[Federal Register Volume 69, Number 101 (Tuesday, May 25, 2004)]
[Notices]
[Pages 29761-29772]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-11507]
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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
[[Page 29762]]
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 30, through May 13, 2004. The last
biweekly notice was published on May 11, 2004 (69 FR 26184).
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, http://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
[[Page 29763]]
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, [email protected];
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
[email protected]. 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 Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://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
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
Duke Energy Corporation, et al., Docket Nos. 50-369 and 50-370, McGuire
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina
Docket Nos. 50-413 and 50-414, Catawba Nuclear Station, Units 1 and 2,
York County, South Carolina
Date of amendment request: March 23, 2004.
Description of amendment request: The amendments would revise
Technical Specification 5.5.7, ``Reactor Coolant Pump Flywheel
Inspection Program,'' to extend the allowable inspection interval to 20
years.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 24, 2003 (68 FR 37590), on possible amendments
to extend the inspection interval for reactor coolant pump (RCP)
flywheels, 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 in license
amendment applications in the Federal Register on October 22, 2003 (68
FR 60422). The licensee affirmed the applicability of the model NSHC
determination in its application dated March 23, 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 proposed change to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines [contained] in Regulatory Guide (RG) 1.174 (<1.0E-6 per
year). Moreover, considering the uncertainties involved in this
evaluation, the risk associated with the postulated failure of an
RCP motor flywheel is significantly low. Even if all four RCP motor
flywheels are considered in the bounding plant configuration case,
the risk is still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits; or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
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 in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or alter the methods governing normal plant operation. In
addition, the change does not impose any new or different
requirements or eliminate any existing requirements, and does not
alter any assumptions made in the safety analysis. The proposed
change is consistent with the safety analysis assumptions and
current plant operating practice. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in a margin of safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
[[Page 29764]]
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel. Therefore,
the proposed change does not involve a significant reduction in a
margin of safety.
The NRC staff proposes to determine that the amendment request
involves NSHC.
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: Stephanie M. Coffin, Acting.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: April 19, 2004.
Description of amendment request: The proposed change revises
Limiting Condition for Operation (LCO) 3.7.3, ``Control Room Emergency
Filtration System,'' to provide specific conditions and required
actions that address degraded control room boundary.
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 Technical Specifications (TS) change involves the
Control Room Emergency Filtration (CREF) System and associated
control room boundary, which provide a radiological controlled
environment from which the plant can be operated following a design
basis accident (DBA). The CREF system and the control room boundary
are not assumed to be initiators of any analyzed accident and do not
affect the probability of accidents. The proposed change adds a Note
to LCO 3.7.3 that allows the control room boundary to be opened
intermittently under administrative controls. A new Condition B is
also added to LCO 3.7.3 to specify a Completion Time of 24 hours to
restore an inoperable control room boundary to OPERABLE status
before requiring the plant to perform an orderly shutdown. The 24-
hour Completion Time is reasonable based on the low probability of a
DBA occurring during this time period and Energy Northwest's
commitment to implement, via administrative controls, appropriate
compensatory measures consistent with the intent of 10 CFR 50,
Appendix A, General Design Criteria (GDC) 19. These compensatory
measures will serve to minimize the consequences of an open control
room boundary and ensure the CREF system can continue to perform its
function. As such, these changes will not affect the function or
operation of any other systems, structures or components. Therefore,
the proposed TS 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 accident previously
evaluated.
The proposed change adds a Note to LCO 3.7.3 that allows the
control room boundary to be opened intermittently under
administrative controls. A new Condition B is also added to LCO
3.7.3 to specify a Completion Time of 24 hours to restore an
inoperable control room boundary to OPERABLE status before requiring
the plant to perform an orderly shutdown. The CREF system and the
control room boundary are designed to protect the habitability of
the control room. The CREF system and the control room boundary are
not accident initiators and do not affect the probability of
accidents. This change is administrative in nature and does not
involve any physical changes to the plant. Therefore, the proposed
TS change does not create the possibility of a new or different kind
of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change adds a Note to LCO 3.7.3 that allows the
control room boundary to be opened intermittently under
administrative controls. A new Condition B is also added to LCO
3.7.3 to specify a Completion Time of 24 hours to restore an
inoperable control room boundary to OPERABLE status before requiring
the plant to perform an orderly shutdown. The 24-hour Completion
Time is reasonable based on the low probability of a DBA occurring
during this time period and Energy Northwest's commitment to
implement, via administrative controls, appropriate compensatory
measures consistent with the intent of 10 CFR 50, Appendix A, GDC
19. These compensatory measures will serve to minimize the
consequences of an open control room boundary and assure that the
CREF system can continue to perform its function. Therefore, the
proposed TS change does not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station (RBS), Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 21, 2003, as supplemented
February 10, 2004.
