[Federal Register Volume 69, Number 148 (Tuesday, August 3, 2004)]
[Notices]
[Pages 46582-46596]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-17346]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from July 9, 2004 through July 22, 2004. The last
biweekly notice was published on July 20, 2004 (69 FR 43457).
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
01F21, 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
[[Page 46583]]
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 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].
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: May 21, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specifications 5.6.6, ``Reactor Coolant System (RCS)
Pressure and Temperature Limits Report (PTLR),'' by adding a reference
to the use of previous Nuclear Regulatory Commission approved Code
Cases N-640 and N-588 as acceptable methods for determining reactor
pressure vessel (RPV) pressure temperature (P-T) 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. Does the proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The use of Code Cases N-588 and N-640 has been approved for
Braidwood and Byron Stations. The use of P-T limits based on these
Code Cases will continue to ensure that
[[Page 46584]]
the RPV integrity is maintained under all conditions.
Thus there is no increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed TS 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 the use or installation of
new equipment. No equipment will be operated in a new or different
manner. No new or different system interactions are created and no
new processes are introduced. The proposed change will not introduce
any new failure mechanisms, malfunctions, or accident initiators not
already considered in the design and licensing bases.
Based on this evaluation, 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 TS change does not involve a significant
reduction in a margin of safety?
The P-T limits provide assurance that RPV integrity is
maintained. The use of Code Cases N-588 and N-640 has been
previously approved by the NRC for Braidwood and Byron Stations and
will continue to ensure that RPV integrity is maintained.
Thus, there is no 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Edward J. Cullen, Deputy General
Counsel, Exelon BSC-Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County,
Pennsylvania
Date of amendment request: June 28, 2004.
Description of amendment request: The proposed amendment would
revise the BVPS-1 Technical Specification (TS) 4.4.5.4.a.8 to modify
the definition of steam generator (SG) tube inspection to exclude the
portion of the tube within the tube sheet below the W* distance. The W*
distance is defined as the distance from the top of the tube sheet to
the bottom of the W* length (7.0 in. on the hot leg side) including the
distance from the top of the tube sheet to the bottom of the WEXTEX
(Westinghouse explosive tube expansion) Transition (approximately 0.25
in.) plus uncertainties (0.12 in.). The proposed amendment would also
revise the SG tube repair criteria of TS 4.4.5.4.a.6 to indicate that
service-induced degradation within the W* distance or less than 8.0 in.
below the top of the tube sheet shall be repaired upon detection. The
proposed amendment would also add TS 4.4.5.2.e to require a 100%
rotating pancake coil probe inspection of the hot leg tube sheet W*
distance, add new W* terminology definitions in TS 4.4.5.4.a.11, and
add a new reporting criteria for W* inspection information to TS
4.4.5.5.d.1 and TS 4.4.5.5.e. This proposed amendment would be
effective for only one operating cycle, as the licensee plans to
replace SGs during the 2006 refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change modifies the [BVPS-1] TSs to incorporate
steam generator (SG) tube inspection scope based on WCAP-14797,
Revision 2 [``Generic W* Tube Plugging Criteria for 51 Series Steam
Generator Tubesheet Region WEXTEX Expansions,'' dated March 2003
(proprietary)]. Of the various accidents evaluated in the [BVPS-1]
Updated Final Safety Analysis Report (UFSAR), the proposed changes
only affect the steam generator tube rupture (SGTR) event evaluation
and the postulated steam line break (SLB) accident evaluation. Loss-
of-coolant accident (LOCA) conditions cause a compressive axial load
to act on the tube. Therefore, since the LOCA tends to force the
tube into the tubesheet rather than pull it out, it is not a factor
in this amendment request. Another faulted load consideration is a
safe shutdown earthquake (SSE); however, the seismic analysis of
Series 51 steam generators has shown that axial loading of the tubes
is negligible during an SSE.
For the SGTR event, the required structural margins of the steam
generator tubes will be maintained by the presence of the tubesheet.
Tube rupture is precluded for cracks in the Westinghouse explosive
tube expansion (WEXTEX) region due to the constraint provided by the
tubesheet. Therefore, Regulatory Guide (RG) 1.121, ``Bases for
Plugging Degraded PWR [pressurized-water reactor] Steam Generator
Tubes,'' margins against burst are maintained for both normal and
postulated accident conditions.
The W* length supplies the necessary resistive force to preclude
pullout loads under both normal operating and accident conditions.
The contact pressure results from the WEXTEX expansion process,
thermal expansion mismatch between the tube and tubesheet and from
the differential pressure between the primary and secondary side.
The proposed changes do not affect the other systems, structures,
components or operational features. Therefore, the proposed change
results in no significant increase in the probability of the
occurrence of an SGTR or SLB accident.
The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary
leakage flow through a postulated broken tube is not affected by the
proposed change since the tubesheet enhances the tube integrity in
the region of the WEXTEX expansion by precluding tube deformation
beyond its initial expanded outside diameter. The resistance to both
tube rupture and collapse is strengthened by the tubesheet in that
region. At normal operating pressures, leakage from primary water
stress corrosion cracking (PWSCC) below the W* length is limited by
both the tube-to-tubesheet crevice and the limited crack opening
permitted by the tubesheet constraint. Consequently, negligible
normal operating leakage is expected from cracks within the
tubesheet region.
SLB leakage is limited by leakage flow restrictions resulting
from the crack and tube-to-tubesheet contact pressures that provide
a restricted leakage path above the indications and also limit the
degree of crack face opening compared to free span indications. The
total leakage, that is, the combined leakage for all such tubes
meet[s] the industry performance criterion, plus the combined
leakage developed by any other alternate repair criteria, will be
maintained below the maximum allowable SLB leak rate limit, such
that off-site doses are maintained less than 10 CFR 100 guideline
values and the limits evaluated in the [BVPS-1] UFSAR.
Therefore, based on the above evaluation, the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed changes do not introduce any changes or
mechanisms that create the possibility of a new or different kind of
accident. Tube bundle integrity is expected to be maintained for all
plant conditions upon implementation of the W* methodology.
The proposed changes do not introduce any new equipment or any
change to existing equipment. No new effects on existing equipment
are created nor are any new malfunctions introduced.
Therefore, 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?
No. The proposed changes maintain the required structural
margins of the steam generator tubes for both normal and accident
conditions. NRC [Nuclear Regulatory Commission] Regulatory Guide
(RG) 1.121 is used as the basis in the development of the W*
methodology for determining that steam generator tube integrity
considerations are
[[Page 46585]]
maintained within acceptable limits. RG 1.121 describes a method
acceptable to the NRC staff for meeting General Design Criteria 14,
15, 31, and 32 by reducing the probability and consequences of an
SGTR. RG 1.121 concludes that by determining the limiting safe
conditions of tube wall degradation beyond which tubes with
unacceptable cracking, as established by inservice inspection,
should be removed from service or repaired, the probability and
consequences of a[n] SGTR are reduced. This RG uses safety factors
on loads for tube burst that are consistent with the requirements of
Section III of the American Society for Mechanical Engineers (ASME)
[Boiler and Pressure Vessel] Code.
