.3, ``Primary Containment Isolation Valves 
(PCIVs).'' The new condition and actions address the unique containment 
isolation features of the hydrogen-oxygen (H<INF>2</INF>O<INF>2</INF>) 
analyzer penetrations. The containment isolation barriers for the 
H<INF>2</INF>O<INF>2</INF> analyzer penetrations consist of two PCIVs 
in series and a closed piping system outside primary containment. 
Editorial changes necessary to accommodate the addition of the proposed 
requirements were also proposed.
    The licensee also requested approval for a proposed exception to 
the Susquehanna Steam Electric Station Final Safety Analysis Report 
commitments regarding conformance of the design of closed systems to 
the criteria of Section 6.2.4 of NUREG-75/087, Revision 1, 1975 
(Standard Review Plan). The exception is related to the boundary valves 
between the H<INF>2</INF>O<INF>2</INF> analyzer and the post-accident 
sampling system (PASS) and is necessary to permit the use of the 
H<INF>2</INF>O<INF>2</INF> analyzer piping system outside primary 
containment as a redundant containment isolation barrier.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change adds a condition to LCO 3.6.1.3 to address 
the unique design of the H<INF>2</INF>O<INF>2</INF> analyzer 
penetration. The H<INF>2</INF>O<INF>2</INF> analyzer penetration 
isolation design requires that both PCIVs and the closed system be 
operable in order to support the single failure criteria and 
containment integrity. As part of the proposed change, an exemption 
[exception] to NUREG-75/087 guidance on closed systems for having 
all closed system boundary valves to be powered from a Class 1E 
power source is being requested. The proposed changes to Technical 
Specifications and Technical Specification Bases have no impact upon 
the safety functions of the H<INF>2</INF>O<INF>2</INF> Analyzer 
PCIVs and closed system. The safety functions of these components 
are to maintain primary containment integrity by limiting leakage 
following an accident to within that assumed in the DBA [design-
basis accident] LOCA [loss of coolant accident] Dose Analysis and to 
open to permit use of the H<INF>2</INF>O<INF>2</INF> Analyzer 
systems post accident. The H<INF>2</INF>O<INF>2</INF> Analyzer PCIVs 
and closed system will be maintained and leak rate tested in 
accordance with the Leakage Rate Test Program, thereby assuring that 
leakage from these components is maintained within the required 
limits. The design of these components is such that they meet the 
applicable design requirements with the exception of the PASS closed 
system boundary valves discussed above. However, the potential for a 
consequential failure of these valves has been evaluated and 
determined to be not credible. Thus, the proposed changes have no 
impact upon the H<INF>2</INF>O<INF>2</INF> Analyzer PCIVs and closed 
system to perform their containment isolation function. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    As discussed above, the proposed change to the Technical 
Specifications does not impact upon the safety function of the 
H<INF>2</INF>O<INF>2</INF> Analyzer PCIVs and closed system. The 
safety functions of these components are to maintain primary 
containment integrity by limiting leakage following an accident to 
within that assumed in the DBA LOCA Dose Analysis and to open to 
permit use of the H<INF>2</INF>O<INF>2</INF> Analyzer systems post 
accident. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not affect the safety function of any 
plant system or component, and does not have any impact on plant 
operation. The proposed change does not involve a significant 
reduction in the margin of safety as currently defined in the bases 
of the applicable Technical Specification section. 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: Bryan A. Snapp, Esquire, Assoc. General 
Counsel, PPL Services Corporation, Inc., 2 North Ninth St., GENTW3, 
Allentown, PA 18101-1179.
    NRC Acting Section Chief: Richard Correia.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch 
Nuclear Plant, Unit 2, Appling County, Georgia

