Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations, 29785-29808 [05-10063]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 70, Number 99 (Tuesday, May 24, 2005)]
[Notices]
[Pages 29785-29808]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 05-10063]


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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 April 29, 2005 through May 12, 2005. The 
last biweekly notice was published on May 10, 2005 (70 FR 24645).

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 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

[[Page 29786]]

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 transmitted either by 
means of facsimile transmission to (301) 415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the ADAMS Public 
Electronic Reading Room on the Internet at the NRC Web site, https://
www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or 
by e-mail to pdr@nrc.gov.

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

    Date of amendment request: January 21, 2005.
    Description of amendment request: The proposed amendment would 
implement the Alternative Source Term (AST) for the analysis of the 
radiological consequences of a design-basis Loss-of-Coolant Accident 
(LOCA). There are no changes proposed to the Operating License or 
Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

    Revision of the LOCA analysis to the Alternative Source Term 
methodology does not affect the design or operation of HBRSEP [H. B. 
Robinson Steam Electric Plant], Unit No. 2. Rather, once the 
occurrence of an accident has been postulated, the new source term 
is an input to evaluate the consequences of the postulated accident. 
The implementation of the Alternative Source Term has been evaluated 
in revisions to the LOCA dose analysis at HBRSEP, Unit No. 2. Based 
on the results of this analysis, it has been demonstrated that the 
dose consequences are within the regulatory guidance provided by the 
NRC. This guidance is presented in 10 CFR 50.67 and Regulatory Guide 
1.183.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.

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

    The proposed change does not affect plant structures, systems, 
or components. The proposed change is to an evaluation methodology 
and does not initiate design basis accidents.
    Thus, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

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

    The proposed change is associated with the implementation of a 
new licensing basis for HBRSEP, Unit No. 2. The new licensing basis 
implements an Alternative Source Term in accordance with 10 CFR 
50.67 and the associated Regulatory Guide 1.183. The results of the 
revised limiting design basis analysis are subject to revised 
acceptance criteria. The analysis has been performed using 
conservative methodologies in accordance with regulatory guidance or 
other methodologies approved by the NRC in prior plant-specific 
license amendments. The dose consequences are within the acceptance 
criteria found in the regulatory guidance associated with 
Alternative Source Terms.
    The proposed change continues to ensure that doses at the 
exclusion area and low population zone boundaries, as well as the 
control room, are within the corresponding regulatory limits. 
Specifically, the margin of safety for the radiological consequences 
of these accidents is considered to be that provided by meeting the 
applicable regulatory limits.
    Therefore, this 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: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.

[[Page 29787]]

    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: February 14, 2005.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements (SRs) for the station batteries as 
specified in Technical Specification (TS) SR 3.8.4.5, the battery 
service test, and TS SR 3.8.4.6, the battery performance test.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the Proposed Changes Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated?
    No. The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed surveillance changes will continue to ensure 
that the DC system is tested in a manner that will verify 
operability. Performance of the required system surveillances, in 
conjunction with the applicable operational and design requirements 
for the DC system, provide assurance that the system will be capable 
of performing the required design functions for accident mitigation 
and also that the system will perform in accordance with the 
functional requirements for the system as described in the Updated 
Final Safety Analysis Report for HBRSEP [H. B. Robinson Steam 
Electric Plant], Unit No. 2. This ensures that the rate of 
occurrence and consequences of analyzed accidents will not change. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the Proposed Changes Create the Possibility of a New or 
Different Kind of Accident From Any Previously Evaluated?
    No. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated. The 
proposed surveillance requirement changes will continue to ensure 
that the DC system is tested in a manner that will verify 
operability. No physical changes to the HBRSEP, Unit No. 2, systems, 
structures, or components are being implemented. There are no new or 
different accident initiators or sequences being created by the 
proposed Technical Specifications changes. Therefore, these changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Do the Proposed Changes Involve a Significant Reduction in 
the Margin of Safety?
    No. The proposed changes do not involve a significant reduction 
in the margin of safety. The proposed DC system surveillance 
requirement changes provide appropriate and applicable surveillances 
for the DC system. The proposed changes to surveillance requirements 
for the DC system will continue to ensure system operability. 
Therefore, these changes do not affect any margin of safety for 
HBRSEP, Unit No. 2.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