For primarily axially oriented cracking located within the
tubesheet, tube burst is precluded due to the presence of the
tubesheet. WCAP-14797, Revision 2, defines a length, W*, of
degradation free expanded tubing that provides the necessary
resistance to tube pullout due to the pressure induced forces (with
applicable safety factors applied). Application of the W* criteria
will preclude unacceptable primary-to-secondary leakage during all
plant conditions. The methodology for determining leakage provides
for large margins between calculated and actual leakage values in
the W* criteria.
Plugging of steam generator tubes reduces the reactor coolant
flow margin for core cooling. Implementation of W* methodology at
[BVPS-1] will result in maintaining the margin of flow that may have
otherwise been reduced by tube plugging.
Based on the above, it is concluded that the proposed changes do
not result in a significant reduction [in a margin of safety] as
defined in the [UFSAR] or [B]ases of the plant [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: Mary O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: July 8, 2004.
Description of amendment request: The amendment request proposes to
delete one-time use footnotes that have expired or have already been
used from the Crystal River Unit 3 (CR-3) Improved Technical
Specifications (ITS). Specifically, obsolete notes will be removed from
ITS 3.8.1, ``AC Sources--Operating (Emergency Diesel Generator),'' ITS
3.7.9, ``Nuclear Services Seawater System,'' and ITS 3.7.18, ``Control
Complex Cooling System.'' This change is administrative in nature and
does not alter any operating license requirements.
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 and states that the amendment
request:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Each footnote was added to ITS through the license amendment
process. The activities supported by the footnotes were performed
and, therefore, the footnotes have no further utility. Deleting the
footnotes is administrative in nature and does not affect plant
conditions that could impact accident probability or consequences.
Therefore, granting this LAR [license amendment request] does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does not create the possibility of a new or different type of
accident from any accident previously evaluated.
The proposed license amendment deletes footnotes that were used
on a one-time basis for several specifications. The proposed LAR
will not result in changes to the design, physical configuration of
the plant or the assumptions made in the safety analysis. Therefore,
the proposed change will not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does not involve a significant reduction in the margin of
safety.
The deletion of the footnotes from the ITS does not affect
properties of plant components or their operation. Therefore,
granting this LAR 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602
NRC Acting Section Chief: Michael L. Marshall, Jr.
Indiana Michigan Power Company, Docket Nos. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of amendment request: June 25, 2004.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) and the bases to reduce the
temperature at which shutdown and control rod drop tests are performed
from greater than or equal to 541 degrees Fahrenheit to greater than or
equal to 500 degrees Fahrenheit. Additionally, the proposed amendment
would make format changes to improve the TS appearance.
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 probability of occurrence of an accident previously evaluated
is not altered by the proposed amendment. The proposed change does not
impact the integrity of the reactor coolant system pressure boundary
and, therefore, does not increase the potential for the occurrence of a
loss-of-coolant accident. The change does not make any physical changes
to the facility design, material or construction standards, and the
proposed change is not an initiator or contributor to any currently
evaluated accident. The format changes are intended to improve
appearance, and do not alter any requirements. Thus, neither the
probability nor the consequences of a previously analyzed accident are
significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The rod drop test is routinely performed during each refueling
outage. Decreasing the test temperature will not create the possibility
of a new or different accident. The proposed test conditions remain
bounded by the analysis of record since the rod drop time assumed in
the accident analysis will not be changed. The format changes are
intended to improve appearance, and do not alter any requirements.
Since no new failure modes are associated with the proposed changes,
the proposed amendment 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 Technical Specification change does not involve a significant
reduction in margin because the acceptance
[[Page 46586]]
criterion for the rod drop time will not change. The proposed change
will reduce the minimum rod drop test temperature from greater than or
equal to 541 degrees Fahrenheit to greater than or equal to 500 degrees
Fahrenheit. This will slightly increase the measured test rod drop
time. The measured test rod drop time, however, will be within the
current Technical Specification limit of 2.4 seconds. The format
changes are intended to improve appearance, and do not alter any
requirements. Therefore, the margin of safety is not impacted by the
proposed amendment.
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: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 15, 2004.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) Section 3.8.1, AC Sources--
Operating, Condition B, to extend the allowed outage time for one
Diesel Generator (DG) inoperable from 7 days to 14 days and TS Section
3.8.3, Diesel Fuel Oil, Lube Oil, and Starting Air, Limiting Condition
for Operation, to allow the use of temporary fuel oil storage tanks to
supply the required fuel oil storage inventory.
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 Standby AC Power System (Diesel Generators) provides onsite
electrical power to vital systems should offsite electrical power be
interrupted. It is not an initiator to any accident previously
evaluated. Therefore, the extended period of operation with one diesel
generator inoperable and the seven day required fuel oil supply being
provided in part by temporary storage tanks will not increase the
probability of an accident previously evaluated.
The Standby AC Power System acts to mitigate the consequences of
design basis accidents that assume a loss of offsite power. For that
purpose, redundant diesel generators are provided to protect against a
single failure. During the Technical Specification seven day allowed
outage time, an operating unit is allowed by the Technical
Specifications to remove one diesel generator from service, thereby
losing this single failure protection. During the requested fourteen
day allowed outage time for fuel oil storage tank cleaning and coating
maintenance activities, the inoperable diesel generator will be
maintained available to start and load, with a minimum of five (5)
hours of fuel available in the day tank. Manual actions contained in
approved procedures to provide fuel from temporary storage tanks to
either the operable diesel generator or the inoperable but available
diesel generator will be implemented. A risk evaluation determined that
the probability of failure to implement the contingency actions is
sufficiently low that it does not adversely impact the availability of
the Standby AC Power System.
The vulnerability to external events, seismic, high winds and fire,
was also evaluated and judged to be not significant due to the low
probability of these events during the period of time this proposed
amendment will be in effect, and the defense in depth strategies being
put in place during the tank maintenance activities.
In the event that fuel stored in the temporary tanks is not
available to support full load operation of the diesel generator beyond
four (4) days, replenishment of fuel oil from offsite can be
accomplished in approximately 24 hours through the use of existing
purchase orders for fuel oil and diesel fuel analysis. Therefore,
during the period of the extended allowed outage time and the use of
temporary fuel oil storage tanks, there is no significant increase in
the 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.
Operation with one diesel generator inoperable but available for an
extended period or with part of the required diesel fuel stored in
temporary tanks does not involve any new mode of plant operation or
different function for plant equipment. Operation in this configuration
does introduce proceduralized manual actions to supply fuel to either
diesel generator from the permanent storage tank or the temporary tank.
These actions can be accomplished within the five hours of full load
diesel operation from fuel stored in the day tank. A risk evaluation
determined that the probability of failure to implement the contingency
actions is sufficiently low that it does not adversely impact the
availability of the Standby AC Power System. There are no new accident
precursors generated due to this temporary extension of allowed outage
time or the use of a temporary fuel oil storage system.
3. Do the proposed changes involve a significant reduction in the
margin of safety?
Response: No.
A single failure of the operable fuel oil transfer pump could
prevent DG operation beyond five hours. Proceduralized manual actions
to supply fuel to either diesel generator from the permanent storage
tank or the temporary tank will be implemented to mitigate this single
failure vulnerability. These actions can be accomplished within the
five (5) hours of full load diesel operation from fuel stored in the
day tank. A risk evaluation determined that the probability of failure
to implement the contingency actions is sufficiently low that it does
not adversely impact the availability of the Standby AC Power System.