Description of amendment request: The amendment would modify the
Technical Specifications (TSs) to delete TS 3.6.4.4, ``Shield Building
Annulus Mixing System,'' in its entirety, revise the Main Steam
Isolation Valve (MSIV) leakage limits contained within TS Surveillance
Requirement 3.6.1.3.10, and delete reference to TS 3.6.4.4 within TS
3.10.1, ``Inservice Leak and Hydrostatic Testing Operation.''
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.
As discussed above, the proposed changes are to delete the
annulus mixing function and deletion of the single MSIV leakage rate
limit. A review of the safety analysis report indicates that
operation (or mis-operation) of the annulus mixing system, or any
component of the annulus mixing system is not considered an
initiator of any accident evaluated in the Updated Safety Analysis
Report. The deletion of the single MSIV leakage limit of 50 scfh in
effect establishes a maximum leakage limit of 150 scfh which is the
current total MSIV leakage limit. The elimination of the single MSIV
acceptable leakage rate limit does not impact any event initiator.
As the proposed changes do not involve any accident initiators,
there is no increase in the probability of an accident previously
evaluated.
The annulus mixing system and the main steam isolation valves
operate following an LOCA [loss-of-coolant accident] to mitigate the
consequences of an accident. Elimination of the annulus mixing
system and the single MSIV leakage limit will lead to some increase
in the dose consequences of a LOCA. The current LOCA dose
consequences evaluation for RBS was revised to account for the
elimination of the annulus mixing system and for increasing the
single MSIV leakage to 150 scfh (applying the total MS-PLCS Division
limit to the single MSIV). The results of the revised evaluation
with the proposed changes show an increase in the calculated dose
consequences, however, the calculated doses were still within the
acceptance limits of 10 CFR 50.67. Thus, while there is an increase
in the dose consequences of an accident previously identified, the
increase is not deemed to be significant.
Therefore, the proposed change does not involve a significant
increase in the
[[Page 29765]]
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 add any equipment, nor is any
equipment replaced with equipment with different performance
characteristics. Thus, no new initiators are added, and therefore,
no new accident types are created as a result of this change. The
proposed changes affect performance characteristics assumed in the
LOCA dose consequences evaluation, however, the nature of the
accidents evaluated in the safety analysis report are not changed.
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.
With respect to dose consequences for the LOCA event, the margin
of safety is considered to be that provided by meeting the 10 CFR
50.67 limits. The revised dose consequences evaluation, which
includes the proposed changes, continues to demonstrate that the
doses at the exclusion area boundary, the low population zone, and
the control room are within the acceptance limits in 10 CFR 50.67.
Therefore, there is no reduction in the margin of safety.
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: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station (RBS), Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: February 16, 2004.
Description of amendment request: The amendment would change
Technical Specification (TS) 3.6.5.1.3, regarding drywell bypass
leakage testing (DWBT). The change would allow for a one-time extension
of the interval (from 10 to 15 years) for performance of the next DWBT.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment to TS SR 3.6.5.1.3 adds a one-time
extension to the current interval for the DWBT. The current interval
of ten years, based on past performance, would be extended on a one-
time basis to 15-years from the date of the last test. The proposed
extension to the DWBT cannot increase the probability of an accident
since there are no design or operating changes involved and the test
is not an accident initiator. The proposed extension of the test
interval does not involve a significant increase in the consequences
since analysis has shown that, the proposed extension of the DWBT
frequency has a minimal impact on plant risk. 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 extension to the interval for the DWBT does not
involve any design or operational changes that could lead to a new
or different kind of accident from any accidents previously
evaluated. The tests are not being modified, but are only being
performed after a longer interval. 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. 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.
An evaluation of extending the DWBT surveillance frequency from
once in 10 years to once in 15 years has been performed using
methodologies based on the ILRT [integrated leak rate testing]
methodologies. This evaluation assumed that the DWBT frequency was
being adjusted in conjunction with the ILRT frequency. This analysis
used realistic, but still conservative, assumptions with regard to
developing the frequency of leakage classes associated with the
DWBT. The results from this conservative analysis indicates that the
proposed extension of the DWBT frequency has a minimal impact on
plant risk and 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: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, (Waterford 3) St. Charles Parish, Louisiana
Date of amendment request: May 7, 2004.
Description of amendment request: The proposed changes will revise
the Waterford 3 Technical Specifications (TS) to clarify the actions of
TS 3.4.5.1, Reactor Coolant System (RCS) Leakage; some of the
surveillance requirements (SRs) of TS 3.4.5.2, RCS Operational Leakage;
and delete duplication in TS 3.3.3.1, Radiation Monitoring
Instrumentation. The proposed change is based on NUREG-1432, ``Standard
Technical Specifications Combustion Engineering Plants,'' Revision 2,
dated April 30, 2001. Also, the proposed change will delete the
containment atmosphere gaseous radioactivity monitoring system from the
TS because this monitor does not meet the requirements of Regulatory
Guide 1.45, Revision 0, ``Reactor Coolant Pressure Boundary Leakage
Detection Systems,'' and Title 10 of the Code of Federal Regulations
(10 CFR), Part 50, Appendix A, General Design Criteria 30, ``Quality of
Reactor Coolant System Pressure Boundary.''
