Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 6780-6795 [E7-2323]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 29 (Tuesday, February 13, 2007)]
[Notices]
[Pages 6780-6795]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-2323]


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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 January 19, 2007, to February 1, 2007. The 
last biweekly notice was published on January 30, 2007 (72 FR 4304).

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

[[Page 6781]]

following amendment requests involve no significant hazards 
consideration. Under the Commission's regulations in 10 CFR 50.92, this 
means that operation of the facility in accordance with the proposed 
amendment would not (1) involve a significant increase in the 
probability or consequences of an accident previously evaluated; or (2) 
create the possibility of a new or different kind of accident from any 
accident previously evaluated; or (3) involve a significant reduction 
in a margin of safety. The basis for this proposed determination for 
each amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. 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:

[[Page 6782]]

(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 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)7ndash;(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.

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

    Date of amendment request: September 26, 2006.
    Description of amendment request: The proposed change deletes 
reference to the containment fan cooler (CFC) condensate flow switch 
from Technical Specification (TS) 3.4.5.1, ``Reactor Coolant System 
Leakage--Leakage Detection Instrumentation,'' and to modify or delete 
associated actions.
    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 Reactor Coolant System (RCS) leakage detection systems are 
passive monitoring systems therefore the proposed changes do not 
affect reactor operations or accident analyses and have no 
radiological consequences. The proposed change continues to require 
diverse methods of monitoring leakage. The gaseous radioactivity 
monitor, although not included in the TSs and the CFC condensate 
flow switches, which are proposed for removal from the TSs, will be 
maintained functional and available.
    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 introduces no new mode of plant operation or 
any plant modification. The RCS leakage detection instrumentation is 
used solely for monitoring purposes and is not part of plant control 
instruments or engineered safety feature actuation circuits. The 
change does not vary or affect any plant operating condition or 
parameter.
    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 modify any of the RCS leakage 
detection instrumentation. The proposed change continues to require 
diverse methods of monitoring leakage. In addition, although not 
required by TS, multiple means of diverse monitoring RCS leakage 
will remain functional and available.
    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: David Terao.

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

    Date of amendment request: January 18, 2007.
    Description of amendment request: The proposed change will revise 
the description of Grand Gulf Nuclear Station Technical Specification 
4.2.2, ``Control Rod Assemblies,'' to allow to the use of hafnium as an 
additional type of control material.
    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 NRC has specifically approved the use of hafnium as neutron 
absorbing material for use in BWR [boiling-water reactor] control 
rod assemblies. The use of hafnium in control rods as a neutron 
absorber material does not significantly alter the neutronic or 
mechanical functional characteristics of the control rods. Control 
rod designs using hafnium have been successfully used in other BWRs. 
Since control rods that utilize hafnium have a longer lifetime, the 
probability of some accidents involving the handling, on-site 
storage, and shipping of irradiated rods will actually be reduced. 
The proposed change does not alter the required number of control 
rods nor does it affect any of the specifications related to the 
control rods (e.g., the shutdown margin and scram timing 
requirements are unaffected).
    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 application of a control rod design using hafnium as an 
absorber material does not produce any new mode of plant operation 
or alter the control rods in such a way as to affect their function 
or operability since the new control rods are designed to be 
compatible with the existing control rods.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

[[Page 6783]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not significantly affect the neutronic 
or mechanical characteristics of the control rods since the hafnium 
containing controls rods are designed to be compatible with the 
existing design and reload licensing criteria; therefore, there is 
no significant change in the margin of safety. It does not change 
the required number of existing control rods. It does not affect the 
existing Technical Specifications related to control rods (e.g., 
required shutdown margin and scram time, etc.).
    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: David Terao.