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

    Date of amendment request: March 3, 2005.
    Description of amendment request: The proposed amendment would 
revise the requirements of Technical Specification (TS) 5.6.5, ``Core 
Operating Limits Report (COLR).'' Specifically, the proposed change 
would add topical report EMF-2103(P)(A), ``Realistic Large Break LOCA 
[loss-of-coolant accident] Methodology for Pressurized Water 
Reactors,'' to the list of documents specified in TS 5.6.5. TS 5.6.5 
lists the approved methodologies that can be used to determine the core 
operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The Proposed Change Does Not Involve a Significant Increase 
in the Probability or Consequences of an Accident Previously 
Evaluated?
    The proposed methodology will be reviewed and approved by the 
NRC prior to its use for HBRSEP [H. B. Robinson Steam Electric 
Plant], Unit No. 2. Analyzed events are assumed to be initiated by 
the failure of plant structures, systems, or components. The 
determination of core operating limits in accordance with this new 
methodology will meet the limitations specified in the NRC safety 
evaluation of the new methodology. The topical report associated 
with the new methodology demonstrates that the integrity of the fuel 
will be maintained and that design requirements will continue to be 
met. The proposed change does not involve physical changes to any 
plant structure, system, or component. Therefore, the probability of 
occurrence for a previously analyzed accident is not significantly 
increased.
    The consequences of a previously analyzed accident are dependent 
on the initial conditions assumed for the analysis, the behavior of 
the fuel during the analyzed accident, the availability and 
successful functioning of the equipment assumed to operate in 
response to the analyzed event, and the setpoints at which these 
actions are initiated. The proposed methodology continues to meet 
applicable design and safety analyses acceptance criteria. The 
proposed change does not affect the performance of any equipment 
used to mitigate the consequences of an analyzed accident. As a 
result, no analysis assumptions are violated and there are no 
adverse effects on the factors that contribute to offsite or onsite 
dose as the result of an accident. The proposed change does not 
affect setpoints that initiate protective or mitigative actions. The 
proposed change ensures that plant structures, systems, or 
components are maintained consistent with the safety analysis and 
licensing bases. Based on this evaluation, there is no significant 
increase in the consequences of a previously analyzed event. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The Proposed Change Does Not Create the Possibility of a New 
or Different Kind of Accident From Any Previously Evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures, or components, other than allowing for 
fuel design in accordance with NRC approved methodologies. The 
proposed methodology continues to meet applicable criteria for Large 
Break Loss of Coolant Accident (LBLOCA) analysis. No new or 
different equipment is being installed. No installed equipment is 
being operated in a different manner. There is no alteration to the 
parameters within which the plant is normally operated or in the 
setpoints that initiate protective or mitigative actions. As a 
result, no new failure modes are being introduced. There are no 
changes in the methods governing normal plant operation, nor are the 
methods utilized to respond to plant transients altered. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The Proposed Change Does Not Involve a Significant Reduction 
in the Margin of Safety?
    The margin of safety is established through the design of the 
plant structures, systems, and components, through the parameters 
within which the plant is operated, through the establishment of the 
setpoints for the actuation of equipment relied upon to respond to 
an event, and through margins contained within the safety analyses. 
The proposed change in the methodology used for LBLOCA analyses does 
not impact the condition or performance of structures, systems, 
setpoints, and components relied upon for accident mitigation. The 
proposed

[[Page 29788]]

change does not significantly impact any safety analysis assumptions 
or results. Therefore, the proposed change does not result in a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David T. Conley, Associate General Counsel 
II--Legal Department, Progress Energy Service Company, LLC, Post Office 
Box 1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Michael L. Marshall, Jr.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423, Millstone Power Station, Unit Nos. 1, 2, and 3, New London County, 
Connecticut