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 revisions do not involve any physical change to
plant design. The less restrictive changes proposed in this
amendment request include relocation of information to the UFSAR
[updated final safety analysis report], addition of a TS 3.0.4
exception, utilization of the diversity and redundancy of the
Waterford 3 leakage detection instrumentation, allowing diversity in
the contingency actions, deletion of SRs, and addition of an allowed
outage time when two of three required leakage detection
instrumentation is inoperable. The less restrictive changes will not
affect the capability of Waterford 3 to detect RCS leakage. At least
one RCS leakage detection instrumentation is always required to
remain operable, and other leakage detection indication, while not
credited specifically for RCS leakage detection, is still available
and required to be operable per other TS
[[Page 29766]]
requirements (i.e., Containment Temperature and Containment
Pressure). Also contingency actions are required (i.e., RCS
Inventory Balance, containment grab samples, flow switch
verification) when any of the RCS leakage detection instrumentation
is inoperable. Performance of the RCS inventory balance is the most
accurate method of determining and quantifying leakage. The RCS
inventory balance is being added as a contingency and replacement
for monitoring instrumentation that has continuous indication and
alarms in the control room.
The more restrictive changes proposed by this revision do not
adversely affect the capability of Waterford 3 RCS leakage detection
instrumentation to detect RCS leakage. The deletion of the
containment atmosphere gaseous radioactivity monitor is considered a
more restrictive change. This monitor does not meet the leakage
detection requirements of Regulatory Guide 1.45 and does not meet
the requirements for retention specified in 10 CFR 50.36. Deletion
of this monitor will reduce the diversity of the Waterford 3
instrumentation for monitoring the containment atmosphere and
require the plant to enter an Action statement when the containment
atmosphere particulate monitor is inoperable. Requiring performance
of an RCS inventory balance when the containment sump monitor is
inoperable provides contingency actions when the plant is in a
degraded RCS leakage detection condition.
The administrative changes proposed by this revision do not
adversely affect the capability of Waterford 3 RCS leakage detection
instrumentation to detect RCS leakage. Relocating the requirements
associated with the RCS Leak Detection System from various TS to
Specification 3.4.5.1 and adding requirement to shutdown when all
required RCS leakage detection instrumentation are inoperable are
administrative in nature. The relocation of information from one TS
to another consolidates information and causes less contusion in the
control room by having all requirements for the leakage detection
instrumentation in one TS. The addition of a specific action to
shutdown when all three leakage detection instrumentation are
inoperable versus an implied requirement to enter TS 3.0.3 is being
performed to be similar to the STS [Standard Technical
Specifications].
None of the above less restrictive, more restrictive, or
administrative changes affects the accident analyses. Since the
proposed changes only affect the requirements for the detection of
RCS leakage, the probability that an accident previously evaluated
will occur remains unchanged. The proposed changes do not prevent
nor limit the diversity of acceptable detection of RCS leakage.
These changes also do not affect the mitigation capability of any
accident previously evaluated. The consequences of an accident
previously evaluated are not affected since the mitigation of
previously evaluated accidents is not affected and leak rate
information will remain available to station personnel.
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 aforementioned revisions do not involve any physical change
to plant design. None of the proposed changes affect[s] the accident
analyses. The RCS water inventory balance is more accurate than
normal leak detection methods in regard to actual RCS leak rates,
and therefore is an excellent alternative when other leak detection
components may become inoperable. The proposed changes do not
prevent acceptable detection of RCS leakage by diverse methods. The
detection of a RCS leak can not cause an accident. Likewise,
detecting a RCS leak, while in its beginning stages, does not create
the possibility of a new or different kind of accident than any
previously analyzed. Therefore, a new or different kind of accident
than that previously analyzed does not result due to the proposed
changes of this submittal.
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 aforementioned revisions do not involve any physical change
to plant design. The proposed changes do not adversely affect the
ability of the RCS leakage detection system to detect RCS leakage.
The ability of the RCS leakage detection instrumentation to detect
leakage within the requirements of Regulatory Guide 1.45 is actually
improved. The containment atmosphere gaseous monitor is being
deleted from TS, because, it does not meet the requirements of
Regulatory Guide 1.45 to detect a 1.0 gpm [gallon per minute] RCS
leakage within 1 hour. Extending the AOT [allowed outage time] when
two of three leakage detection systems is inoperable does not
decrease the margin of safety because one instrument remains
operable, other instrumentation capable of indicating RCS leakage is
available, and an RCS inventory balance is required to be performed
on an increased frequency. The RCS inventory balance is more
accurate than normal leak detection methods in regard to actual RCS
leak rates, and therefore is an excellent alternative when other
Ieak detection components may become inoperable. Maintaining diverse
and accurate RCS leak detection methods available and capable of
prompt leakage detection helps to ensure RCS leaks will be detected
within an acceptable period of time and, therefore, the proposed
changes do not significantly reduce the margin to safety.