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

    Date of amendment request: October 11, 2006.
    Description of amendment request: The proposed amendment would 
modify the plant Improved Technical Specifications (ITSs) to implement 
a more conservative requirement in ITS 3.7.7, ``Nuclear Services Closed 
Cycle Cooling Water (SW) System.'' The current Action A allows the 
plant to operate for up to 72 hours before initiating a shutdown when 
one required SW heat exchanger is inoperable. The proposed revision 
will only allow operation to continue for 8 hours before initiating a 
shutdown when one required SW heat exchanger is inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    (1) Does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The limiting design basis accident for CR-3 includes, as an 
assumption, adequate heat removal capability by the SW system. The 
amendment is being proposed to ensure the SW system performs its 
design basis function. Adequate heat removal is provided by three 
OPERABLE SW heat exchangers. The 8 hour completion time will reduce 
the window that the plant can operate with only two SW heat 
exchangers before a shutdown is required. The proposed change does 
not increase the probability of an accident previously evaluated 
since the amendment is not a modification to plant systems, nor a 
change to plant operation that could initiate an accident. 
Therefore, granting the LAR [license amendment request] does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. The dose consequences of all 
design basis accidents are unchanged by this proposed amendment.
    (2) Does not create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    The function of the SW system considered in the design basis is 
to remove process and operating heat from safety-related components 
during normal as well as transient conditions. The proposed 
amendment to limit the allowed ACTION Completion Time to 8 hours 
will ensure the function of the SW system is consistent with the 
design basis and will not result in changes to the design, physical 
configuration of the plant or the assumptions made in the safety 
analysis. The requirement does not change the function of the system 
nor its ability to perform its design function. No alteration to 
plant configuration or operation is proposed. Therefore, the 
proposed change will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    (3) Does not involve a significant reduction in a margin of 
safety?
    CR-3's design basis considers adequate heat removal by the SW 
system to cool the containment fan assembly cooling coils and fan 
motors, spent fuel pool, SW pump motors and other equipment which 
must function following an accident. This proposed amendment will 
not alter the current design basis. By limiting the allowed ACTION 
Completion Time to 8 hours, the proposed amendment to ITS 3.7.7 will 
limit the time the safety function of the SW system can be 
compromised. Therefore, the amendment does not result in a reduction 
of the margin of safety.

    The NRC staff has reviewed the analysis provided for Florida Power 
Corporation 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 (Acting): Margaret H. Chernoff.

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit 2, Dauphin County, Pennsylvania

    Date of amendment request: December 13, 2006.
    Description of amendment requests: The amendment application 
proposes to delete Technical Specification (TS) 6.8.1.3, which provides 
the requirement for submittal of the annual occupational radiation 
exposure report.
    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? 
No
    The proposed change eliminates the Technical Specification 
reporting requirement for occupational radiation exposure 
information, which is in excess to that required to be submitted by 
regulations. The proposed change involves no changes to plant 
systems or accident analyses. As such, the change is administrative 
in nature and does not affect initiators of analyzed events or 
assumed mitigation of 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? 
No
    The proposed change does not involve a physical alteration of 
the plant, add any new equipment, or require any existing equipment 
to be operated in a manner different from the present design. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety? No
    This change is an administrative change to reporting 
requirements of occupational radiation exposure data and will not 
reduce a margin of safety because it has no effect on any safety 
analyses assumptions. Hence, this change is administrative in 
nature. For these reasons, the proposed change does not involve a 
significant reduction in the margin of safety.

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

[[Page 6784]]