Therefore, during the extended allowed outage time and the use of a
temporary fuel oil storage system, the Standby AC Power System
maintains the ability to provide a source of on-site AC power adequate
for maintaining the safe shutdown of the reactor following abnormal
operational transients and postulated accidents.
IEEE [Institute of Electrical and Electronics Engineers] Design
Standard 308-1970, ``IEEE Criteria for Class 1E Electric Systems for
Nuclear Power Generating Station,'' Section 5.2.4, ``Standby Power
Supply,'' Paragraph 6), ``Energy Storage,'' contains the requirement
for stored energy capacity to be the longer of (a) seven days or (b)
time required to replenish the energy from sources away from the
generating unit's site following the limiting design basis event.
Cooper Nuclear Station's Updated Safety Analysis Report documents that
the Standby AC Power System conforms to the applicable sections of IEEE
308-1970.
The Diesel Generator Diesel Oil Storage and Transfer System will be
configured to ensure a minimum fuel oil inventory to support greater
than four (4) days of full load diesel generator operation is
maintained in the operable permanent storage tank. Existing cross-
[[Page 46587]]
tie capabilities in the fuel storage and transfer system piping, in
conjunction with proceduralized manual actions, ensure the four day
fuel supply is available to either diesel generator. The remaining
three (3) day fuel supply will be stored in temporary non-Class I tanks
and would potentially be vulnerable to external events. The
vulnerability to external events, seismic, high winds and fire, was
evaluated and judged to be not significant due to the low probability
of these events during the period of time this proposed amendment will
be in effect, and the defense in depth strategies being put in place
during the tank maintenance activities.
In the event that fuel stored in the temporary tanks is not
available to support full load operation of the diesel generator beyond
four (4) days, replenishment of fuel oil from offsite can be
accomplished in approximately 24 hours through the use of existing
purchase orders for fuel oil and diesel fuel analysis.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: June 8, 2004.
Description of amendment request: The Humboldt Bay Power Plant,
Unit 3, is a decommissioning nuclear power plant that was permanently
shutdown in July 1976. The plant is currently in a safe storage
(SAFSTOR) condition to ensure that necessary plant systems will be
operated and maintained as needed to preserve safe conditions within
the facility to prevent deterioration until active decommissioning can
commence. All spent fuel is stored in the spent fuel pool. Pacific Gas
and Electric Company (PG&E) has proposed a license amendment to clarify
the technical specifications applicability to current plant conditions
and practices. Specifically, the requested changes clarify that:
(1) Fuel fragments within the spent fuel pool totaling less than
one fuel assembly and damaged fuel assembly UD-6N do not have to be
stored in containers made of neutron absorbing material. Furthermore,
that one additional assembly can be removed from a neutron absorbing
container to perform fuel handling activities.
(2) The control station for Humboldt Bay Units 1 and 2 is
considered to be anywhere on the +27 foot operating deck.
(3) References to certain technical specification section
designators that contain typographical errors have been corrected.
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?
No. The proposed changes provide either clarification to reflect
plant conditions or correct typographical errors. Existing accident
analysis assumptions bound the proposed addition of not storing fuel
fragments, which may be considered as less than or equal to a fuel
assembly, in a container made with neutron absorbing material. The
proposed changes involve no changes to plant systems or accident
analysis, and as such, do not affect initiators of analyzed events
or assumed mitigation of accidents. Therefore, the proposed changes
do not increase the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different type of accident from any accident previously
evaluated?
No. The proposed changes provide either clarification to reflect
plant conditions or correct typographical errors. Existing accident
analysis assumptions bound the proposed addition of not storing fuel
fragments, which may be considered as less than or equal to a fuel
assembly, in a container made with neutron absorbing material. The
proposed changes do not involve a physical alteration to the plant,
add any new equipment, or require existing equipment to be operated
in a manner different from the present design. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The proposed changes provide either clarification to reflect
existing plant conditions or correct typographical errors. Existing
accident analysis assumptions bound the proposed addition of not
storing fuel fragments, which may be considered as less than or
equal to a fuel assembly, in a container made with neutron absorbing
material. They have no effect on plant equipment, operating
practices or safety analysis assumptions. 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: Richard F. Locke, Esquire, Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Claudia Craig.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of amendment request: June 23, 2004.
Description of amendment request: The Humboldt Bay Power Plant,
Unit 3, is a decommissioning nuclear power plant that was permanently
shutdown in July 1976. The plant is currently in a safe storage
(SAFSTOR) condition to ensure that necessary plant systems will be
operated and maintained as needed to preserve safe conditions within
the facility to prevent deterioration until active decommissioning can
commence. All spent fuel is stored in the spent fuel pool. Currently,
the facility operating license only allows maintaining the facility in
SAFESTOR. At the time the license condition for SAFSTOR was specified,
Pacific Gas and Electric Company (PG&E), the licensee, had intended to
maintain SAFSTOR until the Department of Energy (DOE) established a
permanent repository for spent fuel. The licensee has recently
reassessed its near-term options for the facility and in December of
2003 applied for a license to store its spent fuel in an onsite dry
cask independent spent fuel storage installation (ISFSI). Moving the
spent fuel to an ISFSI would permit the licensee to begin significant
decommissioning activities. Consequently, PG&E has submitted a license
amendment request to permit the licensee to proceed with
decontamination and decommissioning activities in accordance with
applicable NRC requirements and the regulations for decommissioning
reactors in 10 CFR 50.82.
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:
[[Page 46588]]
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed change eliminates the restriction to remain in
SAFSTOR status, and allows PG&E to take actions necessary to
decommission and decontaminate the facility in accordance with NRC
regulations. The proposed change involves no changes to plant
systems or accident analysis, and as such, do not affect initiators
of analyzed events or assumed mitigation of accidents. Therefore,
the proposed changes do not increase the probability or consequences
of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different type of accident from any accident previously
evaluated?
No. The proposed change eliminates the restriction to remain in
SAFSTOR status, and allows PG&E to take actions necessary to
decommission and decontaminate the facility in accordance with NRC
regulations. The proposed change does not involve a physical
alteration to the plant, add any new equipment, or require 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
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The proposed change eliminates the restriction to remain in
SAFSTOR status, and allows PG&E to take actions necessary to
decommission and decontaminate the facility in accordance with NRC
regulations. The proposed change has no effect on plant equipment,
operating practices or safety analysis assumptions. 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: Richard F. Locke, Esquire, Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Claudia Craig.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: March 31, 2004.
Description of amendment request: The proposed change will allow
operation in regions of the power/flow map currently restricted by the
requirements of interim corrective actions (ICAs) and certain limiting
conditions for operations (LCOs) of Technical Specification 3.4.1. The
oscillation power range monitor (OPRM) will allow operations in the
regions restricted by the administrative controls mentioned above by
using inputs from the local power range monitoring (LPRM) system to
monitor core conditions and generate a reactor protection system (RPS)
trip when required to prevent a violation of the minimum critical power
ratio (MCPR) safety limit.
