Involving No Significant Hazards Considerations, 5233-5254 [05-1574]

Agencies

• Biweekly Notice; Applications and Amendments to Facility Operating Licenses
• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 20 (Tuesday, February 1, 2005)]
[Notices]
[Pages 5233-5254]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-1574]


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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 7, 2005, through January 19, 2005. 
The last biweekly notice was published on January 18, 2005 (70 FR 
2886).

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

[[Page 5234]]

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, 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 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

[[Page 5235]]

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.

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

    Date of amendment request: September 15, 2004.
    Description of amendment request: The proposed amendment would 
delete requirements from the Technical Specifications (TSs) to maintain 
hydrogen recombiners and hydrogen and oxygen monitors. A notice of 
availability for this TS improvement using the consolidated line item 
improvement process was published in the Federal Register on September 
25, 2003 (68 FR 55416). Licensees were generally required to implement 
upgrades as described in NUREG-0737, ``Clarification of TMI [Three Mile 
Island] Action Plan Requirements,'' and Regulatory Guide (RG) 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
combustible gas control were imposed by order for many facilities and 
were added to, or included, in the TSs for nuclear power reactors 
currently licensed to operate. The revised Title 10 of the Code of 
Federal Regulations (10 CFR) Section 50.44, ``Standards for combustible 
gas control system in light-water-cooled power reactors,'' eliminated 
the requirements for hydrogen recombiners and relaxed safety 
classifications and licensee commitments to certain design and 
qualification criteria for hydrogen and oxygen monitors.
    The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice 
of availability of a model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on September 25, 2003 (68 FR 55416). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated September 15, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of NSHC is presented below:

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 NRC 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 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 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 NRC 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 NRC 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 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 severe accident 
management guidelines, the emergency plan, the emergency operating 
procedures, and site survey monitoring that support modification of 
emergency plan protective action recommendations.
    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 the 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 NRC has found that this hydrogen 
release is not risk-significant

[[Page 5236]]

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, 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 the 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 proposes to determine that the amendment request 
involves NSHC.
    Attorney for licensee: Mr. Thomas S. O'Neill, Associate General 
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road, 
Warrenville, IL 60666.
    NRC Section Chief: Gene Y. Suh.

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: December 16, 2004.
    Description of amendments request: The requested change will delete 
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 16, 2004.
    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 the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating letter 
report of shutdown experience and operating statistics if the 
equivalent data is submitted using an industry electronic database. 
It also eliminates the TS reporting requirement for an annual 
occupational radiation exposure report, which provides information 
beyond that specified in NRC 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 or 
transients. 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 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?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Section Chief: Robert A. Gramm.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: December 1, 2004.
    Description of amendments request: The requested change will delete 
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 1, 2004.
    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 the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating report 
of shutdown experience and operating statistics if the equivalent 
data is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC 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 or transients. 
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 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?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: December 6, 2004.
    Description of amendment request: The requested change will delete

[[Page 5237]]

Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated December 6, 2004.
    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 the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the TSs reporting requirements to 
provide a monthly operating report of shutdown experience and 
operating statistics if the equivalent data is submitted using an 
industry electronic database. It also eliminates the TS reporting 
requirement for an annual occupational radiation exposure report, 
which provides information beyond that specified in NRC 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 or transients. 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 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?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in the margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significant hazards consideration.
    Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB, 
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279 .
    NRC Section Chief: M. Kotzalas (Acting).