    Date of amendment request: May 21, 2001.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to eliminate the response time 
testing requirements for the reactor protector system (RPS) signals of 
reactor high steam dome pressure and reactor vessel water level low.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Southern Nuclear Operating Company (SNC) has reviewed the 
proposed Technical Specifications changes described above and 
determined they do not involve a significant hazards consideration 
based on the following:
    1. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
purpose of the proposed changes is to eliminate response time 
testing requirements for select components in the RPS. However, 
because of the continued application of other existing Technical 
Specifications requirements, such as channel calibrations, channel 
checks, channel functional tests, and logic system functional tests, 
the response time of the RPS will be maintained within the 
acceptance limits assumed in plant safety analyses. This will assure 
successful mitigation of an initiating event. The proposed Technical 
Specifications changes do not affect the capability of the 
associated systems to perform their intended function

[[Page 31714]]

within their required response time. The BWR Owners' Group (BWROG) 
has documented an evaluation in NEDO-32291, Supplement 1, ``System 
Analyses for the Elimination of Selected Response Time Testing 
Requirements'', which was submitted to the NRC for review and 
approval as a Topical Report in December 1997. The BWROG submitted 
additional information to the staff in Addendum 1 to NEDO 32291, 
Supplement 1 in November, 1998. Subsequently, the NRC approved the 
Topical Report by a Safety Evaluation Report (SER) issued in June, 
1999.
    This evaluation demonstrates that response time testing is 
redundant to the other Technical Specifications requirements listed 
in the proceeding paragraph. These other tests are sufficient to 
identify failure modes or degradation in instrument response time 
and ensure operation of the associated systems within acceptance 
limits. Furthermore, Addendum 1 to NEDO 32291, Supplement 1 clearly 
demonstrates defense-in-depth, such that from a realistic basis, 
there is no safety significance even if instrumentation loop 
response times are significantly longer than the loop bounding 
response times.
    2. The proposed changes will not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
The proposed Technical Specifications changes do not affect the 
capability of the RPS to perform its intended function within the 
acceptance limits assumed in plant safety analyses. Periodic 
surveillance of these RPS instrument loop components will continue 
and may be used to detect degradation that could cause the response 
time characteristic to exceed the BRT [bounting response time] 
allowance.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety. The current Technical Specifications 
response times are based on the maximum allowable values assumed in 
the plant safety analyses, which conservatively establish the margin 
of safety. As described above, the proposed Technical Specifications 
changes do not affect the capability of the associated systems to 
perform their intended function within the allowed response time 
used as the basis for the plant safety analyses. Plant and system 
responses to an initiating event will remain in compliance with the 
assumptions of the safety analyses; therefore, the margin of safety 
is not affected.

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

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket No. 50-366, Edwin I. Hatch 
Nuclear Plant, Unit 2, Appling County, Georgia

    Date of amendment request: May 23, 2001.
    Description of amendment request: The proposed amendment would 
change the Safety Limit Minimum Critical Power Ratios (SLMCPR) in 
Technical Specification (TS) 2.1.1.2 to reflect the results of a cycle-
specific calculation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specification changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The derivation of the revised SLMCPRs for Plant Hatch Unit 2 
Cycle 17 for incorporation into the TS, and their use to determine 
cycle-specific thermal limits, have been performed using NRC-
approved methods and procedures. The procedures incorporate cycle-
specific parameters and reduced power distribution uncertainties in 
the determination of the value for SLMCPRs. These calculations do 
not change the method of operating the plant and have no effect on 
the probability of an accident initiating event or transient. The 
basis of the MCPR Safety Limit is to ensure no mechanistic fuel 
damage is calculated to occur if the limit is not violated. The new 
SLMCPRs preserve the existing margin to transition boiling and the 
probability of fuel damage is not increased. Therefore, the proposed 
changes do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes result only from a cycle-specific 
application of NRC-approved methods to the Unit 2 Cycle 17 core 
reload. These changes do not involve any new method for operating 
the facility and do not involve any facility modifications. No new 
initiating events or transients result from these changes. 
Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The margin of safety as defined in the TS bases will remain the 
same. Cycle-specfic SLMCPRs are calculated using NRC-approved 
methods and procedures which are in accordance with the current fuel 
design and licensing criteria. The SLMCPRs remain high enough to 
ensure that greater than 99.9% of all fuel rods in the core are 
expected to avoid transition boiling if the limit is not violated, 
thereby preserving the fuel cladding integrity. Therefore, the 
proposed TS changes do not involve a reduction in the margin of 
safety.