    Date of amendment request: December 21, 2004.
    Description of amendment request: The requested change will delete 
Technical Specification (TS) requirements for annual Occupational 
Radiation Exposure Reports (all units), annual report regarding 
challenges to pressurizer relief and safety valves (Units 2 and 3), and 
Monthly Operating Reports (Units 2 and 3).
    The NRC staff issued a notice of availability of a model no 
significant hazards consideration (NSHC) determination for referencing 
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 21, 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 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: 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. 50-336, Millstone 
Power Station, Unit No. 2, New London County, Connecticut

    Date of amendment request: March 9, 2005.
    Description of amendment request: Current Technical Specifications 
(TSs) require that all operations involving a reduction in reactor 
coolant boron concentration or that involve positive reactivity changes 
be suspended under certain conditions. The requested changes modify the 
TSs to incorporate wording related to the reactor coolant system (RCS), 
electrical power systems, and refueling operations to provide 
operational flexibility during mode changes or addition of coolant 
during shutdown operations. Additionally, changes are to be made to the 
TS bases, as appropriate.
    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:

Criterion 1: Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed change does not in any way alter the SDM [shutdown 
margin] or refueling boron concentration. It limits introduction of 
coolant into the RCS of reactivity more positive than that necessary 
to meet the required SDM or refueling boron concentration. This 
proposed change does not affect the input or assumptions for any 
accidents previously evaluated nor does it affect initiation of an 
accident. Based on this discussion, the proposed amendment does not 
increase the probability or consequence of an accident previously 
evaluated.

Criterion 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 allows introduction of coolant into the RCS 
with different temperature or lower boron concentration, however, 
the required boron concentration or SDM is maintained. The proposed 
amendment does not introduce failure modes, accident initiators, or 
malfunctions that would cause a new or different kind of accident. 
No plant modifications are associated with the change. Therefore, 
the proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?

    Response: No.
    The proposed change provides the flexibility necessary for 
continued safe reactor operations while limiting any potential for 
excess positive reactivity additions. [The] SDM and required boron 
concentration are not affected. Therefore, based on the above, 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: Lillian M. Cuoco, Senior Nuclear Counsel, 
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
    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 23, 2004.
    Description of amendment request: The requested amendment would

[[Page 29789]]

relocate certain Technical Specifications regarding refueling 
operations to the Technical Requirements Manual (TRM).
    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:

Criterion 1: Does the proposed amendment involve a significant increase 
in the probability or consequences of an accident previously evaluated?

    Response: No.
    The communications equipment, refueling machine, and spent fuel 
pool crane are not designed to perform accident mitigation 
functions. The proposed change to relocate selected refueling 
specifications does not modify any plant equipment and does not 
impact any failure modes that could lead to an accident. Relocating 
the specifications to the TRM where changes would be controlled 
under the 10 CFR 50.59 process does not change the ability of the 
communications or refueling equipment to function as expected. 
Additionally, these specifications have no affect on the consequence 
of any analyzed accident since the equipment is not related to 
accident mitigation. Based on this discussion, the proposed 
amendment does not increase the probability or consequences of an 
accident previously evaluated.

Criterion 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[s] do[es] not modify any plant equipment and 
there is no impact on the capability of the existing equipment to 
perform their intended functions to move fuel safely or conduct 
refueling operations while in contact with the control room. No 
system setpoints are being modified and no changes are being made to 
the method in which refueling operations are conducted. No changes 
to the heavy loads program are being proposed by this change. No new 
failure modes are introduced by the proposed changes. The proposed 
amendment does not introduce accident initiators or malfunctions 
that would cause a new or different kind of accident. Therefore, the 
proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3: Does the proposed amendment involve a significant 
reduction in a margin of safety?