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

    Date of amendment request: June 6, 2004.
    Description of amendment request: The proposed change would modify 
the Millstone Power Station, Unit No. 2 Technical Specifications (TSs) 
to extend the 10-year test interval for the Integrated Leakage Rate 
Test program to 15 years from the last Type A test. Specifically, the 
proposed change would revise TS 6.19, ``Containment Leakage Rate 
Testing [CLRT] Program,'' and permit a one-time, 5-year extension of 
the 10-year performance-based Type A test interval. In addition, the 
testing would be in accordance with the CLRT Program, Regulatory Guide 
(RG) 1.163, ``Performance-Based Containment Leak-Test Program'' and 
surveillance testing requirements as proposed in Nuclear Energy 
Institute 94-01 for Type A testing.
    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 license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed extension to Type A testing cannot increase the 
probability of an accident previously evaluated since extension of 
the containment Type A testing is not a physical plant modification 
that could alter the probability of accident occurrence, nor is it 
an activity or modification that by itself could lead to equipment 
failure or accident initiation.
    The proposed one-time, five-year extension to Type A testing 
does not result in a significant increase in the consequences of an 
accident as documented in NUREG-1493. The NUREG notes that very few 
potential containment leakage paths are not identified by Type B and 
C tests. It concludes that even reducing the Type A (ILRT 
[integrated leak rate test]) testing frequency to once per twenty 
years leads to an imperceptible increase in risk.
    DNC (the licensee) provides a high degree of assurance through 
indirect testing and inspection that the containment will not 
degrade in a manner detectable only by Type A testing. The last two 
Type A tests identified containment leakage within acceptance 
criteria, indicating a very leak-tight containment. Inspections 
required by the ASME Code [American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code] are also performed in order to 
identify indications of containment degradation that could affect 
leak-tightness. Separately, Type B and C testing required by 
Technical Specifications, identifies any containment opening from 
design penetrations, such as valves, that would otherwise be 
detected by a Type A test. These factors establish that a one-time, 
five-year extension to the Millstone Unit 2 Type A test interval 
will not represent a significant increase in the consequences of an 
accident.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The proposed revision to the Technical Specifications adds a 
one-time extension to the current interval for Type A testing for 
Millstone Unit 2. The current test interval of ten years, based on 
past performance, would be extended on a one-time basis to fifteen 
years from the last Type A test. The proposed extension to Type A 
testing does not create the possibility of a new or different type 
of accident since there are no physical changes being made to the 
plant and there are no changes to the operation of the plant that 
could introduce a new failure.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed revision to Millstone Unit 2 Technical 
Specifications adds a one-time extension to the current interval for 
Type A testing. The current test interval of ten years, based on 
past performance, would be extended on a one-time basis to fifteen 
years from the last Type A test for Millstone Unit 2. RG 1.174 
provides guidance for determining the risk impact of plant-specific 
changes to the licensing basis. RG 1.174 defines very small changes 
in risk as resulting in increases of CDF [core damage frequency] 
below 10-\6\/yr and increases in LERF [large early 
release frequency] below 10-\7\/yr. Since the ILRT does 
not impact CDF, the relevant criterion is LERF. The increase in 
LERF, resulting from a change in the Type A ILRT test interval from 
a once-per-ten-years to a once-per-fifteen-years is 0.83 x 
10-\8\/yr, based on internal events. Since guidance in 
Reg. Guide 1.174 defines very small changes in LERF as below 
10-\7\/yr, increasing the ILRT interval from ten to 
fifteen years is, therefore, considered non-risk significant and 
will not significantly reduce the margin of safety. The NUREG-1493 
generic study of the effects of extending containment leakage 
testing found that a 20-year interval in Type A leakage testing 
resulted in an imperceptible increase in risk to the public. NUREG-
1493 generically concludes that the design containment leakage rate 
contributes about 0.1 percent of the overall risk. Decreasing the 
Type A testing frequency would have a minimal effect on this risk 
since 95% of the Type A detectable leakage paths would already be 
detected by Type B and C testing. Given that the proposed change 
will continue to meet the current design basis, any reduction in a 
margin of safety would not be significant.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 5238]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone 
Power Station, Unit No. 3, New London County, Connecticut

    Date of amendment request: December 16, 2004.
    Description of amendment request: The proposed amendment would 
revise the current fuel rod average licensing basis burnup limit for 
one lead test assembly (LTA) containing advanced zirconium based alloys 
to a limit not exceeding 71,000 MWD/MTU.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The Westinghouse LTA is very similar in design to the 
Westinghouse fuel that comprises the remainder of the core. The 
reload core design for Millstone Unit 3 Cycle 12, where one LTA will 
operate to high burnup, will meet all applicable design criteria. 
The performance of the Emergency Core Cooling System will not be 
affected by the operation of the LTA and operation of the LTA to 
high burnup will not result in a change to the Millstone Unit 3 
reload design and safety analysis limits. Operation of one 
Westinghouse LTA to high burnup will not result in a measurable 
impact on normal operating releases, and will not increase the 
predicted radiological consequences of accidents postulated in 
Chapter 15 of the Millstone FSAR [final safety analysis report]. 
Therefore, neither the probability of occurrence nor the 
consequences of any accident previously evaluated is significantly 
increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The Westinghouse LTA is very similar in design (both mechanical 
and composition of materials) to the resident Westinghouse fuel. All 
design and performance criteria will continue to be met and no new 
single failure mechanisms will be created. The irradiation of one 
LTA to high burnup does not involve any alteration to plant 
equipment or procedures, which would introduce any new or unique 
operational modes or accident precursors. Therefore, the possibility 
for a new or different kind of accident from any accident previously 
evaluated is not created.
    3. Involve a significant reduction in a margin of safety.
    The operation of one Westinghouse LTA to high burnup does not 
change the performance requirements of any system or component such 
that any design criteria will be exceeded. The normal limits on core 
operation defined in the Millstone Unit 3 Technical Specifications 
will remain applicable for the core in which the high burnup 
assembly is irradiated. Therefore, the margin of safety as defined 
in the Bases to the Millstone Unit 3 Technical Specifications is not 
significantly reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    NRC Section Chief: Darrell Roberts.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of amendment request: September 8, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the technical specifications (TSs) to maintain 
hydrogen recombiners and hydrogen monitors. Licensees were generally 
required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TSs for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 8, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The 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 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. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen 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.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).

