Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 74354-74365 [E7-25416]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 249 (Monday, December 31, 2007)]
[Notices]
[Pages 74354-74365]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-25416]


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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 section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and grants the Commission the 
authority to issue and make immediately effective any amendment to an 
operating license upon a determination by the Commission that such 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 6, 2007 to December 19, 2007. The 
last biweekly notice was published on December 18, 2007 (72 FR 71703).

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

[[Page 74355]]

within 30 days after the date of publication of this notice will be 
considered in making any final determination. Within 60 days after the 
date of publication of this notice, 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.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place 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, Rulemaking, 
Directives and Editing 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 Commission's 
Public Document Room (PDR), located at One White Flint North, Public 
File Area O1F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. The filing of requests for a hearing and petitions for leave 
to intervene is discussed below.
    Within 60 days after the date of publication of this notice, 
person(s) 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 
via electronic submission through the NRC E-Filing system 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 person(s) should consult a current copy of 
10 CFR 2.309, which is available at the Commission's PDR, located at 
One White Flint North, Public File Area 01F21, 11555 Rockville Pike 
(first floor), Rockville, Maryland. Publicly available records will be 
accessible from the Agencywide Documents Access and Management System's 
(ADAMS) Public Electronic Reading Room on the Internet at the NRC Web 
site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request 
for a hearing or petition for leave to intervene is filed within 60 
days, the Commission or a presiding officer designated by the 
Commission or by the Chief Administrative Judge of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the Chief Administrative Judge of the Atomic 
Safety and Licensing Board will issue a notice of a hearing or an 
appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor who fails to satisfy 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.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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 hearing or a petition for leave to intervene must be 
filed in accordance with the NRC E-Filing rule, which the NRC 
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process 
requires participants to submit and serve documents over the Internet 
or in some cases to mail copies on electronic storage media. 
Participants may not submit paper copies of their filings unless they 
seek a waiver in accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
five (5) days prior to the filing deadline, the petitioner/requestor 
must contact the Office of the Secretary by e-mail at 
HEARINGDOCKET@NRC.GOV, or by

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calling (301) 415-1677, to request (1) a digital ID certificate, which 
allows the participant (or its counsel or representative) to digitally 
sign documents and access the E-Submittal server for any proceeding in 
which it is participating; and/or (2) creation of an electronic docket 
for the proceeding (even in instances in which the petitioner/requestor 
(or its counsel or representative) already holds an NRC-issued digital 
ID certificate). Each petitioner/requestor will need to download the 
Workplace Forms Viewer\TM\ to access the Electronic Information 
Exchange (EIE), a component of the E-Filing system. The Workplace Forms 
Viewer\TM\ is free and is available at https://www.nrc.gov/site-help/e-
submittals/install-viewer.html. Information about applying for a 
digital ID certificate is available on NRC's public Web site at https://
www.nrc.gov/site-help/e-submittals/apply-certificates.html.
    Once a petitioner/requestor has obtained a digital ID certificate, 
had a docket created, and downloaded the EIE viewer, it can then submit 
a request for hearing or petition for leave to intervene. Submissions 
should be in Portable Document Format (PDF) in accordance with NRC 
guidance available on the NRC public Web site at https://www.nrc.gov/
site-help/e-submittals.html. A filing is considered complete at the 
time the filer submits its documents through EIE. To be timely, an 
electronic filing must be submitted to the EIE system no later than 
11:59 p.m. Eastern Time on the due date. Upon receipt of a 
transmission, the E-Filing system time-stamps the document and sends 
the submitted an e-mail notice confirming receipt of the document. The 
EIE system also distributes an e-mail notice that provides access to 
the document to the NRC Office of the General Counsel and any others 
who have advised the Office of the Secretary that they wish to 
participate in the proceeding, so that the filer need not serve the 
documents on those participants separately. Therefore, applicants and 
other participants (or their counsel or representative) must apply for 
and receive a digital ID certificate before a hearing request/petition 
to intervene is filed so that they can obtain access to the document 
via the E-Filing system.
    A person filing electronically may seek assistance through the 
``Contact Us'' link located on the NRC Web site at https://www.nrc.gov/
site-help/e-submittals.html or by calling the NRC technical help line, 
which is available between 8:30 a.m. and 4:15 p.m., Eastern Time, 
Monday through Friday. The help line number is (800) 397-4209 or 
locally, (301) 415-4737.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file a motion, in accordance 
with 10 CFR 2.302(g), with their initial paper filing requesting 
authorization to continue to submit documents in paper format. Such 
filings must be submitted by: (1) First class mail addressed to the 
Office of the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff; or (2) courier, express mail, or expedited 
delivery service to the Office of the Secretary, Sixteenth Floor, One 
White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852, 
Attention: Rulemaking and Adjudications Staff. Participants filing a 
document in this manner are responsible for serving the document on all 
other participants. Filing is considered complete by first-class mail 
as of the time of deposit in the mail, or by courier, express mail, or 
expedited delivery service upon depositing the document with the 
provider of the service.
    Non-timely requests and/or petitions and contentions 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 and/or the contentions should be 
admitted, based on a balancing of the factors specified in 10 CFR 
2.309(c)(1)(i)-(viii). To be timely, filings must be submitted no later 
than 11:59 p.m. Eastern Time on the due date.
    Documents submitted in adjudicatory proceedings will appear in 
NRC's electronic hearing docket which is available to the public at 
https://ehd.nrc.gov/EHD_Proceeding/home.asp, unless excluded pursuant 
to an order of the Commission, an Atomic Safety and Licensing Board, or 
a Presiding Officer. Participants are requested not to include personal 
privacy information, such as social security numbers, home addresses, 
or home phone numbers in their filings. With respect to copyrighted 
works, except for limited excerpts that serve the purpose of the 
adjudicatory filings and would constitute a Fair Use application, 
participants are requested not to include copyrighted materials in 
their submission.
    For further details with respect to this amendment action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: November 19, 2007.
    Description of amendment request: The proposed amendment would make 
administrative revisions to delete requirements that are obsolete or 
redundant, or correct and clarify the typing and formatting of other 
requirements. The proposed changes will not result in changes to the 
plant design or the procedural controls for the operation, 
surveillance, or maintenance of the plant.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the Proposed Changes Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated?
    No. The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed changes are administrative. The changes 
delete obsolete or redundant requirements, clarify existing 
requirements, and correct typing and formatting errors. There will 
be no resulting changes to the plant design or procedural controls. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the Proposed Changes Create the Possibility of a New or 
Different Kind of Accident From Any Previously Evaluated?
    No. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated. There 
are no physical changes being made to the plant or to the manner in 
which the plant is operated. Therefore, the changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Do the Proposed Changes Involve a Significant Reduction in 
the Margin of Safety?
    No. The proposed changes do not involve a significant reduction 
in the margin of safety. There are no physical changes being made to 
the plant or to the manner in which