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

    Date of amendment request: December 21, 2006.
    Description of amendment request: The proposed amendment revises 
the licensing basis to reflect a revision to the spent fuel pool 
criticality analysis methodology and a new criticality analysis. In 
addition, associated changes are proposed to Technical Specifications 
3.7.12, ``Spent Fuel Storage,'' and 4.3.1, ``Criticality,'' to reflect 
the results of the new criticality analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously 
evaluated?
    Response: No
    Operation of the facility in accordance with the proposed 
amendment request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
presence of soluble boron in the Spent Fuel Pool (SFP) water being 
used for criticality control does not increase the probability of a 
dropped fuel assembly accident within the pool. The handling of the 
fuel assemblies in the SFP has always been performed and will 
continue to be performed in borated water.
    There is no increase in the probability of the accidental 
misloading of fuel assemblies into the SFP fuel storage racks when 
considering the presence of soluble boron in the pool water for 
criticality control. Fuel assembly placement will continue to be 
controlled pursuant to approved fuel handling procedures and in 
accordance with the spent fuel storage rack limitations specified in 
the Technical Specifications (TS). There is no increase in the 
consequences for an accidental misloading of fuel assemblies in the 
SFP fuel storage racks because the criticality analyses demonstrate 
that the pool will remain subcritical following an accidental 
misloading.
    Soluble boron credit is used to provide margin to offset 
uncertainties, tolerances, and off-normal/accident conditions, and 
to provide subcritical margin such that the SFP keff 
[effective neutron multiplication constant] is maintained less than 
or equal to 0.95. The plant-specific criticality analysis results 
demonstrate that the spent fuel rack keff will remain<1.0 
(at a 95/95 percent probability and confidence level) even with the 
SFP flooded with unborated water.
    There is no increase in the probability of the loss of normal 
cooling to the SFP water when considering the presence of soluble 
boron in the pool water for subcriticality control since a high 
concentration of soluble boron has always been maintained in the SFP 
water.
    A loss of normal cooling to the SFP water causes an increase in 
the temperature of the water passing through the stored fuel 
assemblies. This causes a decrease in water density, which would 
result in a net increase in reactivity when soluble boron is present 
in the water. However, the additional negative reactivity provided 
by the 2100 ppm [parts per million] boron concentration limit, above 
that provided by the concentration required (805 ppm) to maintain 
keff less than or equal to 0.95, will compensate for the 
increased reactivity which could result from a loss of SFP cooling 
event. Because adequate soluble boron will be maintained in the SFP 
water the consequences of a loss of normal cooling to the SFP will 
not be increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No
    Under the proposed amendment, no changes are being made to the 
fuel storage racks themselves, to any other systems, or to the 
physical structures of the Primary Auxiliary Building. Therefore, 
there are no changes proposed to the plant configuration, equipment 
design, or installed equipment.
    Criticality accidents in the SFP are not new or different types 
of accidents. They have been analyzed in the FSAR [Final Safety 
Analysis Report] and in fuel storage criticality analysis reports 
associated with specific licensing amendments. The proposed new SFP 
storage limitations are consistent with the assumptions made in the 
new criticality analysis, and will not have any significant effect 
on normal SFP operations and maintenance, and do not create the 
possibility of a new or different kind of accident. Verifications 
will continue to be performed to ensure that the SFP loading 
configuration meets specified requirements.
    The current TS includes a SFP boron concentration limit that 
conservatively bounds the boration assumption of the new criticality 
analysis. Since soluble boron has always been maintained in the SFP 
water, implementation of this requirement for SFP criticality 
control purposes has have no effect on normal pool operations and 
maintenance. Also, since soluble boron has always been present in 
the SFP, a dilution event has always been a possibility. The loss of 
substantial amounts of soluble boron from the SFP that could lead to 
keff exceeding 0.95 was evaluated as part of the analyses 
in support of this license amendment request. The evaluation 
demonstrates that a dilution of the SFP boron concentration from the 
minimum TS concentration of 2100 to 805 ppm is not credible.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    (3) Does the proposed amendment result in a significant 
reduction in a margin of safety?
    Response: No
    The proposed Technical Specification changes providing the 
resulting spent fuel storage operation limits provide adequate 
safety margin to ensure that the stored fuel assembly array always 
remains subcritical. These limits are based on a plant-specific 
criticality analysis performed in accordance with the present 
Westinghouse spent fuel rack criticality analysis methodology which 
allows credit for soluble boron.
    The criticality analysis takes credit for soluble boron to 
ensure that keff will be less than or equal to 0.95 under 
normal circumstances. While the criticality analysis used credit for 
soluble boron, storage configurations have been defined using 95/95 
keff calculations to ensure that the spent fuel rack 
keff is less than unity (0.995) with no soluble boron. 
Soluble boron credit is used to provide safety margin to offset 
uncertainties, tolerances, and off-normal/accident conditions, and 
to provide subcritical margin such that the SFP keff is 
maintained less than or equal to 0.95.
    The loss of substantial amounts of soluble boron from the SFP 
that could lead to keff exceeding 0.95 was evaluated as 
part of the analyses in support of this license amendment request. 
The evaluation demonstrates that a dilution of the SFP boron 
concentration from the minimum TS concentration of 2100 to 805 ppm 
is not credible. Also, the plant-specific criticality analysis 
results demonstrate that even if a complete dilution were to occur 
the spent fuel rack keff would remain <1.0 (at a 95/95 
percent probability and confidence level) with the SFP flooded with 
unborated water. The plant-specific criticality analysis performed 
in accordance with the conservative analysis methodology of the 
Westinghouse licensing topical report demonstrates that the 
requirements of 10 CFR 50.68 and 10 CFR 50, Appendix A, General 
Design Criterion 62 will be satisfied. 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: Jonathan Rogoff, Esquire, Vice President, 
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street, 
Hudson, WI 54016.
    NRC Acting Branch Chief: Patrick D. Milano.