[[Page 5239]]

    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen monitor requirements, 
including removal of these requirements from TSs, 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 monitor requirements, including removal 
of these requirements from TSs, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen 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 the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TSs, 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 Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TSs 
will not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT 
06385.
    NRC Section Chief: Darrell J. Roberts.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina; Docket Nos. 
50-269, 50-270, and 50-287, Oconee Nuclear Station, Units 1, 2, and 3, 
Oconee County, South Carolina

    Date of amendment request: September 20, 2004.
    Description of amendment request: The proposed amendment deletes 
the requirements from the technical specifications (TS) to maintain 
hydrogen recombiners (McGuire only) and hydrogen monitors (McGuire and 
Oconee). Licensees were generally required to implement upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI Unit 2. Requirements related to 
combustible gas control were imposed by Order for many facilities and 
were added to or included in the TS for nuclear power reactors 
currently licensed to operate. The revised 10 CFR 50.44, ``Standards 
for Combustible Gas Control System in Light-Water-Cooled Power 
Reactors,'' eliminated the requirements for hydrogen recombiners and 
relaxed safety classifications and licensee commitments to certain 
design and qualification criteria for hydrogen and oxygen monitors.
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration determination for referencing in 
license amendment applications in the Federal Register on September 25, 
2003 (68 FR 55416). The licensee affirmed the applicability of the 
model NSHC determination in its application dated September 20, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The 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 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. 
Category 1 in [Regulatory Guide] RG 1.97 is intended for key 
variables that most directly indicate the accomplishment of a safety 
function for design-basis accident events. The hydrogen 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.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from [Technical Specification] TS will not 
prevent an accident management strategy through the use of the 
severe accident management guidelines (SAMGs), the emergency plan 
(EP), the emergency operating procedures (EOP), and site survey 
monitoring that support modification of emergency plan protective 
action recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen 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.

[[Page 5240]]

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 monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen 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 the Margin of Safety

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen 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 Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: October 22, 2004.
    Description of amendment request: The proposed amendments would 
delete the requirements from the Technical Specifications (TSs) to 
maintain hydrogen recombiners and hydrogen monitors. Licensees were 
generally required to implement upgrades as described in NUREG-0737, 
``Clarification of TMI [Three Mile Island] Action Plan Requirements,'' 
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions 
During and Following an Accident.'' Implementation of these upgrades 
was an outcome of the lessons learned from the accident that occurred 
at TMI Unit 2. Requirements related to combustible gas control were 
imposed by Order for many facilities and were added to or included in 
the TSs for nuclear power reactors currently licensed to operate. The 
revised 10 CFR 50.44, ``Standards for Combustible Gas Control System in 
Light-Water-Cooled Power Reactors,'' eliminated the requirements for 
hydrogen recombiners and relaxed safety classifications and licensee 
commitments to certain design and qualification criteria for hydrogen 
and oxygen monitors.
    The Nuclear Regulatory Commission (NRC) staff issued a notice of 
availability of a model no significant hazards consideration (NSHC) 
determination for referencing in license amendment applications in the 
Federal Register on September 25, 2003 (68 FR 55416). The licensee 
affirmed the applicability of the model NSHC determination in its 
application dated October 22, 2004.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

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

    The 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 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. 
Category 1 in RG 1.97 is intended for key variables that most 
directly indicate the accomplishment of a safety function for 
design-basis accident events. The hydrogen 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.
    The regulatory requirements for the hydrogen monitors can be 
relaxed without degrading the plant emergency response. The 
emergency response, in this sense, refers to the methodologies used 
in ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy through the use of the severe accident 
management guidelines (SAMGs), the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of the hydrogen recombiner 
requirements and relaxation of the hydrogen 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 monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen 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

[[Page 5241]]

of radionuclides within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

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

    The elimination of the hydrogen recombiner requirements and 
relaxation of the hydrogen 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 Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    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 Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 25, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure 
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
in license amendment applications in the Federal Register on June 23, 
2004 (69 FR 35067). The licensee affirmed the applicability of the 
model NSHC determination in its application dated October 25, 2004.
    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 the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates the Technical Specifications 
(TSs) reporting requirements to provide a monthly operating report 
of shutdown experience and operating statistics if the equivalent 
data is submitted using an industry electronic database. It also 
eliminates the TS reporting requirement for an annual occupational 
radiation exposure report, which provides information beyond that 
specified in NRC 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 or transients. 
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 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?
    Response: No.
    This is an administrative change to reporting requirements of 
plant operating information and occupational radiation exposure 
data, and has no effect on plant equipment, operating practices or 
safety analyses assumptions. For these reasons, the proposed change 
does not involve a significant reduction in [a] margin of safety.

    Based upon the reasoning presented above, the requested change does 
not involve significance hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: December 30, 2004.
    Description of amendme