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: N. S. Reynolds, Esquire, Winston & Strawn
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: April 29, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3/4.4.10, ``Reactor Coolant
System--Structural Integrity, ASME Code Class 1, 2, and 3 Components,''
to relocate Surveillance Requirement (SR) 4.4.10.1.b which requires
that the reactor vessel internals vent valves be tested and inspected,
to the Technical Requirements Manual (TRM). The Davis-Besse Nuclear
Power Station (DBNPS) TRM is a licensee-controlled document that is
incorporated by reference into the DBNPS Updated Safety Analysis Report
(USAR). Changes to the DBNPS TRM are performed in accordance with the
regulatory requirements of 10 CFR 50.59.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their 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 surveillance requirement relocation
from the Technical Specifications to the USAR TRM does not alter the
design, operation, or testing of any structure, system, or
component. No preciously analyzed accident scenario is changed.
Initiating conditions and assumptions remain as previously analyzed.
Therefore, the proposed changes 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 surveillance requirement relocation
from the Technical Specifications to the USAR TRM does not alter the
design, operation, or testing of any structure, system or component.
The proposed change does not introduce any new or different accident
initiators. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
[[Page 29767]]
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No. The proposed surveillance requirements relocation
from the Technical Specifications to the USAR TRM does not affect
the capabilities of the Reactor Vessel Internals Vent Valves.
Therefore, the proposed change will not affect 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: May 3, 2004.
Description of amendment request: The proposed amendment would
change the facility as described in the Updated Safety Analysis Report
(USAR) for the emergency diesel generators (EDGs). Specifically, the
proposed change would describe a departure from Safety Guide 9,
``Selection of Diesel Generator Set Capacity for Standby Power
Supplies,'' for the frequency and voltage transient during the EDG
automatic loading sequence.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No. The proposed amendment alters the design
requirements for the Emergency Diesel Generators (EDGs).
Specifically, the proposed amendment affects the requirements for
EDG voltage and frequency response following a loss of offsite
power. The EDGs function to mitigate the consequences of accidents
when offsite power is not available. The EDGs are not an initiator
of any analyzed accident.
The effect of this change on the capability of the EDGs, the
onsite electric power system, and essentially powered equipment to
perform their required safety functions has been evaluated, and the
proposed change does not significantly impact the capability of
these systems to perform their required accident mitigation
functions. No previous analyzed accident scenario is affected by the
proposed change.
The proposed change does not affect the initiation of any
analyzed accident. The accident mitigation functions for affected
equipment are maintained. 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 amendment affects the USAR
requirements for EDG voltage and frequency response following a loss
of offsite power. The effect of this change on the capability of the
EDGs, the onsite electric power system, and essentially powered
equipment to perform their required safety functions has been
evaluated, and the proposed change does not significantly impact the
capability of these systems to perform their required safety
functions. The assumptions of the current accident analyses are
maintained and no new or different accident initiators are created.
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 amendment affects the USAR
requirements for EDG voltage and frequency response following a loss
of offsite power. The effect of this change on the capability of the
EDGs, the onsite electric power system, and essentially powered
equipment to perform their required safety functions has been
evaluated, and it is concluded the proposed change does not impact
the capability of these systems to perform their required safety
functions. However, since the proposed change does make changes to
the controlling values for EDG voltage and frequency transient
response that are less restrictive than those presently described in
the USAR, this is considered a reduction in a margin of safety.
The magnitude of voltage and frequency drops which would result
in failure of the EDGs, the onsite power system, or essentially
powered equipment have not been determined due to the limitations of
the transient assessment model and the nonlinear phenomena
associated with that postulated failure. However, based on (1) a
computer model and testing of the diesel engine, engine speed
control governor and actuator, the synchronous generator and
excitation system that demonstrate the EDGs are capable of starting,
accelerating, and carrying the required loads, (2) a comprehensive
evaluation of the impact of the transient voltage and frequency
response on plant equipment and safety functions, (3) the momentary
duration of the voltage and frequency dips, and (4) based on
engineering judgement, the proposed change is not considered to have
a significant effect on the margin of safety. 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: April 23, 2004.
Description of amendment request: The proposed amendments would
revise several Technical Specification (TS) Allowed Outage Times for TS
3.3.3, Accident Monitoring, to be consistent with the Completion Times
in the related Specification in NUREG-1431, Revision 2, ``Standard
Technical Specifications Westinghouse Plants (the Improved Standard
Technical Specifications, or ISTS).''
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. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes revise the Actions and allowed outage times
of the accident monitoring instrumentation. The accident monitoring
instrumentation is not an initiator of any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased by these proposed changes.