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 request: December 29, 2006.
    Description of amendment request: The proposed amendments would

[[Page 6785]]

revise Technical Specification (TS) 5.5.8 to indicate that the 
Inservice Testing Program shall include testing frequencies applicable 
to the American Society of Mechanical Engineers Code for Operations and 
Maintenance (ASME OM Code), and to indicate that there may be some non-
standard frequencies specified as 2 years or less in the Inservice 
Testing Program to which the provisions of Surveillance Requirement 
(SR) 3.0.2 are applicable. The proposed changes are consistent with 
NRC-approved Technical Specification Task Force (TSTF) Travelers TSTF-
479, Revision 0, ``Changes to Reflect Revision of 10 CFR 50.55a,'' and 
TSTF-497, Revision 0, ``Limit Inservice Testing Program SR 3.0.2 
Application to Frequencies of 2 Years or Less.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes revise TS 5.5.8, ``lnservice Testing 
Program,'' for consistency with 10 CFR 50.55a(f)(4) requirements 
regarding inservice testing of pumps and valves. The proposed change 
incorporates revisions to the ASME Code that result in a net 
improvement in the measures for testing pumps and valves.
    The proposed changes do not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. They do not involve the addition or removal of any 
equipment, or any design changes to the facility. Therefore, the 
proposed changes do not represent 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 proposed changes do not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released off-site and there is no increase in individual or 
cumulative occupational exposure. Therefore, this proposed change 
does not create the possibility of an accident of a different kind 
than previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No
    The proposed changes revise TS 5.5.8, ``lnservice Testing 
Program,'' for consistency with the requirements of 10 CFR 
50.55a(f)(4) regarding the inservice testing of pumps and valves. 
The proposed change incorporates revisions to the ASME Code that 
result in a net improvement in the measures for testing pumps and 
valves. The safety function of the affected pumps and valves will be 
maintained. Therefore, this 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 requests involve no significant hazards consideration.
    Attorney for licensee: Antonio Fernandez, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

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: December 29, 2006.
    Description of amendment requests: The proposed amendments will 
revise Technical Specification (TS) 5.5.16 for consistency with the 
requirements of 10 CFR 50.55a(g)(4) for components classified as Code 
Class CC. This regulation requires licensees to update their 
containment inservice inspection requirements in accordance with 
Subsections IWE and IWL of Section XI, Division I of the American 
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code 
as limited by 10 CFR 50.55a(b)(2)(vi) and modified by 10 CFR 
50.55a(b)(2)(viii) and 10 CFR 50.55a(b)(2)(ix). This license amendment 
request is consistent with NRC-approved Industry/Technical 
Specification Task Force (TSTF) Traveler number TSTF-343, ``Containment 
Structural Integrity.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises the Technical Specification (TS) 
administrative controls programs for consistency with the 
requirements of 10 CFR [Part] 50, paragraph 55a(g)(4) for components 
classified as Code Class CC.
    The proposed change affects the frequency of visual examinations 
that will be performed for the concrete surfaces of the containment 
for the purpose of the Containment Leakage Rate Testing Program. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The frequency of visual examinations of the concrete 
surfaces of the containment and the mode of operation during which 
those examinations are performed has no relationship to or adverse 
impact on the probability of any of the initiating events assumed in 
the accident analyses. The proposed change would allow visual 
examinations that are performed pursuant to NRC-approved ASME 
[Code,] Section XI requirements (except where relief has been 
granted by the NRC) to meet the intent of visual examinations 
required by Regulatory Guide 1.163, without requiring additional 
visual examinations pursuant to the Regulatory Guide. The intent of 
early detection of deterioration will continue to be met by the more 
rigorous requirements of the Code-required visual examinations. As 
such, the safety function of the containment as a fission product 
barrier is maintained.
    The proposed change does not impact any accident initiators or 
analyzed events or assumed mitigation of accident or transient 
events. It does not involve the addition or removal of any 
equipment, or any design changes to the facility.
    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 accident from any accident previously evaluated?
    Response: No.
    The proposed change revises the TS Administrative Controls 
programs for consistency with the requirements of 10 CFR [Part] 50, 
paragraph 55a(g)(4) for components classified as Code Class CC.
    The change affects the frequency of visual examinations that 
will be performed for the concrete surfaces of the containments. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The proposed change does not involve a modification to the 
physical configuration of the plant (i.e., no new equipment will be 
installed) or a change in the methods governing normal plant 
operation. The proposed change will not impose any new or different 
requirements or introduce a new accident initiator, accident 
precursor, or a malfunction mechanism. Additionally, there is no 
change in the types or increases in the amounts of any effluent that 
may be released offsite and there is no increase in individual or 
cumulative occupational exposure.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?

