Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 43528-43543 [06-6597]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 71, Number 147 (Tuesday, August 1, 2006)]
[Notices]
[Pages 43528-43543]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 06-6597]


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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 7, 2006 to July 19, 2006. The last 
biweekly notice was published on July 18, 2006 (71 FR 40742).

[[Page 43529]]

Notice of Consideration of Issuance of Amendments to Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination. 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 
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, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the 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

[[Page 43530]]

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 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.

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

    Date of amendments request: September 29, 2005, as supplemented by 
letter dated July 5, 2006.
    Description of amendments request: The amendments revised the 
Physical Security Plan to clarify the description of the owner 
controlled area vehicle checkpoint.
    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 amendment, which will clarify the description of a 
security feature of the Owner Controlled Area (OCA) Checkpoint, does 
not reduce the ability of the Security organization to prevent 
radiological sabotage and, therefore, does not increase the 
probability or consequences of a radiological release previously 
evaluated. The proposed Security Plan changes will not affect any 
important to safety systems or components, their mode of operation 
or operating strategies. The proposed Security Plan changes have no 
affect on accident initiators or mitigation. Therefore, the proposed 
amendment to the Security Plan 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 amendment to clarify the description of a security 
feature of the OCA Checkpoint does not affect the operation of 
systems important to safety. The Security Plan amendment does not 
affect any of the parameters or conditions that could contribute to 
the initiation of any accident. No new accident scenarios are 
created as a result of the proposed Security Plan changes. In 
addition, the design functions of equipment important to safety are 
not altered as a result of the proposed Security Plan changes. 
Therefore, the proposed Security Plan changes will not create the 
possibility of a new or different accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed Security Plan changes will not affect any important 
to safety systems or components, their mode of operation, or 
operating strategies. The proposed Security Plan changes have no 
affect on accident initiators or mitigation. The proposed amendment 
to the Security Plan does not reduce the effectiveness of any 
security/safeguards measures currently in place. Therefore, the 
proposed Security Plan changes will not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Janet S. Mueller, Director, Law Department, 
Arizona Public Service Company, P.O. Box 52034, Mail Station 8695, 
Phoenix, Arizona 85072-2034.
    NRC Branch Chief: David Terao.

Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power 
Station, Kewaunee County, Wisconsin

    Date of amendment request: June 28, 2006.
    Description of amendment request: The proposed amendment changed 
Kewaunee Power Station (KPS) Technical Specifications 3.3.b.3.B and 
3.3.b.4.A to increase the minimum required boron concentration in the 
refueling water storage tank (RWST) from 2400 parts per million (ppm) 
to 2500 ppm.
    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.
    Increasing the minimum required boron concentration in the RWST 
does not add, delete, or modify any KPS systems, structures, or 
components (SSCs). The RWST and its contents are not accident 
initiators. Rather, they are designed for accident mitigation. The 
effects of an increase in the minimum RWST boron concentration from 
2400 ppm to 2500 ppm are bounded by existing evaluations and 
determined to be acceptable. Thus, the proposed increase in minimum 
RWST boron concentration has no adverse effect on the ability of the 
plant to mitigate the effects of design basis accidents.
    Therefore, the proposed amendment does 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.
    Increasing the minimum required boron concentration in the RWST 
does not change

[[Page 43531]]

the design function of the RWST or the SSCs designed to deliver 
borated water in the RWST to the [reactor] core. Increasing the 
minimum required boron concentration in the RWST does not create any 
credible new failure mechanisms or malfunctions for plant equipment 
or the nuclear fuel. The safety function of the borated water in the 
RWST is not being changed.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    An evaluation has been performed showing that maintaining RWST 
boron concentration above 2500 ppm continues to assure acceptable 
results for design basis accident analyses [ ] considering the 
reactivity of the core. Increasing the minimum boron concentration 
in the RWST from 2400 ppm to 2500 ppm increases the margin of safety 
in the KPS safety analyses, since additional post-accident negative 
reactivity will be available to the core. This additional negative 
reactivity more than compensates for the additional reactivity in 
the core due to the unanticipated prolonged shutdown periods in 
Cycle 27. Additionally, the proposed new minimum boron concentration 
of 2500 ppm is within the range required by current safety analyses 
(i.e., 2400 ppm to 2625 ppm), and well below the currently 
acceptable maximum boron concentration of 2625 ppm.
    The proposed amendment does not result in altering or exceeding 
a design basis or safety limit for the plant. All current fuel 
design criteria will continue to be satisfied, and the safety 
analyses of record (except for the postLOCA sump boron 
concentration), including evaluations of the radiological 
consequences of design basis accidents, will remain applicable.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Branch Chief: L. Raghavan.

Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian 
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New 
York

    Date of amendment request: May 31, 2006.
    Description of amendment request: The proposed amendment revised 
the Technical Specification (TS) requirements related to steam 
generator (SG) tube integrity. Specifically, it would revise the TS 
definition of LEAKAGE; TS 3.4.13, ``Reactor Coolant System (RCS) 
Operational Leakage;'' TS 5.5.7 (Indian Point Unit 2) and TS 5.5.8 
(Indian Point Unit 3), ``Steam Generator (SG) Program;'' TS 5.6.7 
(Indian Point Unit 2) and TS 5.6.8 (Indian Point Unit 3), ``SG Tube 
Inspection Report;'' and would create new TS 3.4.17, ``SG Tube 
Integrity.''
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF 449, Revision 4. The NRC staff 
issued a notice of opportunity for comment in the Federal Register on 
March 2, 2005 (70 FR 10298), on possible amendments concerning TSTF-
449, including a model safety evaluation and model no significant 
hazards consideration (NSHC) determination, using the consolidated line 
item improvement process (CLIIP). The NRC staff subsequently issued a 
notice of availability of the models for referencing in license 
amendment applications in the Federal Register on May 6, 2005 (70 FR 
24126). The licensee affirmed the applicability of the following NSHC 
determination in its application dated May 31, 2006.
    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, which is presented below:

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

    The proposed change requires a SG Program that includes 
performance criteria that will provide reasonable assurance that the 
SG tubing will retain integrity over the full range of operating 
conditions (including startup, operation in the power range, hot 
standby, cooldown and all anticipated transients included in the 
design specification). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE.
    A steam generator tube rupture (SGTR) event is one of the design 
basis accidents that are analyzed as part of a plant's licensing 
basis. In the analysis of a SGTR event, a bounding primary to 
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits 
in the licensing basis plus the LEAKAGE rate associated with a 
double-ended rupture of a single tube is assumed.
    For other design basis accidents such as MSLB, rod ejection, and 
reactor coolant pump locked rotor the tubes are assumed to retain 
their structural integrity (i.e., they are assumed not to rupture). 
These analyses typically assume that primary to secondary LEAKAGE 
for all SGs is 1 gallon per minute or increases to 1 gallon per 
minute as a result of accident induced stresses. The accident 
induced leakage criterion introduced by the proposed changes 
accounts for tubes that may leak during design basis accidents. The 
accident induced leakage criterion limits this leakage to no more 
than the value assumed in the accident analysis.
    The SG performance criteria proposed change to the TS identify 
the standards against which tube integrity is to be measured. 
Meeting the performance criteria provides reasonable assurance that 
the SG tubing will remain capable of fulfilling its specific safety 
function of maintaining reactor coolant pressure boundary integrity 
throughout each operating cycle and in the unlikely event of a 
design basis accident. The performance criteria are only a part of 
the SG Program required by the proposed change to the TS. The 
program, defined by Nuclear Energy Institute (NEI) 97-06, Steam 
Generator Program Guidelines, includes a framework that incorporates 
a balance of prevention, inspection, evaluation, repair, and leakage 
monitoring. The proposed changes do not, therefore, significantly 
increase the probability of an accident previously evaluated.
    The consequences of design basis accidents are, in part, 
functions of the DOSE EQUIVALENT 1-131 in the primary coolant and 
the primary to secondary LEAKAGE rates resulting from an accident. 
Therefore, limits are included in the plant technical specifications 
for operational leakage and for DOSE EQUIVALENT 1-131 in primary 
coolant to ensure the plant is operated within its analyzed 
condition. The typical analysis of the limiting design basis 
accident assumes that primary to secondary leak rate after the 
accident is 1 gallon per minute with no more than [500 gallons per 
day or 720 gallons per day] in any one SG, and that the reactor 
coolant activity levels of DOSE EQUIVALENT 1-131 are at the TS 
values before the accident.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary coolant chemistry controls. The 
proposed approach updates the current TSs and enhances the 
requirements for SG inspections. The proposed change does not 
adversely impact any other previously evaluated design basis 
accident and is an improvement over the current TSs.
    Therefore, the proposed change does not affect the consequences 
of a SGTR accident and the probability of such an accident is 
reduced. In addition, the proposed changes do not affect the 
consequences of a main steam line break (MSLB), rod ejection, or a 
reactor coolant pump locked rotor event, or other previously 
evaluated accident.