[[Page 74357]]

the plant is operated. The proposed changes are administrative. The 
changes delete obsolete or redundant requirements, clarify existing 
requirements, and correct typing and formatting errors. Therefore, 
the changes do not involve a significant reduction in any margin of 
safety for HBRSEP [H.B. Robinson Steam Electric Plant], Unit No. 2.

    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: David T. Conley, Associate General Counsel 
II--Legal. Department, Progress Energy Service Company, LLC, Post 
Office Box 1551, Raleigh, North Carolina 27602.
    NRC Branch Chief: Thomas H. Boyce.

Duke Power Company LLC, Docket No. 50-369, McGuire Nuclear Station, 
Unit 1, Mecklenburg County, North Carolina.

    Date of amendment request: February 21, 2007, as supplemented 
August 9, 2007.
    Description of amendment request: The proposed amendment would 
allow, on a one-time basis, an extension of the interval governing the 
conduct of the Integrated Leak Rate Test (ILRT) for McGuire Nuclear 
Station, Unit 1. The proposed amendment would revise administrative 
Technical Specification (TS) 5.5.2, ``Containment Leak Rate Testing 
Program,'' from the currently approved 15-year interval (since the last 
McGuire Nuclear Station, Unit 1, Type A test) to a frequency 
encompassing the end of the McGuire Nuclear Station, Unit 1, End-of-
Cycle 19 refueling outage (approximately 6 months beyond the present TS 
frequency).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. 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.

First Standard

    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed extension to the Type A testing intervals 
cannot increase the probability of an accident previously evaluated 
since extension of the intervals is not a physical plant 
modification that could alter the probability of accident 
occurrence, nor is it an activity or modification by itself that 
could lead to equipment failure or accident initiation. The proposed 
extension to the Type A testing intervals does not result in a 
significant increase in the consequences of an accident as 
documented in NUREG-1493 [``Performance-Based Containment Leak-Test 
Program'', NUREG-1493, September 1995]. The NUREG notes that very 
few potential containment leakage paths are not identified by Type B 
and Type C tests. It concludes that reducing the Type A testing 
frequency to once per twenty years leads to an imperceptible 
increase in risk. McGuire [Nuclear Station, Unit 1 (McGuire Unit 1)] 
provides a high degree of assurance through testing and inspection 
that the containment will not degrade in a manner detectable only by 
Type A testing. Prior Type A tests for McGuire Unit 1 identified 
containment leakage within acceptance criteria, indicating a very 
leak tight containment. Inspections required by the ASME Code 
[American Society of Mechanical Engineers (ASME), Boiler and 
Pressure Vessel Code (Code)] are also performed in order to identify 
indications of containment degradation that could affect leak 
tightness. Separately, Type B and Type C testing, required by TS 
[Technical Specification] identify any containment opening from 
design penetrations, such as valves, that would otherwise be 
detected by a Type A test. These factors establish that an extension 
to the Type A test intervals will not represent a significant 
increase in the consequences of an accident.