[[Page 6786]]

    Response: No.
    The proposed change revises the TS Administrative Controls 
programs for consistency with the requirements of 10 CFR [Part] 50, 
paragraph 55a(g)(4) for components classified as Code Class CC.
    The change affects the frequency of visual examinations that 
will be performed for the concrete surfaces of the containments. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The safety function of the containment as a fission product 
barrier will be maintained.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Antonio Fern[aacute]ndez, Esq., Pacific Gas 
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

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: December 29, 2006.
    Description of amendment requests: The proposed amendments will 
revise Technical Specification (TS) 3.4.1, ``RCS [Reactor Coolant 
System] Pressure, Temperature, and Flow Departure from Nucleate Boiling 
(DNB) Limits,'' and TS 5.6.5, ``CORE OPERATING LIMITS REPORT (COLR). 
This license amendment request proposes to relocate the RCS DNB 
parameters for pressurizer pressure and RCS average temperature to the 
COLR. This relocation is consistent with Technical Specification Task 
Force Traveler TSTF-339, Revision 2, ``Relocate TS Parameters to 
COLR.'' TS 5.6.5 is revised to add topical reports WCAP-8567-P-A, 
``Improved Thermal Design Procedure,'' and WCAP-11596-P-A, 
``Qualification of the PHOENIX-P/ANC Nuclear Design System for 
Pressurized Water Reactor Cores,'' by name and title only. These 
changes are consistent with TSTF-363, Revision 0, ``Revise Topical 
Report References in ITS 5.6.5, COLR.''
    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 changes are programmatic and administrative in 
nature, and do not physically alter safety-related systems or affect 
the way in which safety-related systems perform their functions. The 
proposed changes relocate cycle-specific parameters from Technical 
Specification (TS) 3.4.1 to the Core Operating Limits Report (COLR). 
This does not change plant design or affect system operating 
parameters. The proposed changes do not, by themselves, alter any of 
the parameters. Removal of the cycle-specific parameters from the TS 
does not eliminate existing requirements to comply with the 
parameters. Also, TS 5.6.5 is revised to add topical reports WCAP-
8567-P-A, ``Improved Thermal Design Procedure,'' and WCAP-11596-P-A, 
``Qualification of the PHOENIX-P/ANC Nuclear Design System for 
Pressurized Water Reactor Cores,'' as they are approved analytical 
methods for determining core operating limits.
    Although relocation of the cycle-specific parameters to the COLR 
would allow revision of the affected parameters without prior NRC 
approval, there is no significant effect on the probability or 
consequences of an accident previously evaluated. Future changes to 
the COLR parameters could result in event consequences that are 
either slightly less or slightly more severe than the consequences 
for the same event using the present parameters. The differences 
would not be significant and would be bounded by the existing 
requirement of TS 5.6.5c to meet the applicable limits of the safety 
analyses.
    The cycle-specific parameters being transferred from the TS to 
the COLR will continue to be controlled under existing programs and 
procedures. The Final Safety Analysis Report Update (FSARU) accident 
analyses will continue to be examined with respect to changes in the 
cycle-dependent parameters obtained using NRC reviewed and approved 
reload design methodologies to ensure that the transient evaluation 
of new reload designs are bounded by previously accepted analyses. 
This examination will continue to be performed pursuant to 10 CFR 
50.59 requirements, ensuring that future reload designs use NRC-
approved methodologies and do not involve more than a minimal 
increase in the probability or consequences of an accident 
previously evaluated in the FSARU.
    The proposed changes do not allow for an increase in plant power 
levels, do not increase the production, and do not alter the flow 
path or method of disposal of radioactive waste or byproducts. 
Therefore, the proposed changes do not change the type or increase 
the amount of effluents released offsite.
    The proposed changes to TS 5.6.5b to reference only the topical 
report number and title for five of the topical reports do not alter 
the analytical methods that have been previously reviewed and 
approved by the NRC. This method of referencing topical reports 
would allow the use of current topical reports to support limits in 
the COLR without having to submit a request for an amendment to the 
operating license. Implementation of revisions to these topical 
reports would still be reviewed in accordance with 10 CFR 50.59 and, 
where required, revisions would be submitted to the NRC for approval 
prior to implementation.
    Therefore, the proposed change does 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 accident from any accident previously evaluated?
    Response: No.
    The proposed changes that relocate cycle-specific parameters 
from the TS to the COLR, thus removing the requirement for prior NRC 
approval of revisions to those parameters, do not involve a physical 
change to the plant. No new equipment is being introduced, and 
installed equipment is not being operated in a new or different 
manner. No changes are being made to the parameters within which the 
plant is operated, other than their relocation to the COLR. No 
protective or mitigative action setpoints are affected by the 
proposed changes. The proposed changes will not alter the manner in 
which equipment operation is initiated, nor will the functional 
demands on credited equipment be changed. No change to procedures 
that ensure the plant remains within analyzed limits are being 
proposed, and no change is being made to procedures relied upon to 
respond to an off-normal event. As such, no new failure modes are 
being introduced.
    Relocation of cycle-specific parameters does not influence, 
impact, or contribute in any way to the possibility of a new or 
different kind of accident. The relocated cycle-specific parameters 
will continue to be calculated using the NRC-reviewed and approved 
methodology. The proposed changes do not alter assumptions made in 
the safety analysis, and operation within the core operating limits 
will continue.
    The proposed changes to reference only the topical report number 
and title do not alter the use of the analytical methods that have 
been previously reviewed and approved by the NRC. This method of 
referencing topical reports would allow the use of current topical 
reports to support limits in the COLR without having to submit a 
request for an amendment to the operating license. Implementation of 
revisions to topical reports would still be reviewed in accordance 
with 10 CFR 50.59 and, where required, would receive NRC review and 
approval.
    The addition of WCAP-8567-P-A and WCAP-11596-P-A to TS 5.6.5 is 
a clarification to provide a complete listing of approved analytical 
methods used for determining core operating limits.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety is established through equipment design, 
operating parameters, and the setpoints at which automatic actions 
are

