Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving no Significant Hazards Considerations, 41780-41793 [E7-14350]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 146 (Tuesday, July 31, 2007)]
[Notices]
[Pages 41780-41793]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-14350]


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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 July 4, 2007 to July 18, 2007. The last 
biweekly notice was published on July 17, 2007 (72 FR 39081).

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)

[[Page 41781]]

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. 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, 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 2.309, 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 a hearing or a petition for leave to intervene must 
be filed 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; (2) courier, express

[[Page 41782]]

mail, and expedited delivery services: Office of the Secretary, 
Sixteenth Floor, One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications 
Staff; (3) E-mail addressed to the Office of the Secretary, U.S. 
Nuclear Regulatory Commission, HearingDocket@nrc.gov; or (4) facsimile 
transmission addressed to the Office of the Secretary, U.S. Nuclear 
Regulatory Commission, Washington, DC, Attention: Rulemakings and 
Adjudications Staff at (301) 415-1101, verification number is (301) 
415-1966. A copy of the request for hearing and petition for leave to 
intervene should also be sent to the Office of the General Counsel, 
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and it 
is requested that copies be transmitted either by means of facsimile 
transmission to (301) 415-3725 or by e-mail to OGCMailCenter@nrc.gov. A 
copy of the request for hearing and petition for leave to intervene 
should also be sent to the attorney for the licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this 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, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North 
Carolina

    Date of amendment request: June 15, 2007.
    Description of amendment request: The amendment proposes to 
relocate the inservice testing requirements to the administrative 
section of the technical specifications (TS), remove the inservice 
inspection activities from TS and locate them in an owner-controlled 
program, and establish a TS Bases Control Program. All of these changes 
are proposed to be consistent with NUREG-1431, Revision 3, ``Standard 
Technical Specifications Westinghouse Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment does not involve a significant increase in 
the probability or consequences of an accident previously evaluated, 
and it does not change an accident previously evaluated in the Final 
Safety Analysis Report (FSAR). The proposed change is administrative in 
nature, and it will continue to ensure that the inspection and testing 
requirements required by regulations are met. The American Society of 
Mechanical Engineers (ASME) Code requirements are established, reviewed 
and approved by ASME, the industry, and ultimately endorsed by the NRC 
for inclusion into 10 CFR 50.55a. Updates to the ASME Code reflect 
advances in technology and consider information obtained from plant 
operating experience to provide enhanced inspection and testing. Thus, 
the proposed change will revise TS to appropriately reference the ASME 
Code required by 10 CFR 50.55a for performing inservice testing, 
specifically referencing the ASME Code for Operation and Maintenance of 
Nuclear Power Plants, rather than the ASME Section XI Code.
    The proposed change does not affect operations, and the inspection 
and testing required is not an accident initiator.
    Therefore, this amendment 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 of 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated, and 
it does not change an accident previously evaluated in the Final Safety 
Analysis Report (FSAR). As noted above, the proposed change is 
administrative in nature, the inspection and testing required is not an 
accident initiator, and no new accident precursors are being 
introduced. The proposed change will revise TS to appropriately 
reference the ASME Code required by 10 CFR 50.55a for performing 
inservice testing, which will continue to ensure that the inspection 
and testing requirements required by regulations are met. Since 
inservice testing will continue to be performed in accordance with 
regulations, adequate assurance is provided to ensure that the safety-
related pumps and valves will continue to operate as required. No new 
testing is required that could create a new or different type of 
accident.
    Therefore, this amendment 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?
    Response: No.
    The proposed amendment does not involve a significant reduction in 
a margin of safety. The proposed amendment does not adversely affect a 
plant safety limit or a limiting safety system setting, and does not 
alter a design basis limit for a parameter evaluated in the FSAR. The 
proposed change is administrative in nature, and it will continue to 
ensure that the inspection and testing requirements required by 
regulations are met. Since inservice testing will continue to be 
performed in accordance with regulations, adequate assurance is 
provided to ensure that the safety-related pumps and valves will 
continue to operate as required and perform their intended safety 
function.
    Therefore, this amendment does not involve a significant reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: 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.