    Response: No.
    The relocation of Technical Specification 3/4.9.5, ``Refueling 
Operations, Communications,'' to the TRM does not imply any 
reduction in its importance in [e]nsuring communication between the 
control room and the refueling station. The proposed change will not 
alter the requirement on communication between the control room and 
the refueling station, it will not alter any of the assumptions used 
in the fuel handling accident analysis, nor will it cause any safety 
system parameters to exceed their acceptance limit. The relocation 
of Technical Specification 3/4.9.6, ``Refueling Machine'' to the TRM 
does not alter the requirement for the lifting device on the 
refueling machine to have adequate capacity or for the interlocks to 
be demonstrated operable prior to fuel movement. The assumptions 
used in the accident analysis are not impacted by this change and no 
impact to any safety system parameters will result. The relocation 
of Technical Specification 3/4.9.7, ``Crane Travel--Spent Fuel 
Storage Areas,'' to the TRM will not alter the requirement that the 
crane interlocks and/or physical stops are operable, nor will it 
alter any of the assumptions used in the fuel handling accident 
analysis. Heavy load lifts are administratively controlled by a safe 
load path and crane interlocks. The proposed change[s] do[es] not 
modify any heavy load path criteria. Administrative changes 
associated with the proposed revision such as relocation of 
associated Technical Specification Bases to the TRM will not have an 
impact on any established safety margins.
    The proposed change[s] do[es] not affect any of the assumptions 
used in the accident analysis, nor do they affect any operability 
requirements for equipment important to plant safety. Therefore, the 
proposed change[s] will not result in a significant reduction in the 
margin of safety as defined in the Bases for Technical 
Specifications covered in this License Amendment Request.

    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 J. Roberts.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: November 25, 2002, as supplemented by 
letters dated November 13, and December 16, 2003, September 22, 2004, 
and April 6, 2005.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) for the Ventilation Filter Testing 
Program (VFTP), Annulus Ventilation System (AVS), Auxiliary Building 
Filtered Ventilation Exhaust System (ABFVES), Fuel Handling Ventilation 
Exhaust System (FHVES), and Control Room Area Ventilation System 
(CRAVS), and containment penetrations.
    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:

First Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant increase in the probability or 
consequences of an accident previously evaluated? No.
    This license amendment request proposes amendments to the system 
TS and/or Bases and/or VFTP TS requirements for the AVS, ABFVES, 
FHVES, and CRAVS. It also proposes amendments to the TS and Bases 
for Containment Penetrations. The AVS is in standby during normal 
plant operations and operates only following a Safety Injection 
signal or during a test. It is not an accident initiator. The ABFVES 
is in operation during normal plant operations. However, the ABFVES 
is not used in direct support of any phase of power generation or 
conversion or transmission, shutdown cooling, fuel handling 
operations, or processing of radioactive fluids. Therefore, it is 
not an accident initiator. The FHVES is utilized to support fuel 
handling operations when moving recently irradiated fuel. It is not 
an accident initiator. The CRAVS operates during normal plant 
operations. However, it is not an accident initiator (the CRAVS 
being defined so as to exclude equipment that maintains an 
appropriately low temperature in the control room). The status of 
containment penetrations is required to be controlled so as to 
minimize the consequences of a fuel handling accident or a weir gate 
drop accident. The containment penetrations by themselves are not 
accident initiators. No accident initiators are associated with the 
changes proposed in this license amendment request. For these 
reasons, operation of the facility in accordance with this proposed 
amendment does not involve a significant increase in the probability 
of any accident previously evaluated.
    In support of the proposed amendment, an analysis has been 
performed to determine the radiological consequences of the design 
basis [Loss of Coolant Accident] LOCA at Catawba Nuclear Station. 
The analysis made use of the Alternative Source Term (AST) 
methodology and in general conformed to the regulatory positions of 
Regulatory Guide 1.183 and the draft regulatory positions of DG-
1111. Total Effective Dose Equivalent (TEDE) radiation doses at the 
Exclusion Area Boundary (EAB), boundary of the Low Population Zone 
(LPZ), and to the control room operators were calculated and found 
to be acceptable. TEDEs were calculated for a design basis LOCA 
postulated for a Catawba nuclear unit operating with all low 
enriched uranium (LEU) fuel and with 4 mixed oxide (MOX) lead fuel 
assemblies (LFAs). It was found that insertion of 4 MOX LFAs did not 
produce a significant increase in the TEDEs for a design basis LOCA.
* * * * *
    The new value for the control room TEDE radiation dose is higher 
than the TEDE radiation dose equivalent to the radiation