[[Page 6787]]

initiated. The proposed changes do not physically alter safety-
related systems, nor do they affect the way in which safety-related 
systems perform their functions. No protective or mitigative action 
setpoints are affected by the proposed changes. Therefore, 
sufficient equipment remains available to actuate upon demand for 
the purpose of mitigating an analyzed event. As the proposed changes 
to relocate cycle-specific parameters to the COLR will not affect 
plant design or system operating parameters, there is no detrimental 
impact on any equipment design parameter, and the plant will 
continue to be operated within prescribed limits.
    The development of cycle-specific parameters for future reload 
designs will continue to conform to NRC-reviewed and approved 
methodologies, and will be performed pursuant to 10 CFR 50.59 to 
assure that the plant operates within cycle-specific parameters.
    The proposed changes to reference only the topical report number 
and title do not alter the use of the analytical methods used to 
determine core operating limits that have been reviewed and approved 
by the NRC. This method of referencing topical reports would allow 
the use of current NRC-approved topical reports to support limits in 
the COLR without having to submit a request for an amendment to the 
operating license. Implementation of revisions to topical reports 
would still be reviewed in accordance with 10 CFR 50.59 and, where 
required, receive NRC review and approval.
    The addition of WCAP-8567-P-A and WCAP-11596-P-A to TS 5.6.5 is 
a clarification to provide a complete listing of approved analytical 
methods used for determining core operating limits.
    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 requests involve no significant hazards consideration.
    Attorney for licensee: Antonio Fern[aacute]ndez, Esq., Pacific Gas 
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