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

    Date of amendment request: July 3, 2007.
    Description of amendment request: The proposed change relocates the 
quality and quantity requirements

[[Page 41783]]

associated with the emergency diesel generator (EDG) fuel oil within 
the Technical Specifications (TS) through the creation of a new TS 
Limiting Condition for Operation and the Diesel Fuel Oil Testing 
Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes in the diesel fuel oil testing program will 
continue to ensure that new and stored diesel fuel oil properties are 
maintained within specified limits to assure EDG operation. The testing 
of diesel generator fuel oil is not considered an initiator or a 
mitigating factor in any previously evaluated accidents.
    The deletion of the requirement to drain and inspect the fuel oil 
storage tank (FOST) does not impact any of the previously analyzed 
accidents. Periodic testing of the fuel oil as required by the Diesel 
Fuel Oil Testing Program will identify poor quality oil. Actions are 
included that will require the quality of the oil to be maintained 
within acceptable limits. Draining and inspecting the FOST are not 
considered an accident initiator or mitigating factor in any previously 
evaluated accidents.
    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 change results in changes to the existing diesel fuel 
oil testing program and the deletion of the [Surveillance Requirements] 
associated with the performance of periodic draining and inspection of 
the FOSTs. No plant modifications are required to support the proposed 
TS changes. There is no impact to plant structures, systems, or 
components, or in the design of the plant structures, systems, or 
components.
    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 result in any plant modifications. 
Diesel generator fuel oil quantity and quality will continue to be 
maintained within acceptable limits 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. Hiltz.

Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear 
Power Station (DNPS), Unit 2, Grundy County, Illinois

    Date of amendment request: July 10, 2007.
    Description of amendment request: The proposed amendment would 
revise the values of the safety limit minimum critical power ratio 
(SLMCPR) in Technical Specification (TS) Section 2.1.1, ``Reactor Core 
SLs.''
    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 probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the operability 
of plant systems designed to mitigate those consequences. Limits have 
been established consistent with NRC-approved methods to ensure that 
fuel performance during normal, transient, and accident conditions is 
acceptable. The proposed change conservatively establishes the SLMCPR 
for DNPS, Unit 2, Cycle 21 such that the fuel is protected during 
normal operation and during plant transients or anticipated operational 
occurrences (AOOs).
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of an 
accident are affected.
    The proposed change revises the SLMCPR to protect the fuel during 
normal operation as well as during plant transients or AOOs. 
Operational limits will be established based on the proposed SLMCPR to 
ensure that the SLMCPR is not violated. This will ensure that the fuel 
design safety criterion (i.e., that at least 99.9% of the fuel rods do 
not experience transition boiling during normal operation and AOOs) is 
met. Since the proposed change does not affect operability of plant 
systems designed to mitigate any consequences of accidents, the 
consequences of an accident previously evaluated are not expected to 
increase.
    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
    Creation of the possibility of a new or different kind of accident 
requires creating one or more new accident precursors. New accident 
precursors may be created by modifications of plant configuration, 
including changes in allowable modes of operation. The proposed change 
does not involve any plant configuration modifications or changes to 
allowable modes of operation.
    The proposed change to the SLMCPR assures that safety criteria are 
maintained for DNPS, Unit 2, Cycle 21.
    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 SLMCPR provides a margin of safety by ensuring that at least 
99.9% of the fuel rods do not experience transition boiling during 
normal operation and AOOs if the MCPR limit is not violated. The 
proposed change will ensure the current level of fuel protection is 
maintained by continuing to ensure that at least 99.9% of the fuel rods 
do not experience transition boiling during normal operation and AOOs 
if the MCPR limit is not violated. The proposed SLMCPR values were 
developed using NRC-approved methods. Additionally, operational

[[Page 41784]]

limits will be established based on the proposed SLMCPR to ensure that 
the SLMCPR is not violated. This will ensure that the fuel design 
safety criterion (i.e., that no more than 0.1% of the rods are expected 
to be in boiling transition if the MCPR limit is not violated) is met.
    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. Bradley J. Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Russell Gibbs.