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

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

    Date of amendment request: April 27, 2001.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3.3.6, ``Containment Ventilation 
Isolation Instrumentation,'' to relax the slave relay test frequency 
from every 92 days to every 18 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The results of WCAP-13878 demonstrate that slave relays are 
highly reliable. WCAP-13878 also provides guidance to assure that 
slave relays remain highly reliable. The aging assessment concludes 
that the age/temperature-related degradation of all ND [normally 
deenergized] relays, and NE [normally energized] relays produced 
after 1992, is sufficiently slow such that a refueling frequency 
surveillance interval will not significantly increase the 
probability of slave relay failures. Finally, the evaluation of the 
auxiliary relays actuated during slave relay testing has concluded 
that based on the tests of the auxiliary relays performed during 
other equipment testing, reasonable assurance is provided that 
failures will be identified if the associated slave relays are 
tested on a refueling frequency.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not alter the performance of the CVI 
[containment

[[Page 31715]]

ventilation isolation] systems assumed in the plant safety analysis. 
Changing the interval for periodically verifying CVI slave relays 
(assuring equipment operability) will not create any new accident 
initiators or scenarios.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated for VEGP [Vogtle Electric Generating Plant].
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes do not affect the total CVI response 
assumed in the safety analysis since the reliability of the slave 
relays will not be significantly affected by the decreased 
surveillance frequency.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based on the above safety evaluation, VEGP concludes that the 
changes proposed by this submittal satisfy the no significant 
hazards consideration standards of 10 CFR 50.92(c) and, accordingly, 
a no significant hazards finding is justified.

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

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

    Date of amendment request: December 20, 2000.
    Description of amendment request: The proposed change will delete 
Condition 2.G, ``Reporting to the Commission,'' and Technical 
Specification 6.6.1.a, ``Reportable Event Action.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.92, it has been determined that this 
request involves no significant hazards considerations. The 
determination of no significant hazards was made by applying the 
Nuclear Regulatory Commission established standards contained in 10 
CFR 50.92. These standards assure that any changes to the operation 
of South Texas Project in accordance with this request consider the 
following:
    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This request involves administrative changes only. No actual 
plant equipment or accident analyses will be affected by the 
proposed changes. Therefore, this request does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    This request involves administrative changes only. No actual 
plant equipment or accident analyses will be affected by the 
proposed change and no failure modes not bounded by previously 
evaluated accidents will be created. Therefore, this request does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, Reactor 
Coolant System pressure boundary, and containment structure) to 
limit the level of radiation dose to the public. This request 
involves administrative changes only.
    No actual plant equipment or accident analyses will be affected 
by the proposed change. Additionally, the proposed changes will not 
relax any criteria used to establish safety limits, will not relax 
any safety systems settings, or will not relax the bases for any 
limiting conditions of operation. Therefore, these proposed changes 
will not impact the margin of safety.
    Conclusion: Based upon the analysis provided herein, the 
proposed amendments will not increase the probability or 
consequences of an accident previously evaluated, create the 
possibility of a new or different kind of accident from any accident 
previously evaluated, or involve a reduction in a margin of safety. 
Therefore, the proposed amendments meet the requirements of 10 CFR 
50.92 and do not involve a significant hazards consideration.