[[Page 29790]]

doses currently reported in the UFSAR. However, the limiting control 
room TEDE radiation dose reported in this submittal is lower than 
the acceptance criterion * * * The new LPZ TEDE radiation dose is 
higher than the equivalent TEDE radiation dose currently 
represented. On the other hand, the margin to the acceptance 
criterion is [large] * * *. The TEDE radiation doses newly computed 
at the EAB for the design basis LOCA are lower than the 
corresponding equivalent EAB TEDE radiation dose currently 
represented in the UFSAR. The margin in the EAB TEDE radiation dose 
to the guideline value is [also large]. * * * In all cases, there is 
significant margin between the newly calculated post-LOCA TEDE 
radiation doses and the corresponding regulatory guideline values. 
In the sense that the margins to the germane regulatory guideline 
values are still large, the new values of TEDE radiation doses are 
comparable to the equivalent TEDE associated with the post-LOCA 
radiation doses currently listed in the UFSAR. Furthermore, these 
margins for the design basis LOCA do not significantly decrease with 
insertion of the 4 MOX LFAs. Therefore, the proposed amendment is 
determined to not result in a significant increase in accident 
consequences.
    AST analyses also were completed for the design basis locked 
rotor accident (LRA) and rod ejection accident (REA). Again, these 
design basis accidents were postulated to occur at a Catawba nuclear 
unit operating with either an all LEU core or with 4 MOX LFAs. The 
TEDEs following these design basis accidents were compared to the 
equivalent TEDEs associated with the current license basis analyses. 
The equivalent TEDEs were computed from the post-accident whole body 
and thyroid radiation doses using the method prescribed in 
Regulatory Guide 1.183, as noted above. TEDEs only at offsite 
locations were compared as post-accident control room radiation 
doses are not reported for these design basis accidents in the 
Catawba UFSAR.
* * * * * * *
    For the EAB, LPZ, and control room, the post-LRA TEDEs are seen 
to increase from the values equivalent to the radiation doses from 
the current license basis analyses. (This is attributed primarily to 
the increase in assumed fraction of the fuel pins with clad failure 
following a design basis LRA at Unit 2. * * *) However, the margins 
to the acceptance criteria of 2.5 Rem at the offsite locations and 5 
Rem in the control room are still significant.
* * * * * * *
    For the EAB, LPZ, and control room, the post-REA TEDEs are seen 
to increase from the values equivalent to the radiation doses from 
the current license basis analyses, as they did for the design basis 
LRA. (This is attributed to a number of reasons. These include 
increase in the fraction of gap activity released to containment, 
inclusion of limiting radial peaking in the source term, and 
inclusion of alkali metals.) However, the margins to the acceptance 
criteria of 6.3 Rem at the offsite locations and 5 Rem in the 
control room are still significant * * *.
    The changes proposed to the TS for Containment Penetrations are 
editorial in nature and will have no effect upon accident 
consequences.
    The changes proposed to the VFTP TS for the AVS, ABFVES, and 
FHVES will not result in a significant increase in any accident 
consequences. The changes to make the penetration values for Unit 2 
consistent with Unit 1 for the AVS, ABFVES, and FHVES are acceptable 
because the appropriate safety factors as delineated in the 
applicable regulatory guideline documents are still maintained. The 
change to the flowrate specified for the ABFVES is consistent with 
the design basis operation of this system. Also, the editorial 
changes proposed to the VFTP TS will have no impact on any 
accidents.
    Operation of the facility in accordance with the proposed 
amendment does not involve a significant increase in the 
consequences of an accident previously evaluated.