Exelon Generation Company, LLC, Docket No. 50-373, LaSalle County 
Station, Unit 1, LaSalle County, Illinois

    Date of amendment request: June 18, 2007.
    Description of amendment request: The proposed amendment would 
revise technical specification TS 5.5.13, ``Primary Containment Leakage 
Rate Testing Program,'' to reflect a one-time extension of the LaSalle 
County Station (LSCS), Unit 1, primary containment Type A Integrated 
Leak Rate Test (ILRT) date for the current requirement of no later than 
June 13, 2009, prior to startup following the thirteenth LSCS Unit 1 
refueling outage (L1R13).
    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 will revise LSCS, Unit 1, TS 5.5.13, ``Primary 
Containment Leakage Rate Testing Program,'' to reflect a one-time 
extension of the primary containment Type A Integrated Leak Rate Test 
(ILRT) date to ``prior to startup following L1R13.'' The current Type A 
ILRT interval of 15 years, based on past performance, would be extended 
on a one-time basis by approximately 5% of the current interval.
    The function of the primary containment is to isolate and contain 
fission products released from the reactor Primary Coolant System (PCS) 
following a design basis Loss of Coolant Accident (LOCA) and to confine 
the postulated release of radioactive material to within limits. The 
test interval associated with Type A ILRTs is not a precursor of any 
accident previously evaluated. Type A ILRTs provide assurance that the 
LSCS Unit 1 primary containment will not exceed allowable leakage rate 
values specified in the TS and will continue to perform their design 
function following an accident. The risk assessment of the proposed 
changes has concluded that there is an insignificant increase in total 
population dose rate and an insignificant increase in the conditional 
containment failure probability.
    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?
    Response: No
    The proposed changes for a one-time extension of the Type A ILRT 
for LSCS Unit 1 will not affect the control parameters governing unit 
operation or the response of plant equipment to transient and accident 
conditions. The proposed changes do not introduce any new equipment, 
modes of system operation or failure mechanisms.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the change involve a significant reduction in a margin of 
safety?
    Response: No
    LSCS Unit 1 is a General Electric BWR/5 plant with a Mark II 
primary containment. The Mark II primary containment consists of two 
compartments, the drywell and the suppression chamber. The drywell has 
the shape of a truncated cone, and is located above the cylindrically 
shaped suppression chamber. The drywell floor separates the drywell and 
the suppression chamber. The primary containment is penetrated by 
access, piping and electrical penetrations.
    The integrity of the primary containment penetrations and isolation 
valves is verified through Type B and Type C local leak rate tests 
(LLRTs) and the overall leak tight integrity of the primary containment 
is verified by a Type A ILRT, as required by 10 CFR 50, Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors.'' These tests are performed to verify the essentially leak 
tight characteristics of the primary containment at the design basis 
accident pressure. The proposed changes for a one-time extension of the 
Type A ILRT does not affect the method for Type A, B, or C testing or 
the test acceptance criteria.
    EGC has conducted a risk assessment to determine the impact of a 
change to the LSCS Unit 1 Type A ILRT schedule from a baseline ILRT 
frequency of three times in ten years to once in 15.67 years (i.e., 15 
years plus 8 months) for the risk measures of Large Early Release 
Frequency (i.e., LERF), Total Population Dose, and Conditional 
Containment Failure Probability (i.e., CCFP). This assessment indicated 
that the proposed LSCS ILRT interval extension has a minimal impact on 
public risk.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
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.

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

    Date of amendment request: November 27, 2006.
    Description of amendment request: The proposed amendments would 
modify various technical specification (TS) requirements for emergency 
diesel generators (EDGs). Specifically, the licensee stated that the 
proposed changes would eliminate several accelerated tests and a test 
table, modify acceptance criteria for fast start and load rejection 
tests, and also, eliminate the EDG failure report. The proposed changes 
are consistent with the Nuclear Regulatory Commission's (NRC's) 
regulatory guidance presented in Generic Letter 93-05, ``Line-Item 
Technical Specifications Improvement to Reduce Surveillance 
Requirements for Testing During Power Operation,'' Generic Letter 94-
01, ``Removal of Accelerated Testing and Special Reporting Requirements 
for Emergency Diesel Generators,'' and NUREG-1433, Rev. 3.1, ``Standard 
Technical Specifications, General Electric Plants, BWR/4.''