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

    The proposed performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed SG Program will not 
introduce any adverse changes to the plant design basis or 
postulated accidents resulting from potential tube degradation. The 
result of the implementation of the SG Program will be an 
enhancement of SG tube performance.

[[Page 43532]]

Primary to secondary LEAKAGE that may be experienced during all 
plant conditions will be monitored to ensure it remains within 
current accident analysis assumptions.
    The proposed change does not affect the design of the SGs, their 
method of operation, or primary or secondary coolant chemistry 
controls. In addition, the proposed change does not impact any other 
plant system or component. The change enhances SG inspection 
requirements.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.

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

    The SG tubes in pressurized water reactors are an integral part 
of the reactor coolant pressure boundary and, as such, are relied 
upon to maintain the primary system's pressure and inventory. As 
part of the reactor coolant pressure boundary, the SG tubes are 
unique in that they are also relied upon as a heat transfer surface 
between the primary and secondary systems such that residual heat 
can be removed from the primary system. In addition, the SG tubes 
isolate the radioactive fission products in the primary coolant from 
the secondary system. In summary, the safety function of an SG is 
maintained by ensuring the integrity of its tubes.
    Steam generator tube integrity is a function of the design, 
environment, and the physical condition of the tube. The proposed 
change does not affect tube design or operating environment. The 
proposed change is expected to result in an improvement in the tube 
integrity by implementing the SG Program to manage SG tube 
inspection, assessment, repair, and plugging. The requirements 
established by the SG Program are consistent with those in the 
applicable design codes and standards and are an improvement over 
the requirements in the current TSs.
    For the above reasons, the margin of safety is not changed and 
overall plant safety will be enhanced by the proposed change to the 
TS.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Branch Chief: Richard J. Laufer.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant, Units 3 and 4, Miami-Dade County, Florida