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. The NRC staff has reviewed the licensee's analysis
against the three standards of 10 CFR 50.92(c). The NRC staff's
analysis is presented below:
1. Does the Proposed Change Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated?
The proposed change would allow operation in regions of the
power/flow map currently restricted by administrative controls. The
purpose of the administrative controls were to ensure adequate
capability to detect and suppress conditions consistent with the
onset of a thermal-hydraulic (T-H) event which is postulated to
cause a violation of the MCPR safety limit. The mitigation of a T-H
instability event will be ensured by the RPS trip signal generated
by the OPRM prior to challenging the MCPR safety limit. Since
automatic protective functions of the OPRM will be replacing
administrative controls which require operator action, the
probability or consequence of a T-H instability event is not
significant. Therefore, the proposed change does not result in a
significant increase in the probability or consequence 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?
The proposed change would allow operation in regions of the
power/flow map currently restricted by administrative controls. The
OPRM system uses inputs from the LPRMs to monitor core conditions
and generate a RPS trip when required. Quality requirements for
software design, testing, implementation and module self-testing of
the OPRM system provide assurance that no new equipment malfunctions
due to software errors are created. The design of the OPRM system
also ensures that neither operation nor malfunction of the OPRM
system will adversely impact the operation of other systems, and no
accident or equipment malfunction of these other systems could cause
the OPRM system to malfunction or cause a different kind of
accident. Therefore, operation with the OPRM system 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?
The proposed change would allow operation in regions of the
power/flow map currently restricted by administrative controls. The
margin of safety for the unmitigated T-H instability event will not
be significantly reduced due to the capability of the OPRM to
automatically detect and suppress conditions which might result in
an MCPR safety limit violation. The automatic functions of the OPRM
will be replacing administrative controls which rely on operator
action to prevent an unmitigated T-H instability event. The OPRM
will maintain the margin of safety while significantly reducing the
burden on the control room operators. Therefore, operation with the
OPRM system does not involve a significant reduction in a margin of
safety.
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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: June 29, 2004.
Description of amendment requests: The proposed amendments would
revise the Technical Specifications (TS) to implement the following
miscellaneous changes: (1) Revise the reporting period of TS 2.2.5 from
30 days to 60 days for the safety limit violations Licensee Event
Report, (2) revise the frequency of Surveillance Requirement (SR)
3.4.3.1.2 of TS 3.4.3.1, ``Pressurizer Heatup and Cooldown Limits,'' to
reflect pressurizer spray cyclic limits being governed by the
temperature differentials between the spray nozzle and the spray line,
(3) revise TS 5.5.2.11.f.1 of TS 5.5.2.11, ``Steam Generator (SG) Tube
Surveillance Program,'' to correct typographical errors, (4) remove TS
5.5.2.14, ``Configuration Risk Management Program (CRMP),'' in
accordance with Federal Register Notice Vol. 64, No. 137 (July 19,
1999), and (5) revise TS 5.7.1.5, ``Core Operating Limits Report
(COLR),'' to delete revision numbers and dates from the referenced
documents in this section consistent with the NRC-approved industry
Technical Specifications Task Force (TSTF) Standard Technical
Specifications Traveler number TSTF-
[[Page 46589]]
363, ``Revise Topical Report References in ITS (Improved Technical
Specifications) 5.6.5 COLR,'' and incorporate editorial corrections.
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.
Southern California Edison (SCE) proposes to modify the San
Onofre Units 2 and 3 Technical Specifications (TS) to accomplish
several improvements by providing consistency with current Code of
Federal Regulations (CFR) Licensee Event Report (LER) reporting
requirements, clarifying a pressurizer heatup/cooldown Surveillance
Requirement, TS editorial corrections, removing TS redundancy to the
Maintenance Rule in accordance with Federal Register Notice Vol. 64,
No. 137 (July 19, 1999), and eliminating need for TS amendment
requests for cited Core Operating Limits Report (COLR) reference
revisions consistent with the NRC approved Industry Technical
Specifications Task Force (TSTF) Standard Technical Specifications
Traveler number TSTF-363, ``Revise Topical Report References in ITS
(Improved Technical Specifications) 5.6.5 COLR.'' These proposed
changes do not involve any change in the design or operation of the
plant. Therefore, the proposed 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.
Modifying the Technical Specifications to provide consistency
with current CFR LER reporting requirements, clarify a pressurizer
heatup/cooldown Surveillance Requirement, incorporate editorial
corrections, remove TS redundancy to the Maintenance Rule in
accordance with Federal Register Notice Vol. 64, No. 137 (July 19,
1999), and to eliminate need for TS amendment requests for cited
COLR reference revisions consistent with the NRC approved Industry
Technical Specifications Task Force (TSTF) Standard Technical
Specifications Traveler number TSTF-363, ``Revise Topical Report
References in ITS (Improved Technical Specifications) 5.6.5 COLR''
does not involve any change in the design or operation of the plant.
Therefore, a possibility of a new or different kind of accident from
any accident previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Evaluation of these proposed modifications to the Technical
Specifications to provide consistency with current CFR LER reporting
requirements, clarify a pressurizer heatup/cooldown Surveillance
Requirement, incorporate editorial corrections, remove TS redundancy
to the Maintenance Rule in accordance with Federal Register Notice
Vol. 64, No. 137 (July 19, 1999), and to eliminate need for TS
amendment requests for cited COLR reference revisions consistent
with the NRC approved Industry Technical Specifications Task Force
(TSTF) Standard Technical Specifications Traveler number TSTF-363,
``Revise Topical Report References in ITS (Improved Technical
Specifications) 5.6.5 COLR'' does not involve any change in the
design or operation of the plant and therefore does not create any
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: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: June 30, 2004.
Description of amendment requests: The proposed amendments would
revise Technical Specification (TS) 5.5.2.15, ``Containment Leakage
Rate Testing Program.'' Specifically, the licensee proposes a one-time
extension of the ten-year period of the performance-based leakage rate
testing program for Type A tests as prescribed by Nuclear Energy
Institute 94-01, Revision 0, ``Industry Guideline for Implementing
Performance-Based Option of 10 CFR Part 50, Appendix J.'' The ten-year
interval between integrated leakage rate tests is to be extended to 15
years from the previous integrated leakage rate tests. Under the
current TS requirements, which include an allowance of a 15-month
extension, the next Type A test would be performed during the Cycle 14
refueling outages currently planned for November 2005 (Unit 2) and June
2006 (Unit 3). The requested change reflects a one-time deferral of the
next Type A containment integrated leak rate test to no later than
March 30, 2010 (Unit 2) and September 9, 2010 (Unit 3). This proposed
change is based on and has been evaluated using the ``risk informed''
guidance in Regulatory Guide 1.174, ``An Approach for Using
Probabilistic Risk Assessment in Risk-informed Decisions on Plant-
Specific Changes to the Licensing Basis.''
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 revision to Technical Specifications adds a one
time extension to the current interval for Type A testing (10 CFR
50, Appendix J, Option B, Integrated Leak Rate Testing). The current
test interval of 10 years, based on past performance, would be
extended on a one time basis to 15 years from the last Type A test.