Second Standard

    The proposed amendments will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed revisions to the McGuire TS add a one-time 
extension to the current interval for Type A testing. The current 
test interval of fifteen years, based on past performance, would be 
extended on a one-time basis to approximately fifteen and a half 
years from the last Type A test. The proposed extension to the Type 
A test interval does not create the possibility of a new or 
different type of accident since there are no physical changes being 
made to the plants and there are no changes to the operation of the 
plants that could introduce a new failure mode.

Third Standard

    The proposed amendment will not involve a significant reduction 
in a margin of safety. The proposed revisions to the McGuire TS add 
a one-time extension to the current interval for Type A testing. The 
current test interval of fifteen years, based on past performance, 
would be extended on a one-time basis to approximately fifteen and a 
half years from the last Type A test. The proposed extension to Type 
A test intervals will not significantly reduce the margin of safety. 
The NUREG-1493 generic study of the effects of extending containment 
leakage testing intervals found that a twenty-year interval resulted 
in an imperceptible increase in risk to the public. NUREG-1493 found 
that, generically, the design containment leakage rate contributes 
about 0.1 percent of the overall risk and that decreasing the Type A 
testing frequency would have a minimal effect on this risk, since 95 
percent of the Type A detectable leakage paths would already be 
detected by Type B and Type C testing. Similar proposed changes have 
been previously reviewed and approved by the NRC, and they are 
applicable to McGuire. Based upon the preceding discussion, Duke 
Energy Corporation [Duke Power Company, LLC] has concluded that the 
proposed amendments 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 three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Associate General 
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South 
Church Street, EC07H, Charlotte, NC 28202.
    NRC Branch Chief: Evangelos C. Marinos.

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

    Date of amendment request: November 15, 2007.
    Description of amendment request: The proposed change would 
relocate Surveillance Requirement (SR) 3.8.3.6 from the Technical 
Specifications (TS) to a licensee-controlled document. SR 3.8.3.6 
requires the Emergency Diesel Generator (EDG) Fuel Oil Storage Tanks 
(FOSTs) to be drained, sediment removed, and cleaned on a 10-year 
interval.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The FOSTs provide the storage for the EDG fuel oil, assuring an 
adequate volume is available for each EDG to operate for seven days 
in the event of a loss of offsite power concurrent with a loss of 
coolant accident. The relocation of the SR to drain and clean the 
FOSTs will not impact any of the previously analyzed accidents. 
Sediment in the tank, or failure to perform this SR, does not 
necessarily result in an inoperable storage tank. Fuel oil quantity 
and quality are assured by other TS SRs which remain unchanged. 
These SRs help ensure tank sediment is minimized and ensure that any

[[Page 74358]]

degradation of the tank wall surface that results in a fuel oil 
volume reduction is detected and corrected in a timely manner. As a 
result, adequate controls exist to allow relocation of this 
preventative maintenance cleaning requirement to licensee controlled 
documents.
    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 of accident from any accident previously evaluated?
    Response: No.
    The proposed TS changes do not involve the addition or 
modification of any plant equipment. Also, the proposed change will 
not alter the design configuration, or method of operation of plant 
equipment beyond its normal functional capabilities. The proposed TS 
change does not create any new credible failure mechanisms, 
malfunctions or accident initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not alter or exceed a design basis or 
safety limit. Diesel generator fuel oil quantity and quality will 
continue to be maintained within acceptable limits of the TS to 
assure the ability of the EDG to perform its intended function.
    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: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hilt.