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

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

    Date of amendment request: February 12, 2001.
    Description of amendment request: The proposed amendment will 
revise Technical Specifications surveillance requirement 
4.4.6.2.2.e,which refers to American Society of Mechanical Engineers 
(ASME) Boiler and Pressure Vessel Code Section XI, paragraph IWV-
3427(b) as a requirement for demonstrating that each Reactor Coolant 
System Pressure Isolation Valve specified in TS Table 3.4-1 is 
operable. Part 10 of the ASME Operations and Maintenance (OM) 
Standards, OMa-1988, is currently the applicable code for these valves 
and does not have these requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.91, this analysis provides a determination 
that the proposed change to the Technical Specifications described 
previously does not involve any significant hazards consideration as 
defined in 10 CFR 50.92.
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This Technical Specification change only affects trending of 
valve leakage rate test results to anticipate the expected leakage 
rate performance of Reactor Coolant System pressure isolation 
valves. Redundant pressure isolation valves are included in the 
plant to ensure continued protection of lower pressure systems from 
exposure to the higher pressure of the Reactor Coolant System in the 
event that excessive leakage develops in an isolation valve. In 
addition, leakage rate tests of Reactor Coolant System pressure 
isolation valves will continue to be performed with no change in the 
accepted amount of leakage or frequency. Therefore, the proposed 
change does not involve a significant increase in the probability of 
an accident previously evaluated.
    The limiting event associated with these valves is a Loss of 
Coolant Accident. This has already been reviewed as part of the 
South Texas Project Updated Final Safety Analysis Report. Therefore, 
the proposed change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposed change only removes a requirement for trending of 
pressure isolation valve leakage rates. The proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There is no change in the design of the plant associated with 
this proposed license

[[Page 31716]]

amendment. The only impact of this change is in the prediction of 
when a particular pressure isolation valve may have a leakage rate 
higher than what is allowed. Adverse test results will be addressed 
under the corrective action program and by application of the 
Maintenance Rule. Engineering analysis of test results can take into 
account special circumstances associated with a test that would 
affect the conclusions.
    Leakage rate test measurements of South Texas Project Reactor 
Coolant System isolation valves will continue to be taken pursuant 
to the surveillance requirements of Technical Specification 
4.4.6.2.2, which is consistent with the requirements of code OMa-
1988, paragraph 4.2.2.3.e for analysis of leakage rates. Code OMa-
1988, paragraph 6.3, requires records of tests, including analysis 
of deviations in test values. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.

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

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

    Date of amendment request: February 28, 2001.
    Description of amendment request: The proposed amendment will 
revise the Technical Specifications (TS) to eliminate periodic response 
time testing requirements on selected sensors and selected protection 
channels, and will modify TS Section 1.0 Definitions for ``ENGINEERED 
SAFETY FEATURE (ESF) RESPONSE TIME'' and ``REACTOR TRIP SYSTEM (RTS) 
RESPONSE TIME'' to provide for verification of response time for 
selected components. Surveillances 4.3.1.2 and 4.3.2.2 will be modified 
consistent with the new definitions. The associated Bases will be 
revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.92, it has been determined that this 
request involves no significant hazards consideration. The 
determination of no significant hazards was made by applying the 
Nuclear Regulatory Commission established standards contained in 10 
CFR 50.92. These standards assure that any changes to the operation 
of South Texas Project in accordance with this request consider the 
following:
    (1) Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    This change to the Technical Specifications does not result in a 
condition where the design, material, and construction standards 
that were applicable prior to the change are altered. The same RTS 
[Reactor Trip System] and ESFAS [Engineered Safety Features 
Actuation System] instrumentation is being used; the time response 
allocations/modeling assumptions in the Chapter 15 analyses are 
still the same; only the method of verifying time response is 
changed. The proposed change will not modify any system interface 
and could not increase the likelihood of an accident since these 
events are independent of this change. The proposed activity will 
not change, degrade or prevent actions or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the SAR [Safety Analysis Report]. Therefore, 
the proposed amendment does not result in any increase in the 
probability or consequences of an accident previously evaluated.
    (2) Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Response: No.
    This change does not alter the performance of the pressure and 
differential pressure transmitters and switches, Process Protection 
racks, Nuclear Instrumentation, and Logic Systems used in the plant 
protection systems. All sensors, Process Protection racks, Nuclear 
Instrumentation, and Logic Systems will still have response time 
verified by test before placing the equipment into operational 
service and after any maintenance that could affect the response 
time. Changing the method of periodically verifying instrument 
response times for certain equipment (assuring equipment 
operability) from time response testing to calibration and channel 
checks will not create any new accident initiators or scenarios. 
Periodic surveillance of these instruments will detect significant 
degradation in the equipment response time characteristics. 
Implementation of the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Will the change involve a significant reduction in a margin 
of safety?
    Response: No.
    This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method for selected pressure and differential pressure 
sensors and for Process Protection racks, Nuclear Instrumentation, 
and Logic Systems is modified to allow use of actual test data or 
engineering data. The method of verification still provides 
assurance that the total system response time is within that assumed 
in the safety analysis. Based on the above, it is concluded that the 
proposed license amendment request does not result in a reduction in 
margin of safety.
    Conclusion: Based on the preceding analysis, it is concluded 
that elimination of periodic equipment response time testing is 
acceptable and the proposed license amendment does not involve a 
Significant Hazards Consideration as defined in 10 CFR 50.92.