[[Page 41785]]

    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 are associated with the testing and reporting 
requirements of the eight (four on each unit) Emergency Diesel 
Generators (EDGs). The changes will eliminate unnecessary EDG testing 
requirements that contribute to potential mechanical degradation of the 
EDGs. The changes are based on the NRC guidance and recommendations 
provided in Generic Letter 93-05 or Generic Letter 94-01, or are 
consistent with NUREG-1433. The change to the reporting requirement is 
administrative in nature.
    The probability of an accident is not increased by these changes 
because the EDGs are not assumed to be initiators of any design basis 
event. Additionally, the proposed changes do not involve any physical 
changes to plant systems, structures, or components (SSC), or the 
manner in which these SSC are operated, maintained, or controlled. The 
consequences of an accident will not be increased because the changes 
to the EDGs and associated support systems still provide a high degree 
of assurance that their operability is maintained.
    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 do not alter the physical design, safety 
limits, or safety analysis assumptions, associated with the operation 
of the plant. Accordingly, the proposed changes do not introduce any 
new accident initiators, nor do they reduce or adversely affect the 
capabilities of any plant structure or system in the performance of 
their safety function.
    Therefore, the proposed changes 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 proposed changes to the EDGs either: (1) Modify the test 
acceptance criteria, (2) modify the accelerated testing schedules, or 
(3) eliminate a reporting requirement. The change to the test 
acceptance criteria is based on the recommendations of Regulatory Guide 
1.9, and the change to the reporting requirement is enveloped by other 
NRC reporting requirements. The other changes are consistent with NRC 
guidance, and reduce unnecessary testing and improve EDG reliability. 
Requirements to assure that a common mode failure has not affected the 
remaining operable EDGs have been maintained. The existing routine 
testing frequency, unaffected by these changes, has been shown to be 
adequate for assuring the EDGs are operable based on operating 
experience. The proposed changes do not impact the assumptions of any 
design basis accident, and do not alter assumptions relative to the 
mitigation of an accident or transient event.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Bradley Fewell, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60555.
    NRC Branch Chief: Harold K. Chernoff.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: March 29, 2007.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station, Unit No. 1 Technical Specifications to 
increase the power level required for a reactor trip following a 
turbine trip (P-9 setpoint).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The analysis of the proposed change included an evaluation of loss 
of load/turbine trip transient. With systems functioning as designed, 
the proposed change to the P-9 setpoint does not impact [the] accident 
analyses previously evaluated in the Updated Final Safety Analysis 
Report (UFSAR). In the best estimate case (normal plant conditions; all 
control systems functioning per design), the pressurizer power operated 
relief valves (PORV) and the steam generator safety valves are not 
challenged following the turbine trip without reactor trip. 
Consequently, the proposed change does not adversely affect the 
probability of a small break loss of coolant accident due to a stuck-
open PORV. The sensitivity study that assessed the affects of degraded 
control systems found that a failure of all condenser steam dump valves 
resulted in challenging the PORVs and the steam generator (SG) safety 
valves. However, overfilling of the pressurizer will not occur and this 
Condition 2 event will not initiate a Condition 3 event. The challenge 
to the PORVs with all steam dump banks failed does not violate design 
or licensing criteria. Therefore, the proposed setpoint change does not 
significantly increase the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed setpoint change does not create the possibility of a 
new or different kind of accident than any accident previously 
evaluated in the FSAR. No new accident scenarios, failure mechanisms or 
limiting single failures are introduced as a result of the proposed 
change. The proposed Technical Specification changes have no adverse 
effects on any safety-related system and do not challenge the 
performance or integrity of any safety-related system. The revised 
setpoint for the P-9 function ensures that accident/transient analyses 
acceptance criteria continue to be met. This change makes no 
modifications to the plant that would introduce new accident causal 
mechanisms and has no affect on how the trip functions operate upon 
actuation. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    3. The proposed changes do not involve a significant reduction in 
the margin of safety.
    The proposed Technical Specification changes do not involve a 
significant reduction in a margin of safety. The analyses supporting 
the proposed change to the P-9 setpoint demonstrate that margin exists 
between the setpoint and the corresponding safety analysis limits. The 
calculations are based on plant instrumentation and calibration/

[[Page 41786]]

functional test methods and include allowances associated with the 
setpoint change. The results of analyses and evaluations supporting the 
proposed change demonstrate acceptance criteria continue to be met. The 
reactor trip on turbine trip provides additional protection and 
conservatism beyond that required for protection of public health and 
safety; the safety analyses in chapter 15 of the UFSAR do not take 
credit for this reactor trip. Therefore, the proposed changes do not 
involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Florida Power & Light Company, 
P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: Harold K. Chernoff.