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

    Date of amendment request: December 5, 2007.
    Description of amendment request: The proposed amendment would 
change the Grand Gulf Nuclear Station, Unit 1 (GGNS), Technical 
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to 
add a reference to an analytical method that will be used to determine 
core operating limits. The new reference, NEDC-33383P, ``GEXL97 
Correlation Applicable to ATRIUM-10 Fuel,'' will allow Entergy 
Operations, Inc. (Entergy) to use a Global Nuclear Fuel (GNF) method to 
determine fuel assembly critical power of AREVA ATRIUM-10 fuel. GGNS 
currently operates with a full core of ATRIUM-10 fuel. Entergy plans to 
use the GEXL97 correlation for GGNS operating Cycle 17 currently 
scheduled to begin in the fall 2008. Additionally, an administrative 
change is proposed to an existing reference in TS 5.6.5.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    Core operating limits are established each operating cycle in 
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core 
Operating Limits Report (COLR)''. These core operating limits ensure 
that the fuel design limits are not exceeded during any conditions 
of normal operation or in the event of any Anticipated Operational 
Occurrence (AOO). The methods used to determine the operating limits 
are those previously found acceptable by the NRC and listed in TS 
Section 5.6.5.b.
    A change to TS 5.6.5.b is requested to include an additional 
reference to the list of analytical methods. GGNS currently operates 
with a full core of AREVA ATRIUM-10 fuel but is scheduled to load 
GE14 fuel during the next refueling outage. GGNS plans to use the 
analysis methods of the new fuel vendor, GNF for the analysis of the 
mixed core. The GEXL97 correlation accurately models predicted core 
behavior and appropriately determines the overall critical power 
uncertainty of the method. In addition, the GEXL97 application range 
covers the range of expected operation of the ATRIUM-10 fuel during 
normal steady state and transient conditions in the GGNS reload 
cores. Although a depressurization transient could result in vessel 
pressures below the range of GEXL97, the transient would not 
threaten fuel cladding integrity, since the margin to the MCPR 
[minimum critical power ratio] safety limit increases with 
decreasing reactor pressure.
    Additionally, Entergy proposes an administrative change to the 
GESTAR-Il reference in TS 5.6.5.b. The administrative change does 
not alter any method of analysis as described in the NRC approved 
versions of GESTAR-II. The requested TS changes concern the use of 
analytical methods and do not involve any plant modifications or 
operational changes that could affect any postulated accident 
precursors or accident mitigation systems and do not introduce any 
new accident initiation mechanisms. The proposed changes have no 
effect on the type or amount of radiation released, and have no 
effect on predicted offsite doses in the event of an accident. Thus, 
the proposed change does not affect the probability of an accident 
previously evaluated nor does it increase the radiological 
consequences of any accident previously evaluated.
    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 of accident from any accident previously evaluated?
    Response: No.
    The proposed TS changes will not change the design function, 
reliability, performance, or operation of any plant systems, 
components, or structures. It does not create the possibility of a 
new failure mechanism, malfunction, or accident initiators not 
considered in the design and licensing bases. Plant operation will 
continue to be within the core operating limits that are established 
using NRC approved methods that are applicable to the GGNS design 
and the GGNS fuel.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change adds GEXL97 to the list of analytical 
methods in TS 5.6.5.b that can be used to determine core operating 
limits. Use of the GEXL97 correlation analytical method provides an 
equivalent level of protection as that currently provided. The 
administrative change does not alter any method of analysis as 
described in the NRC approved versions of GESTAR-II. The proposed 
change does not modify the safety limits or set points at which 
protective actions are initiated, and does not change the 
requirements governing operation or availability of safety equipment 
assumed to operate to preserve the margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Terence A. Burke, Associate General 
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson, 
Mississippi 39213.
    NRC Branch Chief: Thomas G. Hiltz.

[[Page 74359]]

Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania.