The proposed extension to Type A testing does not involve a
significant increase in the consequences of an accident since
research documented in NUREG-1493, ``Performance-Based Containment
System Leakage Testing Requirements,'' September 1995, has found
that, generically, very few potential containment leakage paths are
not identified by Type B and C tests. The NUREG concluded that
reducing the Type A testing frequency to one per twenty years was
found to lead to an imperceptible increase in risk. A high degree of
assurance is provided through testing and inspection that the
containment will not degrade in a manner detectable only by Type A
testing. The last Type A tests show leakage to be below acceptance
criteria, indicating a leak tight containment. Inspections required
by the American Society of Mechanical Engineers (ASME) Code Section
XI (Subsections IWE and IWL) and maintenance rule monitoring (10 CFR
50.65, ``Requirements for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants) are performed in order to
identify indications of containment degradation that could affect
that leak tightness. Type B and C testing required by Technical
Specifications will identify any containment opening such as valves
that would otherwise be detected by the Type A tests. These factors
show that a Type A test extension will not represent a significant
increase in the consequences of an accident.
Therefore, the proposed changes do 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 revision to Technical Specifications adds a one
time extension to the current interval for Type A testing (10 CFR
50, Appendix J, Option B, Integrated Leak Rate Testing). The current
test interval of 10 years, based on past performance, would be
extended on a one time basis to 15 years from the last Type A test.
The proposed extension to Type A testing cannot create the
possibility of a new or different type of accident since there are
no physical changes being made to the plant and there are no changes
to the operation of the plant that could introduce a new failure
mode creating
[[Page 46590]]
an accident or affecting the mitigation of an accident. Therefore,
the proposed changes do 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 revision to Technical Specifications adds a one
time extension to the current interval for Type A testing (10 CFR
50, Appendix J, Option B, Integrated Leak Rate Testing). The current
test interval of 10 years, based on past performance, would be
extended on a one time basis to 15 years from the last Type A test.
The proposed extension to Type A testing will not significantly
reduce the margin of safety. The NUREG 1493, ``Performance-Based
Containment System Leakage Testing Requirements,'' September 1995,
generic study of the effects of extending containment leakage
testing found that a 20 year extension in Type A leakage testing
resulted in an imperceptible increase in risk to the public. NUREG
1493 found that, generically, the design containment leakage rate
contributes about 0.1 percent to the individual risk and that the
decrease in Type A testing frequency would have a minimal affect on
this risk since 95% of the potential leakage paths are detected by
Type C testing. Regular inspections required by the American Society
of Mechanical Engineers (ASME) Code Section XI (Subsections IWE and
IWL) and maintenance rule monitoring (10 CFR 50.65, ``Requirements
for Monitoring the Effectiveness of Maintenance at Nuclear Power
Plants) will further reduce the risk of a containment leakage path
going undetected.
Therefore the proposed change does not involve a 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: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Southern Nuclear Operating Company, Inc. Docket Nos. 50-348 and 50-364,
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
Date of amendment request: June 28, 2004.
Description of amendment request: The proposed amendments would
revise existing Technical Specifications (TSs) 3.4.13, ``RCS [Reactor
Coolant System] Operational Leakage,'' TS 5.59, ``Steam Generator [SG]
Tube Surveillance Program,'' and TS 5.610, ``Steam Generator Tube
Inspector Report.'' It would also add a new TS 3.4.17, ``Steam
Generator Tube Integrity.'' These changes would facilitate the
implementation of industry initiative NEI [Nuclear Energy Institute]
97-06, ``Steam Generator Program Guidelines,'' which would allow for a
comprehensive, performance-based approach to managing SG performance at
Farley Nuclear Plant, Units 1 and 2.
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 operating conditions (including startup, operation in
the power range, hot standby, cooldown and all anticipated
transients included in the design specification). The SG performance
criteria are based on tube structural integrity, accident induced
leakage, and operational LEAKAGE.
The structural integrity performance criterion is:
``All inservice SG tubes shall retain structural integrity over
the full range of normal operating conditions (including startup,
operation in the power range, hot standby and cooldown and all
anticipated transients included in the design specification) and
design basis accidents. This includes retaining a safety factor of
3.0 against burst under normal steady state full power operation
primary to secondary pressure differential and a safety factor of
1.4 against burst applied to the design basis accident primary to
secondary pressure differentials. Apart from the above requirements,
additional loading conditions associated with the design basis
accidents, or combination of accidents in accordance with the design
and licensing basis, shall also be evaluated to determine if the
associated loads contribute significantly to burst or collapse. In
the assessment of tube integrity, those loads that do significantly
affect burst or collapse shall be determined and assessed in
combination with the loads due to pressure with a safety factor of
1.2 on the combined primary loads and 1.0 on axial secondary
loads.''
The accident induced leakage performance criterion is:
``The primary to secondary accident induced leakage rate for all
design basis accidents, other than a 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. For FNP Units 1 and 2, leakage is not to exceed 1 gpm [gallons
per minute] total for all three SGs. Exceptions to the 1 gpm limit
can be applied if approved by the NRC in conjunction with approved
alternate repair criteria.''
The operational LEAKAGE performance criterion is:
The RCS operational primary to secondary LEAKAGE through any one
SG shall be limited to 150 gpd [gallons per day].
A steam generator tube rupture (SGTR) event is one of the design
basis accidents analyzed as part of the plant licensing basis. In
the analysis of a SGTR event, a bounding primary to secondary
LEAKAGE rate equal to the operational LEAKAGE rate limits in the
licensing basis plus the LEAKAGE rate associated with a double-ended
rupture of a single tube is assumed.
For other design basis accidents such as main steam line break
(MSLB), rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). For FNP Units 1 and 2, these analyses
assume that primary to secondary LEAKAGE for all SGs is 1 gpm. The
accident induced leakage criterion introduced by the proposed
changes accounts for tubes that may leak during design basis
accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed in this change to the TS
identify the standards against which tube integrity is to be
measured. Meeting the performance criteria provides reasonable
assurance that the SG tubing will remain capable of fulfilling its
specific safety function of maintaining reactor coolant pressure
boundary integrity throughout each operating cycle and in the
unlikely event of a design basis accident. The performance criteria
are only a part of the Steam Generator Program required by the
proposed change to the TS. The program, defined by NEI 97-06, Steam
Generator Program Guidelines, includes a framework that incorporates
a balance of prevention, inspection, evaluation, plugging, and
leakage monitoring.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in the TS for operational leakage and
for DOSE EQUIVALENT I-131 in primary coolant to ensure the plant is
operated within its analyzed condition. The analysis of the limiting
design basis accident assumes that primary to secondary leak rate
after the accident is 1 gpm with no more than 500 gpd in any one SG,
and that the reactor coolant activity levels of DOSE EQUIVALENT I-
131 are at the technical specification values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TS and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TS.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of a MSLB,
[[Page 46591]]
rod ejection, or a reactor coolant pump locked rotor event.
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 TS.
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
result of the implementation of the Steam Generator Program will be
an enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the Steam Generator Program to manage SG
tube inspection, assessment 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 TS.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: Stephanie M. Coffin, Acting.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: April 26, 2004.