The Technical Specifications continue to require the accident
monitoring instrumentation to be operable. Therefore, the accident
monitoring instrumentation will continue to provide sufficient
information on selected plant parameters to monitor and assess these
variables following an accident. The consequences of an accident
during the extended allowed outage time are the same as the
consequences during the current allowed outage time. As a result,
the consequences of any accident previously evaluated are not
significantly increased by these proposed changes. Therefore, the
proposed amendments do not involve a significant increase in the
probability or consequences of any accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
[[Page 29768]]
kind of accident from any previously evaluated.
The proposed changes do not alter the design, physical
configuration, or mode of operation of the plant. The accident
monitoring instrumentation is not an initiator of any accident
previously evaluated. No changes are being made to the plant that
would introduce any new accident causal mechanisms. The proposed
changes do not affect any other plant equipment. Therefore,
operation of the facility in accordance with the proposed amendments
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes do not change the operation, function, or
modes of the plant or equipment operation. The proposed changes do
not change the level of assurance that the accident monitoring
instrumentation will be available to perform its function. The
proposed changes provide a more appropriate time to restore the
inoperable channel(s) to operable status, and only apply when one or
more channels of a required instrument are inoperable. The
additional time to restore an inoperable channel to operable status
is appropriate based on the low probability of an event requiring an
accident monitoring instrument during the interval, providing a
reasonable time for repair, and other means which may be available
to obtain the required information. Therefore, operation of the
facility in accordance with the proposed amendments would 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: William F. Burton, Acting.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: March 15, 2004.
Description of amendment request: Maine Yankee Atomic Power Company
(Maine Yankee) is requesting that the U.S. Nuclear Regulatory
Commission (NRC) release the remaining land under License No. DPR-36,
with the exception of land where the Independent Spent Fuel Storage
Installation is located. Maine Yankee submitted detailed information on
dismantlement activities and final status survey results for the Spray
Building and Spray Pipe with the amendment request, and proposes to
submit dismantlement and survey information for the remaining land area
in four additional submittals. Maine Yankee is seeking review and
approval of the amendment; however, Maine Yankee is requesting that the
NRC condition the effective date of the license amendment to correspond
with the NRC's approval of the final information submittal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The requested license amendment involves release of land
presently considered part of the Maine Yankee plant site under
license DPR-36. The release of this land will occur after all
demolition activities are completed and final status surveys have
been performed to document the final radiological conditions of the
land. When the release occurs, the only remaining radiological
hazard at the site will be contained in the Independent Spent Fuel
Storage Installation (ISFSI). Therefore, the focus of the analysis
is on the potential impact on the probability and consequences of
accidents associated with the ISFSI.
The accident conditions evaluated for the spent fuel storage
casks include the following: accident pressurization, mis-loading of
fuel canisters, drop of the vertical concrete casks, explosion,
fires, maximum anticipated heat load, earthquakes, floods,
lightening strikes, tornado and tornado driven missiles, tip over of
vertical concrete cask, and full blockage of vertical concrete cask
air inlets and outlets. The release of the non-ISFSI land from the
license will not affect the probability of any of these accidents.
Maine Yankee will retain sufficient control over activities
performed on the Owner Controlled Area through rights granted in the
legal land conveyance documents to ensure that there is no impact on
consequences from postulated accidents. Therefore, the proposed
release of the land will not affect the consequences of any of these
postulated accidents.
The proposed action, therefore, does not increase either the
probability or the consequences of any accidents that have been
considered.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The requested amendment involves release of land presently
considered part of the Maine Yankee plant site under license DPR-36.
When the amendment becomes effective, demolition activities will be
complete and all systems, structures and components will have been
removed from the land. The requested release of the land does not
create the possibility of a new or different kind of accident that
could affect the ISFSI that has not been considered in the design,
installation or operation of the ISFSI. As noted above, Maine Yankee
will retain control over activities performed in the Owner
Controlled Area for the ISFSI to assure that no new hazards are
introduced that could create the potential for 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.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The margin of safety defined in the statements of consideration
for the final rule on the Radiological Criteria for License
Termination is described as the margin between the 100 mrem/yr
public dose limit established in 10 CFR 20.1301 for licensed
operation and the 25 mrem/yr dose limit to the average member of the
critical group at a site considered acceptable for unrestricted use.
This margin of safety accounts for the potential effect of multiple
sources of radiation exposure to the critical group. Additionally,
the State of Maine, through legislation, has imposed a 10 mrem/yr
all pathways dose limit, with no more than 4 mrem/yr attributable to
drinking water sources.
The License Termination Plan (LTP) prepared by Maine Yankee
establishes conservative criteria for residual radiation levels
following completion of demolition activities at the site. The LTP
demonstrates that when these conservative criteria are met, the dose
to the average member of the critical group will be below the
regulatory criteria established by the State of Maine, and,
therefore, well below the dose limits established by the NRC. The
proposed release of the site lands, once the criteria established in
the LTP have been met will, therefore, not result in any reduction
in the margin of safety.