    Date of amendment request: November 16, 2007.
    Description of amendment request: The proposed changes revise 
technical specification (TS) action requirements associated with 
inoperable reactor coolant system (RCS) leakage detection systems. A 
new TS action requirement is proposed that will address the 
inoperability of the drywell unit cooler condensate flow rate 
monitoring system concurrent with one other RCS leakage detection 
system, other than the primary containment atmosphere gaseous 
radioactivity monitoring system. This would relax the allowed out-of-
service time for the specified combination of systems and is related to 
the current inoperability of the drywell unit cooler condensate flow 
rate monitoring system. The proposed changes would be effective for the 
remainder of the current operating cycle (Cycle 10), which is currently 
scheduled to end in the spring of 2009, or until the next shutdown of 
sufficient duration to allow for drywell unit cooler condensate flow 
rate monitoring system repairs, whichever comes first.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes continue to maintain an acceptable level of 
reactor coolant system (RCS) leakage detection instrumentation 
required to support plant operations. The level of RCS leakage 
detection capability inherent with the proposed changes will 
continue to provide acceptable early warning detection of potential 
RCS pressure boundary degradation. The proposed changes do not 
impact the physical configuration or design function of plant 
structures, systems, or components (SSCs) or the manner in which 
SSCs are operated, modified, tested, or inspected [with the 
exception of an increase in allowed out-of-service time for a 
concurrent inoperability of the drywell unit cooler condensate flow 
rate monitoring system and another specified RCS leakage detection 
system]. The proposed changes do not impact the initiators or 
assumptions of analyzed events, nor do they impact mitigation of 
accidents or transient events. Therefore, the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes only affect systems associated with the 
detection of leakage resulting from the degradation of the RCS 
pressure boundary. The proposed changes do not alter plant 
configuration or require that new plant equipment be installed. The 
RCS leakage detection systems will continue to function as designed 
in all modes of operation. No new accident type is created as a 
result of the proposed changes. No new failure mode for any 
equipment is created. The proposed changes do not alter assumptions 
made about accidents previously evaluated. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes do not involve any physical changes to 
plant SSCs or the manner in which SSCs are operated, modified, 
tested, or inspected. The proposed changes do not involve a change 
to any safety limits, limiting safety system settings, limiting 
conditions of 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 an accident analysis. 
Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, with changes as noted above, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: J. Bradley Fewell, Esquire, Associate 
General Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

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

    Date of amendment request: October 9, 2007.
    Description of amendment request: A change is proposed to the 
technical specifications (TS) of Quad Cities Nuclear Power Station 
(QCNPS), Units 1 and 2, consistent with Technical Specifications Task 
Force (TSTF) Change Traveler TSTF-423 to the standard TSs for boiling 
water reactor plants, to allow, for some systems, entry into hot 
shutdown rather than cold shutdown to repair equipment, if risk is 
assessed and managed consistent with the program in place for complying 
with the requirements of Title 10 of the Code of Federal Regulations 
(10 CFR) Section 50.65(a)(4). Changes proposed herein will be made to 
the QCNPS, Units 1 and 2, TSs for selected required action end states 
providing this allowance.
    The licensee reviewed the proposed no significant hazards 
consideration (NSHC) determination published in the Federal Register on 
March 23, 2007 (71 FR 14726) and concluded that it is applicable to 
QCNPS, Units 1 and 2. The licensee incorporated the proposed 
determination by reference to satisfy the requirements of 10 CFR 
50.91(a).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC 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 allows a change to certain required end 
states when the TS Completion Times for remaining in power operation 
will be exceeded. Most of the requested technical specification (TS) 
changes are to permit an end state of hot shutdown (Mode 3) rather 
than an end state of cold shutdown (Mode 4) contained in the current 
TS. The request was limited to: (1) Those end states where entry 
into the shutdown mode is for a short interval, (2) entry is 
initiated by inoperability of a single train of equipment or a 
restriction on a plant operational parameter, unless otherwise 
stated in the applicable technical specification, and (3) the 
primary purpose is to correct the initiating condition and return to 
power operation as soon as is practical. Risk insights from both the 
qualitative and quantitative risk assessments were used in specific 
TS assessments. Such assessments are documented in Section 6 of GE 
NEDC-32988, Revision 2, ``Technical Justification to Support Risk 
Informed Modification to Selected Required Action End States for BWR 
Plants.'' They provide an integrated discussion of deterministic and 
probabilistic issues, focusing on specific technical specifications, 
which are used to support the proposed TS end state and associated 
restrictions. The staff finds that the risk insights support the 
conclusions of the specific TS assessments. Therefore, the 
probability of an accident previously evaluated is not significantly 
increased, if at all. The consequences of an accident after adopting 
proposed TSTF-423, are no different than the consequences of an 
accident prior to adopting TSTF-423. Therefore, the consequences of 
an accident previously evaluated are not significantly affected by 
this change. The addition of a requirement to assess and manage the 
risk introduced by this change will further minimize possible 
concerns. Therefore, this

[[Page 74360]]

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 Previously Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
If risk is assessed and managed, allowing a change to certain 
required end states when the TS Completion Times for remaining in 
power operation are exceeded, i.e., entry into hot shutdown rather 
than cold shutdown to repair equipment, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change and the commitment by the licensee to adhere to the guidance 
in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0, 
``Technical Specifications End States, NEDC-32988-A,'' will further 
minimize possible concerns. Thus, this change does not create the 
possibility of a new or different kind of accident from an accident 
previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety.
    The proposed change allows, for some systems, entry into hot 
shutdown rather than cold shutdown to repair equipment, if risk is 
assessed and managed. The BWROG's [Boiling Water Reactor Owners 
Group's] risk assessment approach is comprehensive and follows staff 
guidance as documented in RGs [Regulatory Guides] 1.174 and 1.177. 
In addition, the analyses show that the criteria of the three-tiered 
approach for allowing TS changes are met. The risk impact of the 
proposed TS changes was assessed following the three-tiered approach 
recommended in RG 1.177. A risk assessment was performed to justify 
the proposed TS changes. The net change to the margin of safety is 
insignificant. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    Therefore, the NRC staff proposes to determine that the requested 
amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