Description of amendment request: The proposed amendments would
revise the Technical Specification Section 5.5.12, ``Primary
Containment Leakage Rate Testing Program'' to reflect a one-time
deferral of the Type A Containment Integrated Leak Rate Test (ILRT).
This change would extend the 10 year interval between ILRTs to 15 years
from the previous ILRT.
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 Technical Specification change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed revision to Technical Specification 5.5.12
(``Primary Containment Leakage Rate Testing Program'') involves a
one-time extension to the current interval for Type A containment
testing. The current test interval of ten (10) years would be
extended on a one-time basis to no longer than fifteen (15) years
from the last Type A test. The proposed Technical Specification
change does not involve a physical change to the plant or a change
in the manner which the plant is operated or controlled. The reactor
containment is designed to provide an essentially leak tight barrier
against the uncontrolled release of radioactivity to the environment
for postulated accidents. As such the reactor containment itself and
the testing requirements invoked to periodically demonstrate the
integrity of the reactor containment exist to ensure the plant's
ability to mitigate the consequences of an accident, and do not
involve the prevention or identification of any precursors of an
accident. Therefore, the proposed Technical Specification change
does not involve a significant increase in the probability of an
accident previously evaluated.
The proposed change involves only the extension of the interval
between Type A containment leakage tests. Type B and C containment
leakage tests will continue to be performed at the frequency
currently required by plant Technical Specifications. Industry
experience has shown, as documented in NUREG-1493, that Type B and C
containment leakage tests have identified a very large percentage of
containment leakage paths and that the percentage of containment
leakage paths that are detected only by Type A testing is very
small. HNP [Hatch Nuclear Plant ] Unit 2 ILRT test history supports
this conclusion. NUREG-1493 concluded, in part, that reducing the
frequency of Type A containment leak tests to once per twenty (20)
years leads to an imperceptible increase in risk. The integrity of
the reactor containment is subject to two types of failure
mechanisms which can be categorized as (1) activity based and (2)
time based. Activity based failure mechanisms are defined as
degradation due to system and/or component modifications or
maintenance. Local leak rate test requirements and administrative
controls such as design change control and procedural requirements
for system restoration ensure that containment integrity is not
degraded by plant modifications or maintenance activities. The
design and construction requirements of the reactor containment
itself combined with the containment inspections performed in
accordance with ASME [American Society of Mechanical Engineers]
Section XI, the Maintenance Rule and the containment coatings
program serve to provide a high degree of assurance that the
containment will not degrade in a manner that is detectable only by
Type A testing. Therefore, the proposed Technical Specification
change does not involve a significant increase in the consequences
of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed revision to the Technical Specifications involves a
one-time extension to the current interval for Type A containment
testing. The reactor containment and the testing requirements
invoked to periodically demonstrate the integrity of the reactor
containment exist to ensure the plant's ability to mitigate the
consequences of an accident and do not involve the prevention or
identification of any precursors of an accident. The proposed
Technical Specification change does not involve a physical change to
the plant or the manner in which the plant is operated or
controlled. Therefore, the proposed Technical Specification change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The proposed revision to Technical Specifications involves a
one-time extension to the current interval for Type A containment
testing. The proposed Technical Specification change does not
involve a physical change to the plant or a change in the manner in
which the plant is operated or controlled. The specific requirements
and conditions of the Primary Containment
[[Page 46592]]
Leakage Rate Testing Program, as defined in Technical
Specifications, exist to ensure that the degree of reactor
containment structural integrity and leak-tightness that is
considered in the plant safety analysis is maintained. The overall
containment leakage rate limit specified by Technical Specifications
is maintained. The proposed change involves only the extension of
the interval between Type A containment leakage tests. Type B and C
containment leakage tests will continue to be performed at the
frequency currently required by plant Technical Specifications.
HNP Unit 2 and industry experience strongly supports the
conclusion that Type B and C testing detects a large percentage of
containment leakage paths and that the percentage of containment
leakage paths that are detected only by Type A testing is small. The
containment inspections performed in accordance with ASME Section
XI, the Maintenance Rule and the Coatings Program serve to provide a
high degree of assurance that the containment will not degrade in a
manner that is detectable only by Type A testing. Additionally, the
on-line containment monitoring capability that is inherent to
inerted BWR containments allows for the detection of gross
containment leakage that may develop during power operation. The
combination of these factors ensures that the margin of safety that
is inherent in plant safety analysis is maintained. Therefore, the
proposed Technical Specification 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Stephanie M. Coffin, Acting.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: June 22, 2004.
Description of amendment request: The proposed amendments would
revise the Technical Specification (TS), Appendix A in order to change
the frequency of the logic system functional test, for the 4 kV
emergency busses' loss of power instrumentation, from once every 18
months to once every 24 months.
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.
This is a proposed change to the surveillance requirement (SR)
for the logic system functional test (LSFT) of the loss of power
(LOP) instrumentation for Plant Hatch Units 1 and 2 (SR 3.3.8.1.4).
The LOP instrumentation functions to monitor the voltage on the 4 kV
emergency busses and, if necessary, to disconnect these busses from
the offsite power source and re-connect them to on-site power. This
would, of course, be necessary if a bus experienced a loss of, or a
degraded, voltage. This ensures an adequate response to a loss of
coolant accident (LOCA) if that accident were to occur
simultaneously with a loss of off-site power (LOSP). The probability
of occurrence of a previously evaluated event, such as a LOCA/LOSP,
will not increase since the LOP instrumentation is not being
physically altered as a result of this change in such a manner which
may increase the likelihood of failure. In fact, it is not being
physically altered at all as a result of this submittal.
Additionally, no other safety related equipment or components
designed to prevent the occurrence of a previously evaluated event
are being physically altered or otherwise affected as a result of
this TS change request.
The consequences of a previously evaluated event will not
increase as a result of revising the surveillance frequency for the
LOP instrumentation. Review of surveillance histories demonstrates
adequate performance for the LOP relays in ultimately connecting the
emergency power sources to the distribution bus, justifying the
revision in the surveillance frequency. Therefore, the LOP
instrumentation can be reasonably expected to perform its function
in a LOCA/LOSP event, even with the revised frequency for the LSFT.
For the above reasons, the change in the LSFT frequency does not
involve a significant increase in the probability or consequences of
a previously evaluated event.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The LOP instrumentation is not being physically altered.
Furthermore, its operation and maintenance will remain within the
design bases. The only proposed change is the frequency of the logic
system functional test. Since no new modes of operation are being
introduced, a new or different kind of accident from any previously
evaluated is not created.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The function of the LOP instrumentation is to ensure that the
emergency power distribution busses receive adequate power from
either the off-site or on-site sources. The LOP relays will initiate
a transfer of the emergency 4 kV busses to the on-site diesel
generators on a loss of coolant accident with a concurrent loss of
off-site power. The diesel logic will then sequence the cooling
water pumps and other safety related equipment onto their respective
emergency bus. This sequencing of loads is tested by a different
surveillance requirement which is not affected by this TS change
request and has already been revised to a frequency of once per 24
months. This proposed TS revision only changes the frequency of
performance of the LSFT for the LOP instrumentation. A review of
surveillance histories shows that these relays perform adequately in
the re-connection of the emergency busses to the on-site power
source. Some problems have been noted in the history review with the
loss of off-site power annunciation. However, the annunciator does
not affect the safety function of providing power to the
distribution bus.