Conclusion
Based on the above, Maine Yankee concludes that the proposed
amendment presents no significant hazards consideration under the
standards set forth in 10 CFR 50.92(c), and, accordingly, a finding
of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendment involves no significant hazards consideration.
Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power
Company, 321 Old Ferry Road, Wiscasset, Maine 04578
NRC Section Chief: Claudia M. Craig.
[[Page 29769]]
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: April 19, 2004.
Description of amendment request: The licensee proposed to revise
the Technical Specifications (TSs) to establish an operating cycle (24-
month) calibration surveillance frequency for the Intermediate Range
Monitor (IRM) instrumentation, which would replace the current ``prior
to startup and normal shutdown'' Surveillance Requirement (SR). The
proposed changes also included associated conforming changes. In
addition, the licensee proposed to relocate the Limiting Conditions for
Operation (LCOs) and SRs for selected control rod withdrawal block
instrumentation to the Updated Final Safety Analysis Report (UFSAR), a
licensee-controlled document.
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 are limited to: (1) establishing a 24-month
calibration frequency for the IRM instrumentation in lieu of the
current ``prior to startup and normal shutdown'' requirement and
incorporating the associated conforming changes, and (2) the
relocation of certain instrumentation requirements from the TSs that
do not satisfy the screening criteria for retention in the TSs. The
proposed changes do not introduce any new modes of plant operation,
make any physical changes to the plant, or alter any operational
setpoints in a manner which could degrade the performance of, or
increase the challenges to, any safety system assumed to function in
the accident analysis. In addition, evaluations of the proposed
changes pursuant to NRC and industry guidance demonstrate that the
availability and reliability of equipment and systems required to
prevent or mitigate the radiological consequences of an accident are
not significantly affected. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes establish a 24-month IRM calibration
frequency in lieu of the current ``prior to startup and normal
shutdown'' requirement and relocate certain instrumentation
requirements to the UFSAR. As such, the proposed changes do not
eliminate any requirements or impose any new requirements, and
adequate controls of existing requirements are maintained.
Furthermore, since the proposed changes do not make any physical
changes to the plant, no new accident initiators or failure
mechanisms are introduced, and the accident assumptions and initial
conditions will remain unchanged. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any accident [previously] evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes establish a 24-month IRM calibration
frequency in lieu of the current ``prior to startup and normal
shutdown'' requirement and relocate certain instrumentation
requirements to the UFSAR. Although the proposed changes result in
changes to surveillance intervals, the impact, if any, on system
availability is small based on (1) other more frequent testing that
is performed, (2) the existence of redundant equipment, and (3)
overall system reliability. Consistent with the findings of previous
industry evaluations, the NMP1 [Nine Mile Point Nuclear Station,
Unit No. 1] plant-specific analyses have shown no evidence of time-
dependent failures that would impact the availability of the
affected systems. Furthermore, plant-specific evaluations and the
adoption of the calculated IRM setpoint Allowable Values ensure that
the setpoint margins are maintained for a 24-month (30-month
maximum) calibration frequency. The proposed relocated requirements
are consistent with the Improved Standard TSs (NUREG-1433 and NUREG-
1434) and 10 CFR 50.36, and will be maintained in accordance with 10
CFR 50.59. Accordingly, the proposed changes will have no
significant impact on the condition or performance of structures,
systems, and components relied upon for accident mitigation.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Richard J. Laufer.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: January 20, 2004.
Description of amendment request: This License Amendment Request
(LAR) proposes selective scope application of the alternate source term
(AST) for the fuel handling accident (FHA) in accordance with the
provisions of 10 CFR 50.67. Nuclear Management Company requests the
Nuclear Regulatory Commission (NRC) review and approval of the AST FHA
methodology for application to the Prairie Island Nuclear Generating
Plant. This LAR also proposes revisions to Technical Specifications
(TS) associated with ensuring that safety analyses assumptions are met
for a postulated FHA in containment. Based on the AST FHA analyses,
this LAR proposes to modify TS 3.9.4, ``Containment Penetrations,'' to
apply during the handling of recently irradiated fuel and require all
containment penetrations to be closed during handling of recently
irradiated fuel; and also proposes to remove the requirements of TS
3.3.5, ``Containment Ventilation Isolation Instrumentation'' relating
to movement of irradiated fuel assemblies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification changes require containment
integrity during movement of recently irradiated fuel. With this
change, the Technical Specifications selectively implement 10 CFR
50.67 alternative source term methodologies for a fuel handling
accident and implement portions of the approved industry improved
Standard Technical Specification traveler, TSTF-51, ``Revise
containment requirements during handling irradiated fuel and core
alterations' as it applies to TS 3.9.4, ``Containment
Penetrations.'' This change also removes requirements for
containment ventilation isolation instrumentation during handling
irradiated fuel from TS 3.3.5, ``Containment Ventilation Isolation
Instrumentation'' since the containment purge and inservice purge
system penetrations which are isolated by this instrumentation will
be required to be isolated during movement of recently irradiated
fuel. With the proposed 10 CFR 50.67 alternative source term
methodologies, these filtration systems are not assumed to function
during a fuel handling accident involving fuel which is not recently
irradiated.