    Date of amendment request: April 27, 2006.
    Description of amendment request: The proposed amendments revised 
the Technical Specifications (TSs) relating to Steam Generator (SG) 
inspection. Specifically, TS 3/4.4.5, Surveillance Requirements, and TS 
3/4.4.6, Reactor Coolant System Leakage, would be modified to clearly 
delineate the scope of the inservice inspections required in the tube 
sheet regions of the SGs.
    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?
    Of the various accidents previously evaluated, the proposed 
changes only affect the SG tube rupture (SGTR) event evaluation and 
the postulated steam line break [SLB] accident evaluation. Loss-of-
coolant accident (LOCA) conditions cause a compressive axial load to 
act on the tube. Therefore, since the LOCA tends to force the tube 
into the tubesheet rather than pull it out, it is not a factor in 
this amendment request. Another faulted load consideration is a safe 
shutdown earthquake (SSE); however, the seismic analysis of Series 
44F SGs has shown that axial loading of the tubes is negligible 
during a SSE.
    For the SGTR event, the required structural margins of the SG 
tubes will be maintained by the presence of the tubesheet. Tube 
rupture is precluded for cracks in the hydraulic expansion region 
due to the constraint provided by the tubesheet. Therefore, 
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR 
[Pressurized-Water Reactor] Steam Generator Tubes,'' margins against 
burst are maintained for both normal and postulated accident 
conditions.
    The limited inspection length of 17 inches supplies the 
necessary resistive force to preclude pullout loads under both 
normal operating and accident conditions. The contact pressure 
results from the hydraulic expansion process, thermal expansion 
mismatch between the tube and tubesheet and from the differential 
pressure between the primary and secondary side. The proposed 
changes do not affect other systems, structures, components or 
operational features. Therefore, the proposed change results in no 
significant increase in the probability of the occurrence of a SGTR 
event.
    The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary 
leakage flow through a postulated broken tube is not affected by the 
proposed change since the tubesheet enhances the tube integrity in 
the region of the hydraulic expansion by precluding tube deformation 
beyond its initial expanded outside diameter. The resistance to both 
tube rupture and collapse is strengthened by the tubesheet in that 
region. At normal operating pressures, leakage from primary water 
stress corrosion cracking (PWSCC) below 17 inches from the top of 
the tubesheet is limited by both the tube-to-tubesheet crevice and 
the limited crack opening permitted by the tubesheet constraint. 
Consequently, negligible normal operating leakage is expected from 
cracks within the tubesheet region.
    The probability of a SLB is unaffected by the potential failure 
of a SG tube as the failure of a tube is not an initiator for a SLB 
event. SLB leakage is limited by leakage flow restrictions resulting 
from the crack and tube-to-tubesheet contact pressures that provide 
a restricted leakage path above the indications and also limit the 
degree of crack face opening compared to free span indications. The 
leak rate during postulated accident conditions would be expected to 
be less than twice that during normal operation for indications near 
the bottom of the tubesheet (including indications in the tube end 
welds) based on the observation that while the driving pressure 
increases by about a factor of two, the flow resistance increases 
with an increase in the tube-to-tubesheet contact. While such a 
decrease is rationally expected, the postulated accident leak rate 
is bounded by twice the normal operating leak rate if the increase 
in contact pressure is ignored. Since normal operating leakage is 
limited to less than 150 gpd, the attendant accident condition leak 
rate, assuming all leakage to be from lower tubesheet indications, 
would be bounded by 300 gpd. This value is less than the 500 gpd 
leak rate assumed during a postulated SLB in the Turkey Point Units 
3 and 4 Updated Final Safety Analysis Report (UFSAR).
    Therefore, based on the above evaluation, the proposed changes 
do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes do not introduce any changes or mechanisms 
that create the possibility of a new or different kind of accident. 
Tube bundle integrity is expected to be maintained for all plant 
conditions upon implementation of the limited tubesheet inspection 
depth methodology. The proposed changes do not introduce any new 
equipment or any change to existing equipment. No new effects on 
existing equipment are created nor are any new malfunctions 
introduced.
    Therefore, based on the above evaluation, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes maintain the required structural margins of 
the SG tubes for both normal and accident conditions. NEI [Nuclear 
Energy Institute] 97-06, Rev. 2 and RG 1.121 are used as the basis 
in the development of the limited tubesheet inspection depth 
methodology for determining that SG tube integrity considerations 
are maintained within acceptable limits. RG 1.121 describes a

[[Page 43533]]