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 request: June 27, 2007.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) 3.3.3, ``Post Accident Monitoring 
(PAM) Instrumentation,'' to include containment recirculation sump 
level instrumentation which will be used for indication of 
recirculation sump strainer blockage. Additionally, the amendment would 
revise TS 3.5.2, ``ECCS [Emergency Core Cooling System]--Operating,'' 
by replacing the term ``trash racks and screens'' with the more 
descriptive term ``strainers.'' Finally, the amendment would revise TS 
3.6.14, ``Containment Recirculation Drains,'' to include Limiting 
Conditions for Operation, Actions, and Surveillance Requirements to 
ensure the operability of flow paths credited in the evaluation of 
potential adverse effects of post-accident debris on the containment 
recirculation function pursuant to NRC Generic Letter 2004-02.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated?
    Response: No.
    The proposed change consists of a revision to the Technical 
Specifications (TS) for post accident monitoring (PAM) instrumentation 
to include new containment recirculation sump level instrumentation, a 
revision to the TS for Emergency Core cooling System (ECCS) to replace 
the term ``trash rack and screen'' with the term ``strainer,'' and a 
revision to the TS for containment recirculation drains to add two flow 
paths credited in the evaluation of the effects of post-accident debris 
on the containment recirculation functions pursuant to Nuclear 
Regulatory Commission Generic Letter 2004-02.
    The proposed TS revisions will not increase the probability of an 
accident because the associated components, i.e., the new sump level 
instruments, the new strainers, and the two flow paths, are not, and 
will not become, accident initiators. The activities involving these 
components pursuant to the proposed TS revisions consist of 
implementing Surveillance Requirements for the new sump level 
instruments and flow paths and actions to be taken if these components 
are inoperable. These activities will not increase the likelihood of an 
accident. The TS change associated with the sump strainers is editorial 
in that it reflects the terminology that has been applied to new pocket 
strainers that continue to perform the trash rack and screen functions. 
The change in terminology will not result in any new activities.
    The proposed TS revision will not increase the consequences of an 
accident because the associated components all provide mitigative 
functions for an accident, and their ability to perform their 
mitigative functions is not reduced by the associated TS changes. The 
TS changes associated with the new sump level instrumentation and the 
recirculation [flow paths] will provide increased assurance that these 
components will be available to perform their mitigative function if 
needed. The TS change associated with the sump strainers is editorial 
and does not affect the mitigative capability of the screens.
    Therefore, the proposed change will 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 revisions will not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because the associated components, i.e., the new sump level 
instruments, the new strainers, and the two flow paths, are components 
that will not initiate any accident. The proposed TS changes associated 
with these components will not cause them to be operated in any manner 
not previously evaluated for the specific components or for similar 
components, or cause them to become other than passive components.
    Therefore, the proposed change will 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 margin of safety associated with the proposed TS revisions 
involves the ability of the associated components, i.e., the new sump 
level instruments, the new strainers, and the two flow paths, to assure 
the ECCS and containment spray recirculation function can be adequately 
accomplished. The TS changes associated with the new sump level 
instrumentation and the recirculation [flow paths] will provide 
increased assurance that this function can be fulfilled. The TS change 
associated with the sump strainers is editorial and does not affect 
this function.
    Therefore, the proposed change will not involve a significant 
reduction in the margin of safety.
    The Nuclear Regulatory Commission (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: Kimberly Harshaw, Esquire, One Cook Place, 
Bridgman, MI 49106.
    NRC Acting Branch Chief: Travis Tate.

Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine 
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

    Date of amendment request: May 31, 2007.
    Description of amendment request: The proposed amendment would 
revise the accident source term used in the NMP2 design basis 
radiological consequence analyses in accordance with Title 10 of the 
Code of Federal Regulations (10 CFR), Part 50.67. The revised accident 
source term replaces the current methodology that is based

[[Page 41787]]

on TID-14844, ``Calculation of Distance Factors for Power and Test 
Reactor Sites,'' with the alternative source term (AST) methodology 
described in Regulatory Guide (RG) 1.183, ``Alternative Source Terms 
for Evaluating Design Basis Accidents at Nuclear Power Reactors.'' The 
amendment request is for full implementation of the AST as described in 
RG 1.183, with the exception that TID-14844 will continue to be used as 
the radiation dose basis for equipment qualification and vital area 
access. Proposed changes include the following: Revision of the 
Technical Specification (TS) definition of Dose Equivalent I-131 to be 
consistent with the AST analyses; TS changes that reflect revised 
design requirements regarding the use of the standby liquid control 
system (SLCS) to buffer the suppression pool pH to prevent iodine re-
evolution following a postulated design basis loss-of-coolant accident 
(LOCA); revisions to the TS operability requirements for the control 
room envelope filtration system and the control room envelope air 
conditioning system, consistent with the assumptions contained in the 
AST fuel-handling accident (FHA) analysis; and credit for operation of 
the residual heat removal system in the drywell spray mode for the 
post-LOCA removal of airborne elemental iodine and particulates from 
the drywell atmosphere. Because NMPNS is considering an extended power 
uprate (EPU) project that would increase the maximum licensed reactor 
core power level to 3,988 megawatts thermal (MWt), the AST analyses 
have been performed using a bounding core isotopic inventory that is 
based on operation at 3,988 MWt in lieu of the currently licensed power 
of 3,467 MWt.
    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.
    Adoption of the AST and those plant systems affected by 
implementing AST do not initiate DBAs [design-basis accidents]. The AST 
does not affect the design or manner in which the facility is operated; 
rather, for postulated accidents, the AST is an input to calculations 
that evaluate the radiological consequences. The AST does not by itself 
affect the post-accident plant response or the actual pathway of the 
radiation released from the fuel. It does, however, better represent 
the physical characteristics of the release, so that appropriate 
mitigation techniques may be applied. Implementation of the AST has 
been incorporated in the analyses for the limiting DBAs at NMP2.
    The structures, systems and components affected by the proposed 
change mitigate the consequences of accidents after the accident has 
been initiated. Application of the AST does result in changes to NMP2 
Updated Safety Analysis Report (USAR) functions (e.g., Standby Liquid 
Control system [SLCS]). As a condition of application of AST, NMPNS is 
proposing to use the [SLCS] to control the suppression pool pH 
following a LOCA. These changes do not require any physical 
modifications to the plant. As a result, the proposed changes do not 
involve a revision to the parameters or conditions that could 
contribute to the initiation of a DBA discussed in Chapter 15 of the 
NMP2 USAR. Since design basis accident initiators are not being altered 
by adoption of the AST, the probability of an accident previously 
evaluated is not affected.
    Plant-specific AST radiological analyses have been performed and, 
based on the results of these analyses, it has been demonstrated that 
the dose consequences of the limiting events considered in the analyses 
are within the acceptance criteria provided by the NRC for use with the 
AST. These criteria are presented in 10 CFR 50.67 and Regulatory Guide 
1.183. Even though the AST dose limits are not directly comparable to 
the previously specified whole body and thyroid dose guidelines of 
General Design Criterion 19 and 10 CFR 100.11, the results of the AST 
analyses have demonstrated that the 10 CFR 50.67 limits are satisfied. 
Therefore, it is concluded that adoption of the AST does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Based on the above discussion, it is concluded that the proposed 
changes do 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.
    Implementation of AST and the proposed changes does not alter or 
involve any design basis accident initiators. These changes do not 
involve any physical changes to the plant and do not affect the design 
function or mode of operations of systems, structures, or components in 
the facility prior to a postulated accident. Since systems, structures, 
and components are operated essentially no differently after the AST 
implementation, no new failure modes are created by this proposed 
change.
    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 proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The changes proposed are associated with a new licensing basis for 
analysis of NMP2 DBAs. Approval of the licensing basis change from the 
original source term to the AST is being requested. The results of the 
accident analyses performed in support of the proposed changes are 
subject to revised acceptance criteria. The limiting DBAs have been 
analyzed using conservative methodologies, in accordance with the 
guidance contained in Regulatory Guide 1.183, to ensure that analyzed 
events are bounding and that safety margin has not been reduced. The 
dose consequences of these limiting events are within the acceptance 
criteria presented in 10 CFR 50.67 and Regulatory Guide 1.183. Thus, 
the proposed changes continue to ensure that the doses at the exclusion 
area boundary and low population zone boundary, as well as in the 
control room, are within corresponding regulatory criteria.
    Therefore, by meeting the applicable regulatory criteria for AST, 
it is concluded that the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1700 K Street, NW., Washington, DC 20006.
    NRC Branch Chief: Mark G. Kowal.