Second Standard

    Does operation of the facility in accordance with the proposed 
amendment create the possibility of a new or different kind of 
accident from any accident previously evaluated? No.
    This proposed amendment does not involve addition, removal, or 
modification of any plant system, structure, or component. These 
changes will not affect the operation of any plant system, 
structure, or components as directed in plant procedures.
    The analysis performed in support of this license amendment 
request, together with the analyses of the design basis fuel 
handling accident and weir gate drop reported in previously 
submitted and NRC approved license amendment requests, includes full 
scope implementation of AST methodology. This analysis does not 
represent any change in the post-accident operation of any plant 
system, structure, or component.
    Operation of the facility in accordance with this amendment does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.

Third Standard

    Does operation of the facility in accordance with the proposed 
amendment involve a significant reduction in the margin of safety? 
No.
    Margin of safety is related to confidence in the ability of 
fission product barriers to perform their design functions following 
any of their design basis accidents. These barriers include the fuel 
cladding, the Reactor Coolant System, and the containment. The 
performance of these barriers either during normal plant operations 
or following an accident will not be affected by the changes 
associated with the license amendment request.
    The AVS is associated with the containment fission product 
barrier. Its post-accident operation will not be affected by 
implementation of the amendment to its TS. The operation of the 
ABFVES either during normal plant operations or following an 
accident will not be affected by implementation of the amendment to 
its TS. The operation of the FHVES either during normal plant 
operations or following an accident will not be affected by 
implementation of the amendment to its TS. The operation of the 
CRAVS either during normal plant operations or following an accident 
will not be adversely affected by the proposed changes to its TS 
Bases. The operation of Containment Penetrations following an 
accident will not be adversely affected by the proposed change to 
its TS.
    As noted, an analysis of radiological consequences of the design 
basis LOCA at Catawba Nuclear Station has been performed in support 
of this license amendment request. The design basis LOCA scenarios 
were selected based on extensive evaluations of Catawba, its design 
basis, and its anticipated response to a design basis LOCA. Credit 
was taken only for safety related systems, structures, and 
components in simulating the mitigation of radiological consequences 
of the LOCA. Limiting values were taken for performance 
characteristics of the Class 1E systems modeled in the analysis. The 
radiological consequences (TEDE radiation doses at the EAB, LPZ, and 
in the control room) are within the regulatory guideline values with 
significant margin.
    The changes proposed to the VFTP TS for the AVS, ABFVES, and 
FHVES will not result in a significant reduction in the margin of 
safety. These changes are supported by regulatory guidance 
documents, and are consistent with existing system operation. Also, 
the editorial changes proposed to the VFTP TS will not have any 
impact on safety.
    Operation of the facility in accordance with the proposed 
amendment does not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas
    Date of amendment request: September 30, 2004, as supplemented by 
letter dated April 26, 2005.
    Description of amendment request: The proposed amendment would 
change the existing steam generator (SG) tube surveillance program to 
be consistent with that being proposed by the Technical Specifications 
Task Force (TSTF) in TSTF-449. These proposed changes would revise 
Technical Specification (TS) 1.1 on definitions, TS 3.4.13 on reactor 
coolant system

[[Page 29791]]

operational leakage, TS 5.5.9 on SG program, and TS 5.6.7 on SG tube 
inspection reports, and add a new TS 3.4.16 on SG tube integrity. Also, 
as a result of the licensee replacing the SGs with SGs having a new 
Alloy 690 thermally treated tubing design, the TSs would be revised to 
reflect this replacement. The September 30, 2004, application was 
noticed in the Federal Register on November 9, 2004 (69 FR 64987).
    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 requires a Steam Generator Program that 
includes performance criteria that will provide reasonable assurance 
that the steam generator (SG) tubing will retain integrity over the 
full range of design basis operating conditions (including startup, 
power operation, hot standby, cooldown, anticipated transients and 
postulated accidents). The SG performance criteria are based on tube 
structural integrity, accident induced leakage, and operational 
LEAKAGE. These criteria assure that the probability of an accident 
will not be increased.
    The primary to secondary accident induced leakage rate for any 
design basis accidents, other than an SG tube rupture, shall not 
exceed the leakage rate assumed in the accident analysis in terms of 
total leakage rate for all SGs and leakage rate for an individual 
SG. [The primary to secondary accident induced leakage rate is 
relatively inconsequential for the SG tube rupture analysis.] The 
operational LEAKAGE performance criterion meets current NRC 
regulations and NEI [Nuclear Energy Institute] 97-06 criteria for 
reactor coolant system (RCS) operational primary to secondary 
LEAKAGE through any one SG of 150 gallons per day. These criteria 
assure that accident doses will stay within regulatory and licensing 
basis limits.
    Therefore, the proposed change does not affect the probability 
or consequences of any ANO-1 [Arkansas Nuclear One, Unit 1] analyzed 
accidents.
    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 performance based requirements are an improvement 
over the requirements imposed by the current technical 
specifications. Implementation of the proposed Steam Generator 
Program will not introduce any adverse changes to the plant design 
basis or postulated accidents resulting from potential tube 
degradation. The proposed change does not affect the design of the 
SGs, their method of operation, or primary or secondary coolant 
chemistry controls. The proposed 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.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    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 Steam Generator Program to manage SG 
tube inspection, assessment, repair, and plugging. The requirements 
established by the Steam Generator Program are consistent with those 
in the applicable design codes and standards and are an improvement 
over the requirements in the current technical specifications.
    Therefore, the margin of safety is not changed by the proposed 
change to the ANO-1 TSs.