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: January 11, 2007.
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TSs) to support replacement of the 
steam generators (SGs) at Diablo Canyon Power Plant, Unit Nos. 1 and 2. 
Revisions are proposed to TS 3.3.2, ``Engineered Safety Feature 
Actuation System (ESFAS) Instrumentation,'' TS 5.5.9, ``Steam Generator 
(SG) Program,'' and TS 5.6.10, ``Steam Generator (SG) Tube Inspection 
Report.'' The replacement SGs are to be installed during the Diablo 
Canyon Power Plant, Unit No. 2, 14th refueling outage (2R14), currently 
scheduled for February 2008, and the Unit No. 1, 15th refueling outage 
(1R15), currently scheduled for January 2009.
    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 revised engineered safety feature actuation system (ESFAS) 
steam generator (SG) Water Level-High High feedwater isolation 
Nominal Trip Setpoint and Allowable Value have been determined using 
the existing setpoint methodology approved for Diablo Canyon Power 
Plant. The setpoint analysis for the replacement steam generators 
(RSGs) accounts for the setpoint uncertainties specific to the RSG 
design. The revised Feedwater Isolation SG Water Level-High High (P-
14) Nominal Trip Setpoint and Allowable Value are applied using a 
conservative surveillance requirement methodology. The function of 
the ESFAS instrumentation is unchanged. The Feedwater Isolation SG 
Water Level-High High (P-14) ESFAS instrumentation will continue to 
function in a manner consistent with the plant design basis and 
satisfy all the requirements of the safety analyses.
    The probability and consequences of accidents previously 
evaluated in the Final Safety Analysis Report (FSAR) Update are not 
adversely affected because the revised Feedwater Isolation SG Water 
Level-High High (P-14) Nominal Trip Setpoint and Allowable Value 
continue to assure a conservative plant response to high SG level, 
consistent with the safety analyses and licensing basis.
    The proposed changes revise and clarify the surveillance 
requirements for ESFAS Function 5.b, Feedwater Isolation SG Water 
Level-High High (P-14). These changes ensure that this function will 
actuate as assumed in the safety analyses.
    The proposed changes to TS 5.5.9 delete the alternate repair 
criteria (ARC) for the existing SGs, incorporate tube inspection 
periods applicable to Alloy 690 thermally treated tubes, and delete 
the TS 5.6.10 reporting requirements for ARC. The TS 5.5.9 SG 
structural integrity, accident induced leakage, and operational 
leakage performance criteria will continue to be met for the RSGs. 
Meeting the SG performance criteria provides reasonable assurance 
that the SG tubes will remain capable of maintaining reactor coolant 
pressure boundary integrity throughout each operating cycle and in 
the unlikely event of a design basis accident. Removal of the ARC 
for the existing SGs will ensure that all tubes found by inservice 
inspection to contain flaws with a depth equal to or exceeding 40 
percent of the nominal tube wall thickness will be plugged as 
required by TS 5.5.9.c. With the revised SG tube inspection period, 
the SGs will continue to meet the SG program defined by NEI [Nuclear 
Energy Institute] 97-06, ``Steam Generator Program Guidelines,'' 
which incorporates a balance of prevention, inspection, evaluation, 
repair, and leakage monitoring.
    Removal of the ARC will reduce the allowable accident induced 
leakage following a main steamline break accident. The proposed 
changes do not have any impact on the accident induced leakage 
assumed in the other design basis accidents. The changes do not have 
any impact on the allowable SG operational leakage, allowable 
reactor coolant system activity, or the allowable SG secondary 
activity.
    The proposed changes will not affect the probability of any 
accident initiators. There will be no degradation in the performance 
of, or an increase in the number of challenges imposed on, safety-
related equipment assumed to function during an accident. There will 
be no change to accident mitigation performance. The proposed 
changes will not alter any assumptions or change any mitigation 
actions in the radiological consequence evaluations in the FSAR 
Update.
    Therefore the proposed changes 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 accident from any accident previously evaluated?
    Response: No.
    The proposed changes will not affect the normal method of plant 
operation or create new methods of plant operation related to the 
Feedwater Isolation SG Water Level-High High (P-14) ESFAS setpoints. 
The proposed changes to the Feedwater Isolation SG Water Level-High 
High (P-14) instrumentation surveillance requirements will provide 
assurance that the plant will operate within the limits assumed in 
the safety analyses. The assumptions made in the setpoint analyses 
for the Feedwater Isolation SG Water Level-High High (P-14) ESFAS 
instrument do not create any new accidents, accident initiators, or 
failure mechanisms.
    The proposed changes, which delete the TS 5.5.9 ARC for the 
existing SGs, incorporate tube inspection periods for Alloy 690 
thermally-treated tubes in TS 5.5.9, and delete the ARC reporting 
requirements in TS 5.6.10, will not introduce any adverse changes to 
the plant design basis or postulated accidents resulting from 
potential tube degradation. The primary-to-secondary leakage that 
may be experienced during all plant conditions will be monitored to 
ensure it remains within current safety analysis assumptions. The 
proposed changes do not adversely affect the method of operation of 
the SGs or the primary or secondary coolant controls and do not 
impact other plant systems or components.