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

    Date of amendment request: April 4, 2007.
    Description of amendment request: The licensee has proposed 
amending the existing license to allow the results of near-term 
surveys, performed on a portion of the plant site, to be included

[[Page 41788]]

in the eventual Final Status Survey (FSS) for license termination.
    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 change would allow survey results for a specific area 
within the licensed site area, performed prior to Humboldt Bay Power 
Plant (HBPP) Unit 3 decommissioning and dismantlement activities, to be 
used in the overall licensed site area Final Status Survey (FSS) for 
license termination. The FSS will be performed following completion of 
HBPP Unit 3 decommissioning and dismantlement activities. This proposed 
change would not change plant systems or accident analysis, and as 
such, would not affect initiators of analyzed events or assumed 
mitigation of accidents. Therefore, the proposed change does not 
increase 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 change does not involve a physical alteration to the 
plant or require existing equipment to be operated in a manner 
different from the present design. Implementation of a cross 
contamination prevention and monitoring plan will be done in accordance 
with plant procedures and licensing bases documents. Therefore, the 
proposed change does not create the possibility of a new or different 
kind of accident from any accident evaluated.
    (3) Does the change involve a significant reduction in a margin of 
safety?
    Response: No.
    The proposed change has no effect on existing plant equipment, 
operating practices, or safety analysis assumptions. 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: Ms. Jennifer K. Post, Pacific Gas and 
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
    NRC Branch Chief: Bruce Watson.

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: March 22, 2007.
    Description of amendment request: The proposed amendment supports 
full-scope implementation of an alternative source term (AST) 
methodology, in accordance with Section 50.67, ``Accident source 
term,'' of Title 10 of the Code of Federal Regulations (10 CFR) with 
the exception that Technical Information Document (TID) 14844, 
``Calculation of Distance Factors for Power and Test Reactor Sites,'' 
will continue to be used as the radiation dose basis for equipment 
qualification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The implementation of AST assumptions has been evaluated in 
revisions to the analyses of the following limiting DBAs [design-basis 
accidents].
     Loss-of-Coolant Accident.
     Fuel Handling Accident.
     Control Rod Ejection Accident.
     Locked Rotor Accident.
     Main Steam Line Break Accident.
     Steam Generator Tube Rupture Accident.
    Based upon the results of these analyses and evaluations, it has 
been demonstrated that, with the requested changes, the dose 
consequences of these limiting events satisfies the dose limits in 10 
CFR 50.67 and are within the regulatory guidance provided by the NRC 
for use with the AST methodology. The AST is an input to calculations 
used to evaluate the consequences of an accident and does not affect 
the plant response or the actual pathway of the activity released from 
the fuel. Therefore, it is concluded that AST does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Implementation of AST provides for elimination of the Fuel Handling 
Building ventilation system filtration TS [Technical Specification] 
requirements and elimination of Control Room ventilation filtration TS 
requirements in Modes 5 or 6. It also eliminates containment integrity 
TS requirements while handling irradiated fuel and during core 
alterations. The equipment affected by the proposed changes is 
mitigative in nature and relied upon after an accident has been 
initiated. The affected systems are not accident initiators; and 
application of the AST methodology is not an initiator of a design 
basis accident.
    Elimination of the requirement to suspend operations involving 
positive reactivity additions that could result in loss of required 
SHUTDOWN MARGIN or required boron concentration if the control room 
ventilation system is inoperable in Modes 5 or 6 does not increase the 
probability of an accident because the proposed change does not affect 
the design and operational controls to prevent dilution events. These 
same design and operational controls prevent a loss of SHUTDOWN MARGIN 
or a boron dilution event so that radiological consequences from these 
events are precluded.
    The proposed changes do not involve physical modifications to plant 
equipment and do not change the operational methods or procedures used 
for moving irradiated fuel assemblies. The proposed changes do not 
affect any of the parameters or conditions that could contribute to the 
initiation of any accidents. Relaxation of operability requirements 
during the specified conditions will not significantly increase the 
probability of occurrence of an accident previously analyzed. Since 
design basis accident initiators are not being altered by adoption of 
the AST, the probability of an accident previously evaluated is not 
affected.
    Administrative changes to delete a footnote from Technical 
Specification surveillance requirement 4.7.7.e.3) and a note from 
ACTION 20 of Technical Specification Table 3.3-3, in which the 
provisions of the notes have expired, does not impact the probability 
or consequences of an accident previously evaluated.
    Based on the above discussion, the proposed changes do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not involve a physical change. The change 
will allow the automatic start feature of systems no longer credited in 
the accident analyses for mitigation to be disabled through the STPNOC 
[STP Nuclear Operating Company]