For the above reasons, the margin of safety is not reduced by
this proposed Technical Specifications change.
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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Stephanie M. Coffin, Acting.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant (BFN), Units 2 and 3, Limestone County, Alabama
Date of amendment request: July 8, 2004 (TS-448)
Description of amendment request: The proposed amendment requests
the modification of Technical Specification Section 5.5.12 ``Primary
Containment Leakage Rate Testing Program'' to allow a one-time 5-year
extension to the 10-year frequency of the performance-based leakage
rate testing program for Type A tests. The proposed changes are
submitted on a risk-informed basis as described in Regulatory Guide
1.174, An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis.
The risk-informed analysis supporting the proposed changes indicates
that the increase in risk from extending the integrated leak rate test
interval from 10 to 15 years is insignificant.
Basis for proposed no significant hazards consideration
determination:
[[Page 46593]]
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:
TVA has evaluated whether or not a significant hazards
consideration is involved with the proposed amendment by focusing on
the three standards set forth in 10 CFR 50.92, ``Issuance of
Amendment,'' as discussed below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed revision to TS adds a one-time extension to the
current interval for Type A testing. The current test interval of 10
years, based on past performance, would be extended on a one-time
basis to 15 years from the last Type A test. The proposed extension
to Type A testing cannot increase the probability of an accident
previously evaluated since the containment Type A testing extension
is not a modification and the test extension is not of a type that
could lead to equipment failure or accident initiation.
The proposed extension to Type A testing does not involve a
significant increase in the consequences of an accident since
research documented in NUREG-1493 has found that, generically, very
few potential containment leakage paths are not identified by Type B
and C tests. The NUREG concluded that reducing the Type A (ILRT)
testing frequency to once per 20 years was found to lead to an
imperceptible increase in risk. These generic conclusions were
confirmed by a plant specific risk assessment.
Testing and the containment inspection programs in place at BFN
provide a high degree of assurance that the containment will not
degrade in a manner detectable only by Type A testing. The last four
Type A tests show leakage to be below acceptance criteria,
indicating a very leak tight containment. Type B and C testing
required by TS will identify any containment opening such as valves
that would otherwise be detected by the Type A tests. Inspections,
including those required by the American Society of Mechanical
Engineers code are also performed in order to identify indications
of containment degradation that could affect that leak tightness.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No. The change does not create the possibility of a new or
different kind of accident from any accident previously analyzed.
The proposed revision to TS adds a one-time extension to the current
interval for Type A testing. The current test interval of 10 years,
based on past performance, would be extended on a one-time basis to
15 years from the last Type A test. The proposed extension to Type A
testing cannot create the possibility of a new or different type of
accident since there are no physical changes being made to the plant
and there are no changes to the operation of the plant that could
introduce a new failure mode creating an accident or affecting the
mitigation of an accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. BFN Units 2 and 3 are General Electric BWR/4 plants with
Mark I primary containments. The Mark I primary containment consists
of a drywell, which encloses the reactor vessel; reactor coolant
recirculation system and branch lines of the Reactor Coolant System;
a toroidal-shaped pressure suppression chamber containing a large
volume of water; and a vent system connecting the drywell to the
water space of the suppression chamber. The primary containment is
penetrated by personnel access hatches, piping, and electrical
penetrations.
The integrity of the primary containment penetrations and
isolation valves is verified through Type B and Type C local leak
rate tests and the overall leak-tight integrity of the primary
containment is verified by a Type A integrated leak rate test as
required by 10 CFR 50, Appendix J, ``Primary Reactor Containment
Leakage Testing for Water-Cooled Power Reactors.'' These tests are
performed to verify the essentially leak-tight characteristics of
the primary containment at the design basis accident pressure. The
proposed change for a one-time extension of the Type A tests does
not affect the method for Type A, B, or C testing, or the test
acceptance criteria. In addition, based on previous Type A testing
results, TVA does not expect additional degradation during the
extended period between Type A tests, which would result in a
significant reduction in a margin of 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 request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Acting Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: July 8, 2004.
Description of amendment request: The proposed amendment will
revise the Technical Specification (TS) to remove the term ``inter-
rack'' and associated wording from Surveillance Requirements 3.8.4.6
and 3.8.4.10 for the 125 Volt (V) Direct Current (DC) Electrical Power
Subsystems of the Emergency Diesel Generators (DGs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The proposed TS change eliminates an inaccurate term and
associated wording, but the actual TS amendment does not result in
any change to the actual surveillance field test for the associated
batteries. The proposed wording will only clarify the surveillances.
Prior field tests were adequate to verify proper battery connection
integrity since it tested the inside (inter-tier) jumper cable
connections as if they were interchangeable with inter-rack.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed TS change does not alter the configuration of
the plant's 125 V DC Electrical Power Subsystems of the Emergency
DGs. The change does not directly affect plant operation. The change
will not result in the installation of any new equipment or system
or the modification of any existing equipment or systems. No new
operations procedures, conditions, or modes will be created by this
proposed change. 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
margin of safety?
No. The battery connection continuity check for the 125 V DC
Electrical Power Subsystems of the Emergency DGs will continue to be
monitored by the same process as previously performed. 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Acting Section Chief: Michael L. Marshall, Jr.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the
[[Page 46594]]
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 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].
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek
Nuclear Generating Station (OCNGS), Ocean County, New Jersey, Docket
No. 50-289, Three Mile Island Nuclear Station, Unit 1 (TMI-1), Dauphin
County, Pennsylvania
Date of application for amendments: March 8, 2004.
Brief description of amendment: The amendments deleted the License
Condition entitled ``Long Range Planning Program'' from the OCNGS and
TMI-1 operating licenses. In addition, for TMI-1, the amendment
relocated a requirement (regarding surveillance of the depth of water
in the spent fuel pool) from the Long Range Planning Program to the
Technical Specifications.
Date of Issuance: July 13, 2004.
Effective date: These license amendments are effective as of their
date of issuance, and shall be implemented within 30 days of issuance.
Amendment Nos.: 244 and 250
Facility Operating License Nos. DPR-16 and DPR-50: Amendments
revised the Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19563 and 19564). The Commission's related evaluation of this amendment
is contained in a Safety Evaluation dated July 13, 2004.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: August 6, 2003, as supplemented
February 13 and June 16, 2004.
Brief description of amendment: The amendment revised the reactor
building tendon surveillance criteria to incorporate a reference to
Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a.
The amendment also includes an administrative change to provide
consistency between Technical Specification Definition 1.22 (MEMBERS OF
THE PUBLIC) and the definition contained in 10 CFR 20.1003, and a
change to correct a typographical error in a reference title.
Date of issuance: July 13, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 251.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68655) and March 16, 2004 (69 FR 12363). The February 13, 2004,
supplemental letter provided clarifying information and expanded the
scope of the application as originally noticed. Therefore, the original
proposed no significant hazards consideration determination was changed
and republished. The June 16, 2004, supplement provided clarifying
information, did not expand the scope of the application and did not
change the NRC staff's proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 13, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: December 24, 2003.