[[Page 6788]]

    Therefore, the proposed changes do not create the possibility of 
a new or different accident from any accident previously evaluated.
    3. [Do] the proposed change[s] involve a significant reduction 
in a margin of safety?
    Response: No.
    The FSAR Update Excessive Heat Removal due to Feedwater System 
Malfunctions event credits the Feedwater Isolation SG Water Level-
High High (P-14) ESFAS instrumentation. The safety analysis limit 
assumed for the Feedwater Isolation SG Water Level-High High (P-14) 
ESFAS instrumentation for this event has not changed for the safety 
analyses for the RSGs. None of the acceptance criteria for Excessive 
Heat Removal due to Feedwater System Malfunctions event are changed 
as a result of the revised Feedwater Isolation SG Water Level-High 
High (P-14) Nominal Trip Setpoint and Allowable Value. The 
instrument surveillance requirement changes for the Feedwater 
Isolation SG Water Level-High High (P-14) function ensure that the 
instrumentation will actuate as assumed in the safety analysis.
    The safety function of the SGs is maintained by ensuring the 
integrity of the tubes. SG tube integrity is a function of the 
design, environment, and the physical condition of the SG tubes. The 
proposed changes, which delete the TS 5.5.9 ARCs for the existing 
SGs, incorporate tube inspection periods for Alloy 690 thermally 
treated tubes in TS 5.5.9, and delete the ARC reporting requirements 
in TS 5.6.10, do not adversely impact the SG tube design or 
operating environment. SG tube integrity will continue to be 
maintained by implementing the SG Program to manage SG tube 
inspection, assessment, and repair. The requirements established by 
the SG program are consistent with those in the applicable design 
codes and standards.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Antonio Fern[aacute]ndez, Esq., Pacific Gas 
and Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Branch Chief: David Terao.

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

    Date of amendment request: May 17, 2006.
    Description of amendment request: The licensee has proposed to 
modify the Physical Security Plan (PSP) to allow leaving certain 
security posts temporarily under emergency conditions requiring 
personnel to evacuate occupied plant areas for their health and safety.
    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.
    Allowing the security posts and monitoring requirements of PSP, 
Sections 3.1.4 and 4.3, and Table 7-1, to not be continuously 
maintained has no impact on the probability of an accident from 
occurring, especially acts of nature such as earthquakes and 
tsunamis.
    The HBPP Defueled Safety Analysis Report, Appendix A, and NRC 
Safety Evaluation Report (SER), Section 10, dated April 29, 1987, 
evaluate various accidents at HBPP. Because all fuel has been 
removed from the reactor vessel and stored in the spent fuel pool, 
the majority of accidents analyzed pertain to events that could only 
affect spent fuel or the spent fuel pool. All accidents affecting 
spent fuel or the spent fuel pool do not require security personnel 
action to protect the public health and safety, or to maintain 
offsite radiological doses well within regulatory limits. In 
addition, NRC SER, Section 10.7, ``Impact of Tsunami Flooding,'' 
analyzes the impact of tsunami flooding. That analysis identifies a 
likely impact of the tsunami to be a release of the radwaste tank 
radionuclide contents to the bay and some damage to the reactor 
building. For both situations, no security personnel action is 
required to maintain offsite radiological doses well within 
regulatory limits.
    Allowing the security posts and monitoring requirements of PSP, 
Sections 3.1.4 and 4.3, and Table 7-1, to not be continuously 
maintained temporarily, under emergency conditions, does not create 
problems that could increase the consequences of an accident. The 
primary function of the manning and monitoring requirements of PSP, 
Sections 3.1.4 and 4.3, and Table 7-1, is to monitor, detect and 
assess unauthorized intrusion into the protected area, and has 
nothing to do with the probability or consequences of plant 
accidents.
    If security personnel evacuate PSP, Section 3.1.4 and Table 7-1, 
security posts during a tsunami, those security personnel will be 
able to return to the PSP, Section 3.1.4 and Table 7-1, security 
posts after the tsunami and assess damage or intrusion by observing 
alarms and/or physical conditions as well as resume implementation 
of security post and monitoring requirements of PSP, Sections 3.1.4 
and 4.3, and Table 7-1. In addition, upon evacuation, security 
personnel notify offsite security backup pe