[[Page 41789]]

modification process. Implementation of AST provides increased 
operating margins for filtration system efficiencies. Application of 
AST provides for relaxation of certain Control Room ventilation system 
filtration requirements. The Fuel Handling Building filtration and 
holdup is no longer credited in the AST analyses. Therefore, the Fuel 
Handling Building Exhaust Air Ventilation system is no longer required 
in the Technical Specifications. It also relaxes containment integrity 
requirements while handling irradiated fuel and during core 
alterations. Elimination of the requirement to suspend operations 
involving positive reactivity additions that could result in loss of 
required SHUTDOWN MARGIN or required boron concentration if the control 
room ventilation system is inoperable in Mode 5 or Mode 6 does not 
create the possibility of a new or different kind of accident because 
these events have already been analyzed in the safety analysis with a 
conclusion that adequate measures exist to prevent these events.
    Similarly, the proposed changes do not require any physical changes 
to any structures, systems or components involved in the mitigation of 
any accidents. Therefore, no new initiators or precursors of a new or 
different kind of accident are created. New equipment or personnel 
failure modes that might initiate a new type of accident are not 
created as a result of the proposed changes.
    Administrative changes to delete a footnote from Technical 
Specification surveillance requirement 4.7.7.e.3) and a note from 
ACTION 20 of Technical Specification Table 3.3-3, in which the 
provisions of the notes have expired, does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Based on the above discussion, the proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    Approval of a change from the original source term methodology 
(i.e., TID 14844) to an AST methodology, consistent with the guidance 
in RG [NRC Regulatory Guide] 1.183, will not result in a significant 
reduction in the margin of safety. The safety margins and analytical 
conservatisms associated with the AST methodology have been evaluated 
and were found acceptable. The results of the revised DBA analyses, 
performed in support of the proposed changes, are subject to specific 
acceptance criteria as specified in RG 1.183. The dose consequences of 
these DBAs remain within the acceptance criteria presented in 10 CFR 
50.67 and RG 1.183.
    Elimination of the requirement to suspend operations involving 
positive reactivity additions that could result in loss of required 
SHUTDOWN MARGIN or required boron concentration if the control room 
ventilation system is inoperable in Mode 5 or Mode 6 does not result in 
a reduction in a margin to safety because adequate measures exist to 
preclude radiological consequences from these events.
    The proposed changes continue to ensure that the doses at the 
exclusion area boundary (EAB) and low population zone boundary (LPZ), 
as well as the Control Room and Technical Support Center, are within 
the specified regulatory limits.
    Administrative changes to delete a footnote from Technical 
Specification surveillance requirement 4.7.7.e.3) and a note from 
ACTION 20 of Technical Specification Table 3.3-3, in which the 
provisions of the notes have expired, does not impact the margin of 
safety.
    Therefore, based on the above discussion, 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 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: A.H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Branch Chief: Thomas G. Hiltz.

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: May 21, 2007.
    Description of amendment request: The license amendment request 
proposes revising the Technical Specification (TS) Surveillance 
Requirement (SR) 4.5.2.d for the inspection of Emergency Core Cooling 
System (ECCS) sumps for consistency with t