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

    Date of amendment request: November 29, 2007.
    Brief description of amendments: Revision to Technical 
Specification (TS) 3.6.7, (``Spray Additive System,'' to allow 
modifications to the facility potentially required to comply with U.S. 
Nuclear Regulatory Commission (NRC) Generic Letter 2004-02, ``Potential 
Impact of Debris Blockage on Emergency Recirculation during Design 
Basis Accident at Pressurized Water Reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No
    The proposed change[s] [do] not impact the initiation or 
probability of occurrence of any accident.
    The accidents evaluated in the Final Safety Analysis report 
(FSAR) that could be affected by this proposed change are those 
involving the pressurization of the containment and those involving 
recirculation of fluid within the Emergency Core Cooling System 
(ECCS) or the Containment Spray System (e.g., loss of coolant 
accidents (LOCAs)).
    The change to a minimum pH [potential of Hydrogen] of 7.1 will 
not result in a significant increase in the radiological 
consequences of a LOCA as-described below.
    The equilibrium spray pH during the recirculation phase 
resulting from this change will be greater than or equal to 7.1. The 
pH range for the spray will be bounded by the water spray solution 
which is borated water with a maximum of 2600 parts per million 
(ppm) boron buffered to a final spray solution pH much less than the 
10.5 as described in the current FSAR Section 3.11(B) for the 
postulated spray solution environment. The maximum pH is the 
limiting parameter for equipment qualification. Since the resulting 
pH level will be closer to neutral using the lower limit of 7.1, 
post-LOCA corrosion of containment components will not be increased. 
Post-LOCA hydrogen generation will be reduced. There will not be an 
adverse radiation dose effect on any safety-related equipment. Thus, 
the potential for failures of the ECCS or safety-related equipment 
following a LOCA will not be increased as a result of the proposed 
change.
    This modification affects the Containment Spray System which is 
intended to respond to and mitigate the effects of a LOCA. The 
Containment Spray System will continue to function in a manner 
consistent with the plant design basis. There will be no degradation 
in the performance of nor an increase in the number of challenges to 
equipment assumed to function during an accident situation.
    Therefore, these Technical Specification (TS) revisions do not 
affect the probability of any event initiators. There will be no 
adverse changes to normal plant operating parameters, Engineered 
Safety Features (ESF) actuation setpoints, or accident mitigation 
capabilities.
    The proposed change allows the Spray Additive System currently 
used to mitigate the consequences of an accident to maintain the 
equilibrium sump pH at greater than or equal to 7.1 to minimize 
chloride-induced stress corrosion cracking in austenitic stainless 
components important to safety located inside containment. 
Therefore, the proposed changes will not increase the probability of 
an accident or malfunction of equipment important to safety 
previously evaluated in the FSAR.
    The offsite and control room doses will continue to meet the 
requirements of [Title 10 of the Code of Federal Regulations (10 
CFR) part 100] 10 CFR 100, 10 CFR 50 Appendix A [General Design 
Criterion] GDC 19, [Standard Review Plan] SRP 15.6.5.11, and SRP 
6.4.11. The proposed new pH limit will provide satisfactory 
retention of iodine in the sump water, as well as provide adequate 
pH control to minimize the potential of chloride-induced stress 
corrosion cracking of austenitic stainless steel components.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the revised Surveillance for the 
Containment Spray Additive System provides for a required minimum 
equilibrium pH in containment post accident. There are no electrical 
or mechanical components being added whose failure could prevent the 
system from functioning.
    No new accident scenarios, transient precursors, or limiting 
single failures are introduced as a result of the proposed changes. 
There will be no adverse effect or challenges imposed on any safety-
related system as a result of this proposed change. The amount of 
sodium hydroxide (NaOH) will provide a minimum equilibrium sump pH 
of 7.1 following mixing. Therefore, the possibility of a new or 
different type of accident is not created.
    There are no changes which would cause the malfunction of 
safety-related equipment, assumed to be operable in the accident 
analyses, as a result of the proposed Technical Specification 
changes. The possibility of a malfunction of safety-related 
equipment with a different result is not created.
    Therefore, the proposed change[s] [do] not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No
    The only function of the chemical additive system is to provide 
pH control of the post-accident containment recirculation sump 
water, since the borated water from the Refueling Water Storage Tank 
(RWST) used as the containment spray pump suction source during 
injection is sufficient to remove iodine from the containment 
atmosphere following a LOCA. The net effect on the pH control 
function of reducing the