method acceptable to the NRC staff for meeting General Design 
Criteria 14, 15, 31, and 32 by reducing the probability and 
consequences of an SGTR. RG 1.121 concludes that by determining the 
limiting safe conditions of tube wall degradation beyond which tubes 
with unacceptable cracking, as established by inservice inspection, 
should be removed from service or repaired, the probability and 
consequences of a SGTR are reduced. This RG uses safety factors on 
loads for tube burst that are consistent with the requirements of 
Section III of the ASME [American Society of Mechanical Engineers] 
Code.
    For axially oriented cracking located within the tubesheet, tube 
burst is precluded due to the presence of the tubesheet. For 
circumferentially oriented cracking, WCAP [Westinghouse Commercial 
Atomic Power] --16506-P defines a length of degradation free 
expanded tubing that provides the necessary resistance to tube 
pullout due to the pressure induced forces (with applicable safety 
factors applied). Application of the limited tubesheet inspection 
depth criteria will preclude unacceptable primary-to-secondary 
leakage during all plant conditions. The methodology for determining 
leakage provides for large margins between calculated and actual 
leakage values in the proposed limited tubesheet inspection depth 
criteria.
    Plugging of the SG tubes reduces the reactor coolant flow margin 
for core cooling. Implementation of the 17 inch inspection length at 
Turkey Point Units 3 and 4 will result in maintaining the margin of 
flow that may have otherwise been reduced by tube plugging.
    Based on the above, it is concluded that the proposed changes do 
not result in any reduction of margin with respect to plant safety 
as defined in the UFSAR or Bases of the plant Technical 
Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Branch Chief: Michael L. Marshall, Jr.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: November 14, 2005.
    Description of amendment request: The proposed amendment revised 
the table of Primary Containment Isolation Instrumentation to eliminate 
the trip generated by the main steamline radiation monitors.
    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 change deletes the Main Steamline Radiation Monitor 
(MSLRM) trip function from TS [technical specification]. The MSLRM 
is not an initiator of any accident previously evaluated. As a 
result, the probability of any accident previously evaluated is not 
significantly increased. The consequences of an accident previously 
evaluated, specifically the Control Rod Drop Accident (CRDA), have 
been evaluated consistent with the DAEC [Duane Arnold Energy Center] 
licensing basis utilizing the Alternative Source Term (10 CFR 
50.67). As demonstrated by the dose calculations, the consequences 
of the accident are within the regulatory acceptance criterion. As a 
result, the consequences of any accident previously evaluated are 
not significantly increased. Therefore, the proposed change does 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.
    No new or different accidents result from utilizing the proposed 
change. The changes do not involve a change in the methods governing 
normal plant operation. The equipment proposed to be removed from 
the plant, the MSLRM, is only credited in the CRDA analysis and no 
other event in the safety analysis. The proposed changes are 
consistent with the revised safety analysis assumptions for a CRDA 
included in this application.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change deletes the requirement for the MSLRM 
isolation function. Analyses performed consistent with the DAEC 
licensing basis, demonstrate that the removal of this isolation will 
not cause a significant reduction in the margin of safety, as the 
resulting offsite dose consequences are being maintained within 
regulatory 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 request involves no significant hazards consideration.
    Attorney for licensee: Mr. R.E. Helfrich, Florida Power & Light 
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: L. Raghavan.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: December 22, 2005.
    Description of amendment request: The proposed amendment revised 
the reactor-pressure vessel material surveillance program described 
within the Duane Arnold Energy Center (DAEC) Updated Final Safety 
Analysis Report from a plant-specific program to the Boiling Water 
Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance 
Program (ISP).
    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 change implements an integrated surveillance 
program that has been evaluated by the NRC [Nuclear Regulatory 
Commission] staff as meeting the requirements of paragraph III.C of 
Appendix H to 10 CFR 50. Consequently, the proposed change does not 
significantly increase the probability of any accident previously 
evaluated. The proposed change provides the same assurance of RPV 
[reactor pressure vessel] integrity. As a result, the consequences 
of any accident previously evaluated are not significantly 
increased.
    Therefore, the proposed change does 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.
    The proposed change revises the DAEC licensing bases to reflect 
participation in the BWRVIP ISP. The ISP was approved by the NRC 
staff as an acceptable material surveillance program which complies 
with 10 CFR 50, Appendix H. The proposed change maintains an 
equivalent level of RPV material surveillance and does not introduce 
any new accident initiators. The proposed change will not impact the 
manner in which the plant is designed or operated. This change will 
not affect the reactor pressure vessel, as no physical changes are 
involved. The proposed change will not cause the reactor pressure 
vessel or interfacing systems to be operated outside of any design 
or testing limits. Furthermore, the proposed changes will not alter 
any assumptions

[[Page 43534]]

previously made in evaluating the radiological consequences of any 
accident.
    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 amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change has been evaluated as providing an 
acceptable alternative to the plant-specific RPV material 
surveillance program that meets the requirements of the regulations 
for RPV material surveillance. The material surveillance program 
requirements contained in 10 CFR 50, Appendix H provide assurance 
that adequate margins of safety exist for the reactor coolant system 
against nonductile or rapidly propagating failures during normal 
operation, anticipated operational occurrences, and system 
hydrostatic tests.
    The BWRVIP ISP has been approved by the NRC staff as an 
acceptable material surveillance program which complies with I0 CFR 
50, Appendix H. The ISP will provide the material surveillance data 
which will ensure that the safety margins required by NRC 
regulations are maintained for the DAEC reactor coolant system.
    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: Mr. R. E. Helfrich, Florida Power & Light 
Company, P. O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: L. Raghavan.

FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of amendment request: April 28, 2006.
    Description of amendment request: The proposed amendment modified 
technical specifications (TSs) requirements for inoperable snubbers by 
adding Limiting Condition for Operation (LCO) 3.0.8. The changes are 
consistent with Nuclear Regulatory Commission approved Industry/
Technical Specification Task Force (TSTF) standard TS change TSTF-372, 
Revision 4.
    The NRC staff issued a notice of availability of a model safety 
evaluation and model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed 
the applicability of the model NSHC determination in its application 
dated April 28, 2006.
    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 allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. Entrance into Actions 
or delaying entrance into Actions is not an initiator of any 
accident previously evaluated. Consequently, the probability of an 
accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on the delay time allowed 
before declaring a TS supported system inoperable and taking its 
Conditions and Required Actions are no different than the 
consequences of an accident under the same plant conditions while 
relying on the existing TS supported system Conditions and Required 
Actions. Therefore, the consequences of an accident previously 
evaluated are not significantly increased by this change. Therefore, 
this 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 allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. The proposed change 
does not involve a physical alteration of the plant (no new or 
different type of equipment will be installed) or a change in the 
methods governing normal plant operation. Thus, this 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?
    Response: No.
    The proposed change allows a delay time before declaring 
supported TS systems inoperable when the associated snubber(s) 
cannot perform its required safety function. The proposed change 
restores an allowance in the pre-ISTS conversion TS that was 
unintentionally eliminated by the conversion. The pre-ISTS TS were 
considered to provide an adequate margin of safety for plant 
operation, as does the post-ISTS conversion TS. Therefore, this 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. R.E. Helfrich, Florida Power & Light 
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Branch Chief: L. Raghavan.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook 
Nuclear Plant, Unit 1, Berrien County, Michigan

    Date of amendment request: April 10, 2006.
    Description of amendment request: The proposed amendment revised 
Surveillance Requirement 3.8.1.11 of the Donald C. Cook Technical 
Specifications, raising the emergency diesel generator full load 
rejection voltage test limit from 5000 volts to 5350 volts.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee provided a 
no significant hazards determination analysis, which is reproduced 
below:

    1. Does the proposed change involve a significant increase in 
the probability of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    Probability of Occurrence of an Accident Previously Evaluated.
    The proposed change is an increase in the Technical 
Specification (TS) Surveillance Requirement (SR) limit on maximum 
voltage following an emergency diesel generator (DG) full load 
rejection. The DGs' safety function is solely mitigative and is not 
needed unless there is a loss of offsite power. The DGs do not 
affect any accident initiators or precursors of any accident 
previously evaluated. The proposed increase in the TS SR limit does 
not affect the DGs' interaction with any system whose failure or 
malfunction can initiate an accident. Therefore, the probability of 
occurrence of an accident previously evaluated is not significantly 
increased.
    Consequences of an Accident Previously Evaluated.
    The DG safety function is to provide power to safety related 
components needed to mitigate the consequences of an accident 
following a loss of offsite power. The purpose of the TS SR voltage 
limit is to assure DG damage protection following a full load 
rejection. The technical analysis performed to support this proposed 
amendment has demonstrated that the DGs can withstand voltages above 
the new proposed limit without a loss of protection. The proposed 
higher limit will continue to provide assurance that the DG is 
protected, and the safety function of the DG will be unaffected by 
the proposed change. Therefore, the consequences of an accident 
previously evaluated will not be significantly increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    There are no new DG failure modes created and the DGs are not an 
initiator of any new

[[Page 43535]]

or different kind of accident. The proposed increase in the TS SR 
limit does not affect the interaction of the DGs with any system 
whose failure or malfunction can initiate an accident. 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 margins of safety applicable to the proposed change are 
those associated with the ability of the DGs to perform their safety 
function. The technical analysis performed to support this amendment 
demonstrates that this ability will be unaffected. The increase in 
the TS SR limit will not affect this ability. Therefore, the 
proposed change does not involve a significant reduction in margin 
of safety.
    The NRC staff evaluated the licensee's analysis, and based on 
this evaluation, the NRC staff proposes to determine that the 
requested amendment does not involve a significant hazards 
consideration.