This amendment does not alter the methodology or equipment used
directly in fuel handling operations. None of the containment
integrity features including the containment equipment hatch,
personnel air locks or any other containment penetration
[[Page 29770]]
are used to handle fuel. Therefore, containment integrity and
ventilation systems, and spent fuel pool ventilation systems are not
accident initiators and therefore these changes do not increase the
probability of a previously evaluated accident.
The total effective dose equivalent (TEDE) doses from the
analysis supporting this amendment request have been compared to
equivalent total effective dose equivalent (TEDE) doses estimated
with the guidelines of Regulatory Guide 1.183 Footnote 7. The new
values are shown to be comparable to the results of the previous
analysis.
A fuel handling accident analysis utilizing alternative source
term methodologies allowed by 10 CFR 50.67 demonstrated that the
dose consequences of a postulated fuel handling accident remain
within the limits of 10 CFR 50.67 without taking credit for
containment closure or ventilation systems assuming the fuel has not
recently been in a critical reactor. The alternative source term
fuel handling accident analysis also demonstrated that the more
restrictive dose guidelines of Regulatory Guide 1.183 are also met
without taking credit for these mitigation features. Since the
alternative source term fuel handling accident analysis results are
within the regulatory limits and regulatory guidelines without
taking credit for these mitigation features, revising this Technical
Specification for containment closure does not involve a significant
increase in the consequences of a previously evaluated accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed Technical Specification changes require containment
integrity during movement of recently irradiated fuel. With this
change, the Technical Specifications selectively implement 10 CFR
50.67 alternative source term methodologies for a fuel handling
accident and implement portions of the approved industry improved
Standard Technical Specification traveler, TSTF-51, ``Revise
containment requirements during handling irradiated fuel and core
alterations'' as it applies to TS 3.9.4, ``Containment
Penetrations.'' This change also removes requirements for
containment ventilation isolation instrumentation during handling
irradiated fuel from TS 3.3.5, ``Containment Ventilation Isolation
Instrumentation'' since the containment purge and inservice purge
system penetrations which are isolated by this instrumentation will
be required to be isolated during movement of recently irradiated
fuel. With the proposed 10 CFR 50.67 alternative source term
methodologies, these filtration systems are not assumed to function
during a fuel handling accident involving fuel which is not recently
irradiated.
The proposed Technical Specification changes do not involve
plant design, hardware, system operation, or procedures involved
with actual handling of irradiated fuel. The proposed changes
include application of new methodology for fuel handling accident
analysis and revises requirements for equipment operability during
movement of irradiated fuel assemblies. These changes do not create
the possibility for a new or different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed Technical Specification changes require containment
integrity during movement of recently irradiated fuel. With this
change, the Technical Specifications selectively implement 10 CFR
50.67 alternative source term methodologies for a fuel handling
accident and implement portions of the approved industry improved
Standard Technical Specification traveler, TSTF-51, ``Revise
containment requirements during handling irradiated fuel and core
alterations' as it applies to TS 3.9.4, ``Containment
Penetrations.'' This change also removes requirements for
containment ventilation isolation instrumentation during handling
irradiated fuel from TS 3.3.5, ``Containment Ventilation Isolation
Instrumentation'' since the containment purge and inservice purge
system penetrations which are isolated by this instrumentation will
be required to be isolated during movement of recently irradiated
fuel. With the proposed 10 CFR 50.67 alternative source term
methodologies, these filtration systems are not assumed to function
during a fuel handling accident involving fuel which is not recently
irradiated.
The assumptions and input used in the fuel handling accident
analysis are conservative. The design basis fuel handling accident
has been defined to identify conservative conditions. The source
term and radioactivity releases have been calculated pursuant to
Regulatory Guide 1.183, Appendix B and with conservative assumptions
concerning prior reactor operations. The control room atmospheric
dispersion factor has been calculated with conservative assumptions
associated with the release. These conservative assumptions and
input ensure that the radiation doses cited in this license
amendment request are the upper bounds to radiological consequences
of a fuel handling accident in containment or the spent fuel pool.
The analysis shows that there is a significant margin between the
offsite radiation doses calculated for the postulated fuel handling
accident using the alternate source term and the dose limits of 10
CFR 50.67 and acceptance criteria of Regulatory Guide 1.183. The
proposed changes will not degrade the plant protective boundaries,
will not cause a release of fission products to the public, and will
not degrade the performance of any structures, systems, and
components important to safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, 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 Systems (ADAMS) Public
Electronic
[[Page 29771]]
Reading Room on the Internet at the NRC Web site, http://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
NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-
415-4737 or by e-mail to [email protected].
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 application for amendments: April 17, 2003, as supplemented
July 29, 2003.