[[Page 74361]]

amount of buffer is that the equilibrium sump pH will be lowered to 
a minimum of 7.1. There will be no change to the current Technical 
Specification acceptance limits on RWST volume and boron 
concentration. The resulting equilibrium sump pH level from this 
change will be closer to neutral; therefore, the post-LOCA corrosion 
of containment components will not be increased (i.e., would be 
reduced).
    Because the long term pH will be maintained greater than or 
equal to 7.1, margin to minimize the potential for stress corrosion 
cracking is maintained.
    The radiological analysis, as discussed in the technical 
analysis above, is shown not to be impacted. There will be no change 
to the [departure from nucleate boiling ratio] DNBR Correlation 
Limit, the design DNBR limits, or the safety analysis DNBR limits 
discussed in Bases Section 2.1.1. There will be no effect on the 
manner in which Safety Limits or Limiting Safety System Settings are 
determined nor will there be any effect on those plant systems 
necessary to assure the accomplishment of protection functions. 
There will be no adverse impact on Departure of Nucleate Boiling 
Ratio limits, [heat flux hot channel factor] FQ, [nuclear 
enthalpy rise hot channel factor] F-delta-H, LOCA peak cladding 
temperature, peak local power density, or any other margin of 
safety.
    Therefore the proposed change[s] [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: Timothy P. Matthews, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, NW., Washington, DC 20036.
    NRC Branch Chief: Thomas G. Hiltz.

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: October 2, 2007.
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 3.5.4, ``Refueling Water Storage 
Tank (RWST),'' Surveillance Requirement (SR) 3.5.4.2, to increase the 
minimum required borated water volume from ``>= [greater than or equal 
to] 400,000 gallons (81.5% indicated level)'' to ``>= 455,300 gallons 
(93.6% level),'' to reflect the new sump design required to comply with 
U.S. Nuclear Regulatory Commission (NRC) Generic Letter 2004-02, 
``Potential Impact of Debris Blockage on Emergency Recirculation during 
Design-Basis Accident at Pressurized Water Reactors.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. [Do] the proposed change[s] involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change[s] [revise] the minimum RWST borated water 
volume. The RWST borated water volume is not an initiator of any 
accident previously evaluated. As a result, the probability of an 
accident previously evaluated is not affected. The proposed 
change[s] [do] not alter or prevent the ability of structures, 
systems, and components from performing their intended function to 
mitigate the consequences of an initiating event within the assumed 
acceptance limits. The effect on containment flood level, equipment 
qualification, and containment sump pH remain within the limits 
assumed in the design and accident analyses. The proposed change[s] 
[do] not affect the source term, containment isolation, or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated. Further, the 
proposed change[s] [do] not increase the types or amounts of 
radioactive effluent that may be released offsite, nor significantly 
increase individual or cumulative occupational/public radiation 
exposures. The proposed change[s] are consistent with the safety 
analysis assumptions and resultant consequences.
    Therefore, the proposed change[s] [do] not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. [Do] the proposed change[s] create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The change[s] [do] not involve a physical alteration of the 
plant (i.e., no new or different components or physical changes are 
involved with this change) or a change in the methods governing 
normal plant operation. The change[s] [do] not alter any assumptions 
made in the safety analysis.
    Therefore, the proposed change[s] will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change[s] to revise the required RWST minimum 
borated water volume [do] not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by [these] change[s]. The proposed change[s] will 
not result in plant operation in a configuration outside of the 
design basis.
    Therefore, the proposed change[s] [do] not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) 
are satisfied. Therefore, the NRC staff proposes to determine that 
the amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: Thomas G. Hiltz.

Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power 
Plant (HBPP), Unit 3 Humboldt County, California.