    Attorney for licensee: James M. Petro, Jr., Esquire, One Cook 
Place, Bridgman, MI 49106.
    NRC Branch Chief: L. Raghavan.

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

    Date of amendment request: June 16, 2006.
    Description of amendment request: The proposed amendment revised 
Technical Specification (TS) 3.10.1, ``Inservice Leak and Hydrostatic 
Testing Operation,'' to extend the scope to include provisions for 
temperature increases above 212 [deg]F as a consequence of inservice 
leak or hydrostatic testing, and as a consequence of control rod scram 
time testing initiated in conjunction with the inservice leak test or 
hydrostatic test, when initial test conditions are below 212 [deg]F.
    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.
    Current TS LCO [Limiting Condition for Operation] 3.10.1 allows 
average RCS [reactor coolant system] temperature to exceed 212 
[deg]F when required during the conduct of hydrostatic and inservice 
leak tests without requiring entry into plant operating Mode 3, Hot 
Shutdown. Extending this allowance to testing in which average RCS 
temperature exceeds 212 [deg]F as a consequence of maintaining 
pressure and to the performance of scram time testing that is 
initiated in conjunction with the hydrostatic and inservice leak 
tests will not impact any accident initiator. Thus, the proposed 
change does not affect the probability of any accident.
    The proposed changes do not involve any modification of 
equipment used to mitigate accidents, and do not impact any system 
used in the mitigation of design basis accidents. The proposed 
changes do not involve modified operation of equipment or [a] system 
used to mitigate accidents. Thus, the proposed changes do not affect 
the consequences of an accident.
    Based on the above, NPPD concludes that 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 TS revisions to TS LCO 3.10.1 do not involve 
physical modification of the plant or a change in plant operation. 
The proposed TS revisions do not revise or eliminate any existing 
requirements, and do not impose any additional requirements. The 
proposed changes do not alter assumptions made in the safety 
analysis, and are consistent with the safety analysis assumptions 
and current plant operating practice. Allowing the performance of 
control rod scram time testing, while in plant operating Mode 4 with 
average RCS temperature greater than 212 [deg]F, does not create the 
possibility of a different kind of accident.
    Based on the above NPPD[,] concludes that these 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 do not impact the design or operation of 
the Reactor Protection System or the Emergency Core Cooling System. 
Allowing completion of scram time testing that was initiated in 
conjunction with inservice leak or hydrostatic testing prior to 
reactor criticality and startup will eliminate the need for 
unnecessary plant maneuvers to control reactor temperature and 
pressure, thereby resulting in enhanced safe operation.
    Based on the above, NPPD concludes that these 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. John C. McClure, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Branch Chief: David Terao.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: January 18, 2006.
    Description of amendment request: The proposed amendment deleted 
the reference to the hydrogen monitors in Technical Specification (TS) 
3.6.11, ``Accident Monitoring Instrumentation'' consistent with the 
NRC-approved Industry/Technical Specification Task Force (TSTF) 
Standard Technical Specification Change Traveler, TSTF-447, 
``Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen 
Monitors.''
    The NRC staff issued a notice of availability of ``Model 
Application Concerning Technical Specification Improvement To Eliminate 
Hydrogen Recombiner Requirement, and Relax the Hydrogen and Oxygen 
Monitor Requirements for Light Water Reactors Using the Consolidated 
Line Item Improvement Process (CLIIP)'', in the Federal Register on 
September 25, 2003 (68 FR 55416). The notice included a model safety 
evaluation (SE), a model no significant hazards consideration (NSHC) 
determination, and a model application.
    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, by confirming the applicability of the model NSHC 
determination to NMP-1 and incorporating it by reference in its 
application. The model NSHC determination is presented below:

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

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen [and

[[Page 43536]]

oxygen] monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG [Regulatory Guide] 1.97 Category 1, is intended for key variables 
that most directly indicate the accomplishment of a safety function 
for design-basis accident events. The hydrogen [and oxygen] monitors 
no longer meet the definition of Category 1 in RG 1.97. As part of 
the rulemaking to revise 10 CFR 50.44 the Commission found that 
Category 3, as defined in RG 1.97, is an appropriate categorization 
for the hydrogen monitors because the monitors are required to 
diagnose the course of beyond design-basis accidents. [Also, as part 
of the rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 2, as defined in RG 1.97, is an appropriate categorization 
for the oxygen monitors, because the monitors are required to verify 
the status of the inert containment.]
    The regulatory requirements for the hydrogen [and oxygen] 
monitors can be relaxed without degrading the plant's, emergency 
response. The emergency response, in this sense, refers to the 
methodologies used in ascertaining the condition of the reactor 
core, mitigating the consequences of an accident, assessing and 
projecting offsite releases of radioactivity, and establishing 
protective action recommendations to be communicated to offsite 
authorities. Classification of the hydrogen monitors as Category 3, 
[classification of the oxygen monitors as Category 2] and removal of 
the hydrogen [and oxygen] monitors from TS will not prevent an 
accident management strategy through the use of the SAMGs [severe 
accident management guidelines], the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen [and oxygen] monitor 
requirements, including removal of these requirements from TS, does 
not involve a significant increase in the probability or the 
consequences of any accident previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, will not result in 
any failure mode not previously analyzed. The hydrogen recombiner 
and hydrogen [and oxygen] monitor equipment was intended to mitigate 
a design-basis hydrogen release. The hydrogen recombiner and 
hydrogen [and oxygen] monitor equipment are not considered accident 
precursors, nor does their existence or elimination have any adverse 
impact on the pre-accident state of the reactor core or post 
accident confinement of radionuclides within the containment 
building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen [and oxygen] monitor requirements, 
including removal of these requirements from TS, in light of 
existing plant equipment, instrumentation, procedures, and programs 
that provide effective mitigation of and recovery from reactor 
accidents, results in a neutral impact to the margin of safety.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3 hydrogen monitors are adequate to provide rapid 
assessment of current reactor core conditions and the direction of 
degradation while effectively responding to the event in order to 
mitigate the consequences of the accident. The intent of the 
requirements established as a result of the TMI [Three Mile Island], 
Unit 2 accident can be adequately met without reliance on safety-
related hydrogen monitors.
    [Category 2 oxygen monitors are adequate to verify the status of 
an inerted containment.]
    Therefore, this change does not involve a significant reduction 
in [a] margin of safety. [The intent of the requirements established 
as a result of the TMI, Unit 2 accident can be adequately met 
without reliance on safety-related oxygen monitors.]
    Removal of hydrogen [and oxygen] monitoring from TS will not 
result in a significant reduction in their functionality, 
reliability, and availability.
    The NRC staff has reviewed the model NSHC determination and its 
applicability to NMP-1. Based on this review, 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: Richard J. Laufer.

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

    Date of amendment request: June 7, 2006.
    Description of amendment request: The amendment deleted Required 
Action D.1.2 in Technical Specification (TS) 3.7.10, ``Control Room 
Emergency Ventilation System (CREVS),'' and Required Action C.1.2 in TS 
3.7.11, ``Control Room Air Conditioning System (CRACS).'' These 
required actions are for the condition where the required actions and 
completion time (CT) of TS 3.7.10 Condition A (one CREVS train 
inoperable) and TS 3.7.11 Condition A (one CRACS train inoperable) are 
not met in Modes 5 or 6, or during movement of irradiated fuel 
assemblies. The deleted required actions, and associated CTs, are to 
verify the operable CREVS (or CRACS) train is capable of being powered 
by an emergency power source.
    The amendment would also delete the phrase ``in MODES 1, 2, 3, or 
4'' from Condition A (one emergency exhaust system (EES) train 
inoperable) of TS 3.7.13, ``Emergency Exhaust System (EES),'' and 
revise Condition D to state the following: ``Required Action and 
associated Completion Time of Condition A not met during movement of