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

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, 11555 Rockville Pike (first floor),

[[Page 31717]]

Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/NRC/ADAMS/index.html. If you do not have access to ADAMS or 
if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to <a href="/cdn-cgi/l/email-protection#9bebffe9dbf5e9f8b5fcf4ed"><span class="__cf_email__" data-cfemail="23534751634d51400d444c55">[email protected]</span></a>.

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

    Date of application for amendment: March 1, 2001.
    Brief description of amendment: The amendment increases the reactor 
core isolation cooling system surveillance test upper pressure limit.
    Date of issuance: May 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 139.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17964).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 31, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 6, 2000 (U-603329).
    Brief description of amendment: The amendment relocates Technical 
Specification Figure 3.6.4.1-1, ``Secondary Containment Drawdown Time 
for 1500 cfm Boundary Leakage'' to plant procedures.
    Date of issuance: June 1, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 140.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 29, 2000 (65 
FR 71132).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 1, 2001.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone 
Nuclear Power Station, Unit No. 2, New London County, Connecticut

    Date of application for amendment: December 21, 2000.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.7.11 to allow plant operation to continue if the 
temperature of the Ultimate Heat Sink (UHS) exceeds the TS limit of 75 
deg.F provided the water temperature, averaged over the previous 24-
hour period, is at or below 75  deg.F. This operational flexibility 
only applies if the UHS temperature is between 75  deg.F and 77  deg.F. 
The action time requirements if the UHS temperature exceeds 77  deg.F, 
or if the 24-hour averaged value exceeds 75  deg.F still apply. An 
associated footnote that is no longer applicable was deleted, and the 
associated TS Bases were modified to reflect these changes.
    Date of issuance: May 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 257.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 18, 2001 (66 FR 
20007).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 31, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 6, 2001, as 
supplemented by letter dated May 1, 2001.
    Brief description of amendment: The amendment revised the Technical 
Specifications associated with the reactor coolant system leakage 
detection systems, to make them consistent with the requirements in 
NUREG-1432, ``Standard Technical Specifications, Combustion Engineering 
Plants.''
    Date of issuance: May 29, 2001.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 231.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 7, 2001 (66 FR 
13803).
    The May 1, 2001, supplemental letter provided clarifying 
information that was within the scope of the original Federal Register 
notice and did not change the staff's initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 29, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: January 24, 2000.
    Brief description of amendment: Entergy Operations, Inc. requests 
revisions to the Grand Gulf Nuclear Station Technical Specifications 
which specify the minimum useable fuel oil inventories to be maintained 
in the Division 1, 2, and 3 Diesel Generator Fuel Oil Storage Tanks.
    Date of issuance: May 24, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 147.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 22, 2000 (65 FR 
15381).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 24, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: November 7, 2000 as 
supplemented by letter dated March 23, 2001.
    Brief description of amendments: The amendments would revise the 
technical specifications (TS) to extend the TS surveillance test 
interval (STI) from a 92-day STI to an 18-month STI, for the solid 
state protection system (SSPS) slave relay types that meet the 
acceptance criteria for the reliability assessments performed in 
accordance with the methodology described in the NRC approved 
Westinghouse Electric Corporation Topical Reports.
    Date of issuance: May 31, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 121, 121, 115, and 115.