    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: March 30, 2005.
    Description of amendment request: The proposed amendment adopts the 
following Nuclear Regulatory Commission (NRC) approved Technical 
Specification Task Force (TSTF) changes that affect the Boiling Water 
Reactor (BWR)/6 Improved Standard Technical Specifications:

--------------------------------------------------------------------------------------------------------------------------------------------------------
            TSTF No.                                 Description                            TS section affected                  Type of change
--------------------------------------------------------------------------------------------------------------------------------------------------------
046, Rev. 1.....................  Clarify the Containment Isolation Valve            SR 3.6.1.3.4....................  Administrative.
                                   Surveillance Requirement (SR) to apply only to    SR 3.6.4.2.2....................
                                   automatic isolation valves.                       SR 3.6.5.3.3....................
222, Rev. 1.....................  Control Rod Scram Time Testing...................  SR 3.1.4.1......................  Administrative.
                                                                                     SR 3.1.4.4......................
264, Rev........................  Delete flux monitors specific overlap SRs........  SR 3.3.1.1.5....................  Less Restrictive.
                                                                                     SR 3.3.1.1.6....................
                                                                                     Table 3.3.1.1-1.................
275, Rev. 0.....................  Clarify requirements for Diesel Generator (DG)     Table 3.3.5.1-1, Footnote (a)...  Administrative.
                                   start signal on Reactor Pressure Vessel (RPV)
                                   level--low, low, low during RPV cavity flood-up.
276, Rev. 2.....................  Revise DG full load rejection test...............  SR 3.8.1.9......................  Less Restrictive.
                                                                                     SR 3.8.1.10.....................
                                                                                     SR 3.8.1.14.....................
300, Rev. 0.....................  Eliminate DG loss of coolant accident-Start SRs    SR 3.8.2.1......................  Less Restrictive.
                                   while in shutdown when emergency core cooling
                                   system is not required.
322, Rev. 2.....................  Secondary Containment Integrity SRs..............  SR 3.6.4.1.3....................  Administrative.
                                                                                     SR 3.6.4.1.4....................
400, Rev. 1.....................  Clarification of SR on bypass of DG automatic      SR 3.8.1.13.....................  Administrative.
                                   trips.
416, Rev. 0.....................  SR 3.5.1.2 Notation..............................  LCO 3.5.1.......................  Administrative.
                                                                                     SR 3.5.1.2......................
                                                                                     LCO 3.5.2.......................
                                                                                     SR 3.5.2.4......................
--------------------------------------------------------------------------------------------------------------------------------------------------------