    Date of amendment request: November 5, 2007.
    Description of amendment request: The licensee has proposed 
amending the technical specifications (TS) to delete many 
operational and administrative requirements upon transfer of spent 
nuclear fuel assemblies and fuel fragment containers from the Spent 
Fuel Pool (SFP) to the Humboldt Bay Independent Spent Fuel Storage 
Installation (ISFSI). Some TS requirements will be relocated to the 
HBPP Quality Assurance Plan.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes reflect the transfer of spent fuel from the 
Spent Fuel Pool to the Humboldt Bay (HB) Independent Spent Fuel 
Storage Installation. Design basis accidents related to the SFP are 
discussed in the Humboldt Bay Power Plant Unit 3 Defueled Safety 
Analysis Report (DSAR). These postulated accidents are predicated on 
spent fuel being stored in the SFP. With the removal of the spent 
fuel from the SFP, there are no important-to-safety systems, 
structures or components required to function or to be monitored. In 
addition, there are no remaining credible accidents involving spent 
fuel or the SFP that require actions of a Certified Fuel Handler or 
Noncertified Fuel Handler to prevent occurrence or to mitigate 
consequences. The proposed change to the Design Features section of 
the Technical Specifications (TS) clarifies that the spent fuel is 
being stored in dry casks within an ISFSI. The probability or 
consequences of accidents at the ISFSI are evaluated in the HB ISFSI 
Final Safety Analysis Report (FSAR) and are independent of the 
accidents evaluated in the HBPP Unit 3 DSAR. Therefore, the proposed 
changes will 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 evaluated?
    Response: No.
    The proposed changes reflect the reduced operational risks as a 
result of the spent fuel being transferred to dry casks within an 
ISFSI. The proposed changes do not modify any systems, structures or 
components. The

[[Page 74362]]

plant conditions for which the HBPP Unit 3 DSAR design basis 
accidents relating to spent fuel and the SFP have been evaluated are 
no longer applicable. The aforementioned proposed changes do not 
affect any of the parameters or conditions that could contribute to 
the initiation of an accident. Design basis accidents associated 
with the dry cask storage of spent fuel are already considered in 
the HB ISFSI FSAR. No new accident scenarios are created as a result 
of deleting nonapplicable operational and administrative 
requirements. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from those 
previously evaluated.
    (3) Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed changes reflect the reduced operational risks as a 
result of the spent fuel being transferred to dry casks within an 
ISFSI. The design basis and accident assumptions within the HBPP 
Unit 3 DSAR and the TS relating to spent fuel are no longer 
applicable. The proposed changes do not affect remaining plant 
operations, nor structures, systems, or components supporting 
decommissioning activities. In addition, the proposed changes do not 
result in a change in initial conditions, system response time, or 
in any other parameter affecting the course of a decommissioning 
activity accident analysis. Therefore, the proposed changes will not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Jennifer K. Post, Pacific Gas and 
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
    NRC Branch Chief: Andrew Persinko.

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

    Date of amendment request: October 31, 2007.
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3.8.1, ``Essential Service Water System 
(ESW),'' and TS 3.8.1, ``AC [Alternating Current] Sources--Operating.'' 
A note would be added to Condition A, one ESW train inoperable, of TS 
3.8.1, and Condition B, one diesel generator (DG) inoperable, of TS 
3.8.1 would be revised. The revisions are to allow a one-time 
completion time extension from 72 hours to 14 days to support a planned 
replacement of ESW piping prior to December 31, 2008, in the licensee's 
fall 2008 refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    [The only change to the plant is that existing ESW piping will 
be replaced in the fall 2008 refueling outage. There are no other 
changes to the plant and no hardware or equipment will be added to 
the plant. This replacement is to address localized degradation of 
the ESW piping due to microbiologically induced corrosion.]
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed to the protection systems. The same 
reactor trip system (RTS) and engineered safety feature actuation 
system (ESFAS) instrumentation will continue to be used. The 
protection systems will continue to function in a manner consistent 
with the plant design basis. The use of polyethylene (PE) piping 
[(i.e., replacing existing ESW piping by PE piping)] in the ESW 
system in accordance with ASME [American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code] Code Case N-755, with 
justified materials and design exceptions as noted in [the 
licensee's letter dated August 30, 2007 (ULNRC-05434), which 
requested relief from the ASME Code to replace the ESW piping by the 
PE piping], will [have the PE piping that replaces the ESW piping] 
provide an acceptable level of quality and safety. There will be no 
changes to the essential service water (ESW) system or [the] 
ultimate heat sink (UHS) surveillance and operating limits. [The 
licensee's letter dated August 30, 2007,] demonstrates the 
acceptability of using the PE piping in this buried ASME Class 3 
application [(i.e., replacing existing ESW piping)].
    The proposed changes will not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configurations of the facility or the manner in 
which the plant is operated and maintained. The proposed changes 
will not alter or prevent the ability of structures, systems, and 
components (SSCs) from performing their intended [safety] functions 
to mitigate the consequences of an initiating event within the 
assumed acceptance limits.
    The proposed changes do not affect the way