Brief description of amendments: These amendments revise the
Required Actions requiring suspension of operations involving positive
reactivity additions and various notes that preclude reduction of boron
concentration.
Date of issuance: May 6, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 266 and 243.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28841).
The July 29, 2003, letter clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register on May 27, 2003 (68
FR 28841).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated May 6, 2004.
No significant hazards consideration comments received: No.
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 application for amendments: November 5, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications to adopt the provisions of Industry/Technical
Specification Task Force change TSTF-359, ``Increase Flexibility in
Mode Restraints.''
Date of issuance: April 29, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 213, 207.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7520)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 29, 2004.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: November 5, 2003.
Brief description of amendments: The amendments revised the
Technical Specifications to adopt the provisions of Industry/Technical
Specification Task Force change TSTF-359, ``Increase Flexibility in
Mode Restraints.''
Date of issuance: April 29, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 221, 203.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7520)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 29, 2004.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: February 9, 2004, as
supplemented by letter dated March 2, 2004.
Brief description of amendment: The amendment removed the
pressurizer heatup and cooldown limits, and the associated action and
surveillance requirements, from the Technical Specifications and placed
them in the Technical Requirements Manual.
Date of issuance: May 4, 2004.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 253.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9860).
The March 2, 2004, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 4, 2004.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Units Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: July 18, 2002, as supplemented
November 14, 2002, and December 11, 2003.
Brief description of amendments: The amendments relocate Technical
Specification (TS) 3/4 9.7 regarding the Spent Fuel Storage Pool
Building cranes and TS 3/4 9.13 (Unit 1) and TS 3/4 9.12 (Unit 2)
regarding spent fuel cask cranes to the respective units' Updated Final
Safety Analysis Report.
Date of Issuance: April 28, 2004
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: 190 and 134
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (67 FR
50954). The November 14, 2002, and December 11, 2003, supplements did
not affect the original proposed no significant hazards determination,
or expand the scope of the request as noticed in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 28, 2004.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,
Unit No. 1, St. Lucie County, Florida
Date of application for amendment: May 22, 2002, as supplemented by
letters dated December 5, 2002, and February 11, 2004.
Brief description of amendment: The amendment revised Technical
Specification 6.9.1.11.b to add two NRC-approved topical reports to the
Core Operating Limits Report methodology list, and delete superseded
reports. Also, the method of listing topical reports was revised to be
consistent with Technical Specifications Task Force 363, which has been
approved by the NRC.
Date of Issuance: May 6, 2004.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 191.
Facility Operating License No. DPR-67: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 25, 2002 (67 FR
42827).
[[Page 29772]]
The supplemental letters provided clarifying information that was
within the scope of the initial notice and did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 6, 2004.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: August 22, 2003, as supplemented
by letters dated January 12 and March 11, 2004.
Brief description of amendment: The amendment revised Section
3.7.1, ``Service Water (SW) System and Ultimate Heat Sink (UHS),'' by
adding a new Condition G to allow continued operation with short-term
elevated UHS temperatures.
Date of issuance: May 7, 2004.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 113.
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 30, 2003 (68
FR 56344).
The January 12 and March 11, 2004, letters provided clarifying
information within the scope of the original application, and did not
change the staff's initial proposed no significant hazards
consideration determination. The staff's related evaluation of the
amendment is contained in a Safety Evaluation dated May 7, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: January 30, 2004.
Brief description of amendment: The amendment relocates the
requirements for hydrogen monitors from the Technical Specifications to
the Technical Requirements Manual.
Date of issuance: May 13, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 174.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9862).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 13, 2004.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: July 25, 2003, as supplemented on
December 5, 2003
Brief description of amendment: The amendment modifies Technical
Specification (TS) 2.1.4, ``Reactor Coolant System (RCS) Leakage
Limits,'' by (1) adding a requirement for no RCS pressure boundary
leakage, (2) combining the existing RCS leakage limits into a format
similar to the Improved Standard TS (ISTS), and (3) replacing the
existing basis associated with this TS with a basis similar in format
and content to the ISTS.
Date of issuance: May 7, 2004.
Effective date: As of the date of issuance, to be implemented
within 90 days from issuance.
Amendment No.: 226.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: August 19, 2003 (68 FR
49818).
The December 5, 2003, supplemental letter provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 7, 2004.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: December 30, 2003, and its
supplement dated March 11, 2004.
Brief description of amendments: The amendments eliminate the
requirements in the technical specifications associated with hydrogen
recombiners and hydrogen monitors.
Date of issuance: May 4, 2004.
Effective date: May 4, 2004, and shall be implemented within 60
days from the date of issuance.
Amendment Nos.: Unit 1--168; Unit 2--169.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9864).
The March 11, 2004, supplemental letter provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 4, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 14th May 2004.
For the Nuclear Regulatory Commission.
Eric J. Leeds,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 04-11507 Filed 5-24-04; 8:45 am]
BILLING CODE 7590-01-P