[[Page 29792]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

    Response: No.
    The proposed changes to the TS [Technical Specifications] 
involve both administrative and less restrictive changes. The 
administrative changes involve wording changes that clarify 
requirements without changing the original intent. As such, these 
types of changes do not affect initiators of analyzed events and do 
not affect the mitigation of any accidents or transients.
    The less restrictive changes involve modifications to 
Surveillance Requirements. The modified Surveillance Requirements do 
not cause the plant to be operated in a new or different manner and 
the required equipment continues to be tested in a manner and at a 
frequency necessary to provide confidence that the equipment can 
perform its intended safety function. Consequently, no initiators to 
accidents previously evaluated are affected and no mitigating 
equipment assumed in accidents previously evaluated is adversely 
affected.
    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 changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed), do 
not change the design function of any equipment, and do not change 
the methods of normal plant operation. Accordingly, the proposed 
changes do not create any new credible failure mechanisms, 
malfunctions, or accident initiators not previously considered in 
the GGNS [Grand Gulf Nuclear Station] design and licensing basis.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

3. Does the proposed change involve a significant reduction in a margin 
of safety?

    Response: No.
    The proposed changes have no affect on any safety analysis 
assumptions or methods of performing safety analyses. The changes do 
not adversely affect system OPERABILITY or design requirements and 
the equipment continues to be tested in a manner and at a frequency 
necessary to provide confidence that the equipment can perform its 
intended safety functions. 10 CFR 50.36 (c)(3) requires the TS to 
include Surveillance Requirements relating to test, calibration, or 
inspection to assure that the necessary quality of systems and 
components is maintained, that facility operation will be within 
safety limits, and that the limiting conditions for operation will 
be met. The GGNS TS Surveillance Requirements will continue to 
provide this assurance with the proposed adoption of the NRC 
approved TSTF changes.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Allen G. Howe.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of amendment request: December 14, 2004.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.3.G, ``Scram Discharge Volume 
[SDV],'' to allow vent or drain lines with inoperable valves to be 
isolated instead of requiring the valves to be restored to Operable 
status or to be in Hot Shutdown within 12 hours.
    The NRC staff issued a Notice of Opportunity for Comment in the 
Federal Register on February 24, 2003 (68 FR 8637), on possible 
amendments to revise the action for one or more SDV vent or drain lines 
with an inoperable valve, including a model safety evaluation and model 
no significant hazards consideration (NSHC) determination, using the 
consolidated line-item improvement process. The NRC staff subsequently 
issued a Notice of Availability of the models for referencing license 
amendment applications in the Federal Register on April 15, 2003 (68 FR 
18294). The licensee affirmed the applicability of the model NSHC 
determination (modified slightly as a result of the Pilgrim Nuclear 
Power Station TS format) in its application dated December 14, 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.

    A change is proposed to allow the affected SDV vent and drain 
line to be isolated when there are one or more SDV vent or drain 
lines with vent or drain valves inoperable instead of requiring the 
valves to be restored to operable status or be in Hot Shutdown 
within 12 hours. With one SDV vent or drain valve inoperable in one 
or more lines, the isolation function would be maintained since the 
redundant valve in the affected line would perform its safety 
function of isolating the SDV. Following the completion of the 
required action, the isolation function is fulfilled since the 
associated line is isolated. The ability to vent and drain the SDV 
is maintained and controlled through administrative controls. This 
requirement assures the reactor protection system is not adversely 
affected by the inoperable valves. With the safety functions of the 
valves being maintained, the probability or consequences of an 
accident previously evaluated are not significantly increased.

Criterion 2: The proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

    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 previously evaluated.

Criterion 3: The proposed change does not involve a significant 
reduction in [a] margin of safety.

    The proposed change ensures that the safety functions of the SDV 
vent and drain valves are fulfilled. The isolation function is 
maintained by redundant valves and by the required action to isolate 
the affected line. The ability to vent and drain the SDV is 
maintained through administrative controls. In addition, the reactor 
protection system will prevent filling of the SDV to the point that 
it has insufficient volume to accept a full scram. Maintaining the 
safety functions related to isolation of the SDV and insertion of 
control rods ensures that the proposed change does not involve a 
significant reduction in the margin of safety.

    Based on the reasoning presented above, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: Darrell J. Roberts.

[[Page 29793]]

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: January 21, 2005.
    Description of amendment reques