Brief description of amendment: The amendment deleted requirements
from the Technical Specifications (TSs) 3.7.A.7.c and 4.7.A.7.c
associated with hydrogen analyzers. The associated TS Bases are also
deleted.
Date of issuance: July 22, 2004.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 206.
Facility Operating License No. DPR-35: The amendment revised the
TSs.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19568).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 22, 2004.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: August 14, 2003, as supplemented by
letters dated January 22, and May 6, 2004.
Description of amendment request: This license amendment modifies
Technical Specification (TS) Table 3.3.6.1-1, ``Primary Containment and
Drywell Isolation Instrumentation,'' Item 1.f, to increase the
analytical limit for detected temperature and the resulting TS
Allowable Value related to the setpoint for the Main Steam Line Turbine
Building Temperature--High system isolation function. Additionally, it
authorizes the use of the GOTHIC 7.0 computer program to perform
analyses of main steamline leaks in the turbine building for Perry
Nuclear Power Plant to replace the currently approved COMPARE computer
program for performing the analyses listed above.
Date of issuance: July 9, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 130.
[[Page 46595]]
Facility Operating License No. NPF-58: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: (69 FR 696) January 6,
2004.
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 9, 2004.
No significant hazards consideration comments received: No.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: February 27, 2004.
Brief description of amendment: The amendment deletes Technical
Specification Section 5.6.2.6, ``Post-Accident Sampling,'' requirements
to maintain a Post-Accident Sampling System.
Date of issuance: July 6, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 213.
Facility Operating License No. DPR-72: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19571).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 6, 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, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: October 23, 2002, as
supplemented by letters dated August 28, 2003, December 11, 2003,
February 3, 2004, and March 25, 2004.
Brief description of amendments: These amendments revised Technical
Specification Section 5.6, ``Design Features--Fuel Storage,'' for St.
Lucie Units 1 and 2 to include the design of a new cask pit spent fuel
storage rack for each unit, and increase each unit's spent fuel storage
capacity by combining the cask pit rack and existing spent fuel pool
storage rack capacities. The cask pit racks will be used to store spent
fuel to allow refueling outage fuel offloads and nonoutage fuel
shuffles and, for Unit 1, to store new fuel prior to loading it into
the reactor.
Date of Issuance: July 9, 2004.
Effective Date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 192 and 135.
Renewed Facility Operating License Nos. DPR-67 and NPF-16:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 28, 2003 (68 FR
4244), as corrected March 31, 2003 (68 FR 15487). The August 28, 2003,
December 11, 2003, February 3, 2004, and March 25, 2004, 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 an Environmental Assessment dated July 2, 2004 and in a Safety
Evaluation dated July 9, 2004.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 29, 2004, as supplement by
letter dated April 8, 2004.
Brief description of amendment: The amendment revises Technical
Specification 3.4.9 Pressure Temperature (P/T) limit curve Figures
3.4.9-1, 3.4.9-2, and 3.4.9-3.
Date of issuance: July 14, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 204.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 16, 2004 (69 FR
12371). The April 8, 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 amendment is contained
in a Safety Evaluation dated July 14, 2004.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 30, 2004, as supplemented by
letter dated June 17, 2004.
Brief description of amendment request: The proposed amendment
would revise the Cooper Nuclear Station (CNS) Technical Specifications
(TSs), by adding a temporary note to allow a one-time extension of a
limited number of TS Surveillance Requirements (SRs). The temporary
note states that the next required performance of the SRs may be
delayed until the current cycle refueling outage, but no later than
February 2, 2005, and it expires upon startup from the refueling
outage. With the exception of one SR, the period of additional time
requested occurs during the next planned refueling outage.
Date of issuance: July 14, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 205.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 12, 2004 (69
FR 7023). The June 17, 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 amendment is contained
in a Safety Evaluation dated July 14, 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: September 24, 2002, and its
supplements dated November 21, 2003, and March 9, 2004.
Brief description of amendments: The amendments revise Technical
Specification (TS) Section 3.4.11, ``Pressurizer Power Operated Relief
Valves (PORVs),'' to credit the automatic actuation of the pressurizer
PORVs for mitigating the plant transient of inadvertent actuation of
the safety injection (SI) system. The amendments also modify the
wording in Criteria A, B, and E of TS 3.4.11 to reflect the new
requirement of ensuring automatic function of PORVs and adds two new
surveillance requirements. The licensee withdrew the changes to TS
3.4.10, ``Pressurizer Safety Valves,'' in its letter dated March 9,
2004.
Date of issuance: July 2, 2004.
Effective date: July 2, 2004, and shall be implemented within 30
days from the date of issuance.
[[Page 46596]]
Amendment Nos.: Unit 1--171; Unit 2--172.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 24, 2002 (67
FR 78522)
The November 21, 2003, and March 9, 2004, supplemental letters
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 amendments is contained
in a Safety Evaluation dated July 2, 2004.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: September 19, 2003.
Brief description of amendment: This amendment revised Surveillance
Requirement 4.2.4.2 to specifically identify the Power Distribution
Monitoring System being used in determining the Quadrant Power Tilt
Ratio with one inoperable Power Range Channel.
Date of issuance: July 6, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 168.
Renewed Facility Operating License No. NPF-12: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: March 30, 2004 (69 FR
16623).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 6, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Units 1 and 2, Hamilton County, Tennessee
Date of application for amendment: March 5, 2004.
Brief description of amendment: The amendment revises the reactor
coolant pump flywheel inspection interval from 10 years to 20 years.
Date of issuance: July 8, 2004.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: 293 and 283.
Facility Operating License No. DPR-77 and DPR-79: Amendment revises
the technical specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19577).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 8, 2004.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: April 8, 2004.
Brief description of amendment: The amendment revises TS 5.5.7,
``Reactor Coolant Pump Flywheel Inspection Program,'' to increase the
inspection interval from 10 years to 20 years.
Date of issuance: July 12, 2004.
Effective date: July 12, 2004, and shall be implemented within 90
days from the date of issuance.
Amendment No.: 163.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 11, 2004 (69 FR
26193).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 12, 2004.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: April 30, 2003, as supplemented by
letters dated December 18, 2003, and April 13, 2004.
Brief description of amendment: The amendment revises several
surveillance requirements (SRs) in Technical Specification (TS) 3.8.1
on alternating current sources for plant operation. The revised SRs
have notes deleted or modified to allow the SRs to be performed, or
partially performed, in reactor modes that previously were not allowed
by the TSs. The proposed changes to SRs 3.8.4.7 and 3.8.4.8 for direct
current sources were withdrawn by letter dated April 13, 2004.
Date of issuance: July 12, 2004.
Effective date: July 12, 2004, and shall be implemented within 90
days of the date of issuance including the incorporation of the changes
to the TS Bases for TS 3.8.1 as described in the licensee's letters
dated April 30 and December 18, 2003, and April 13, 2004.
Amendment No.: 154.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34673).
The December 18, 2003, and April 13, 2004, supplemental letters
provided additional clarifying information, did not expand the scope of
the application as 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 July 12, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 26th day of July 2004.
For the Nuclear Regulatory Commission.
James E. Lyons,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 04-17346 Filed 8-2-04; 8:45 am]
BILLING CODE 7590-01-P