[[Page 31718]]

    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 21, 2001 (66 
FR 11053).
    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 amendments is contained 
in a Safety Evaluation dated May 31, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 25, 2000.
    Brief description of amendments: Revised the Action Statements 
associated with Technical Specification (TS) Table 3.3.7.5-1,``Accident 
Monitoring Instrumentation,'' concerning the Drywell Hydrogen/Oxygen 
(H<INF>2</INF>/O<INF>2</INF>) Concentration Analyzers, and the 
associated TS Bases.
    Date of issuance: As of date of issuance and shall be implemented 
within 30 days.
    Effective date: May 24, 2001.
    Amendment Nos.: 151 and 115.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 27, 2000 (65 
FR 81929).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 24, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: August 7, 2000, as supplemented 
February 6, 2001.
    Brief description of amendment: This amendment will change 
Technical Specification (TS) Section Bases 3/4.3.1 and 3/4.3.2 to 
clarify the actions that must be performed when Steam and Feedwater 
Rupture Control System (SFRCS) components and SFRCS-actuated components 
are inoperable. Specifically, the changes will provide guidance on 
which TS actions are applicable for SFRCS-actuated components. The 
changes will also add a new TS 3/4.7.1.8 which would provide 
appropriate requirements for the Main Feedwater Control Valves and the 
Startup Feedwater Control Valves. Additionally, the changes add TS 3/
4.7.1.9 which will provide requirements for the Turbine Stop Valves. 
The changes are consistent with the intent of NUREG-1430, ``Standard 
Technical Specifications--Babcock and Wilcox Plants,'' Revision 1, 
April 1995.
    Date of issuance: May 29, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days.
    Amendment No.: 246.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 1, 2000 (65 FR 
65342).
    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 May 29, 2001.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: April 6, 2001.
    Brief description of amendment: The amendment revises the Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications (TSs) Section 6.2, 
``Organization,'' and Section 6.13, ``High Radiation Area'' to reflect 
the title change from Shift Supervisor to Shift Manager.
    Date of issuance: June 1, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 154.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 2, 2001 (66 FR 
22031).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 1, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: October 30, 2000.
    Brief description of amendments: The amendments approve the 
insertion of breakaway ceramic pins into the latches of eight double-
leaf doors in the auxiliary building special ventilation zone in order 
to restrain the doors and reduce the frequency of open-door position 
alarms. The ceramic latch pins are designed to break at forces well 
below the differential pressure that would be generated in the 
auxiliary building as a result of a postulated high-energy line break 
(HELB), and thereby allow the doors to swing open and create a relief 
path from the auxiliary building. Therefore, the modification provides 
the restraints needed to reduce the frequency of open-door position 
alarms; but without impeding the doors' steam relief function that was 
assumed in the design-basis HELB analysis.
    Date of issuance: May 30, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 157 and 148.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revise the Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15928).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 30, 2001.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: February 16, 2001 (ULNRC-04390).
    Brief description of amendment: The amendment revises Technical 
Specification 5.2.1.c to replace the title ``Vice President and Chief 
Nuclear Officer'' with ``Senior Vice President and Chief Nuclear 
Officer.''
    Date of issuance: May 30, 2001.
    Effective date: May 30, 2001, to be implemented within 60 days from 
the date of issuance.
    Amendment No.: 145.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 4, 2001 (66 FR 
17971)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 30, 2001.
    No significant hazards consideration comments received: No.

[[Page 31719]]

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

    Date of application for amendments: December 12, 2000, as 
supplemented by letters dated January 8, and February 22, 2001.
    Brief Description of amendments: The amendments revise Technical 
Specification Section 3.17 and associated Bases. The proposed changes 
will accommodate a vacuum-assisted fill technique for backfilling 
isolated reactor coolant system (RCS) loops from the active volume of 
the RCS.
    Date of issuance: May 22, 2001.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 226 and 226.
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: March 21, 2001 (66 FR 
15932).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 22, 2001.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland this 5th day of June 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-14755 Filed 6-11-01; 8:45 am]
BILLING CODE 7590-01-U


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