[Federal Register Volume 69, Number 216 (Tuesday, November 9, 2004)]
[Notices]
[Pages 64984-64996]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-24804]
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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, October 15, 2004, through October 28, 2004.
The last biweekly notice was published on October 26, 2004 (69 FR
62467).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. (Note:
Public access to ADAMS has been temporarily suspended so that security
reviews of publicly available documents may be performed and
potentially sensitive information removed. Please check the NRC Web
site for updates on the resumption of ADAMS access.) The filing of
requests for a hearing and petitions for leave to intervene is
discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
[[Page 64985]]
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/. (Note: Public access to ADAMS has been
temporarily suspended so that security reviews of publicly available
documents may be performed and potentially sensitive information
removed. Please check the NRC Web site for updates on the resumption of
ADAMS access.) 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. (Note: Public access to ADAMS has been
temporarily suspended so that security reviews of publicly available
documents may be performed and potentially sensitive information
removed. Please check the NRC Web site for updates on the resumption of
ADAMS access.) 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].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: September 16, 2004.
Description of amendment request: The proposed amendment would
revise the scope and the frequency of Surveillance Requirement (SR)
3.7.6.1 for verification of one complete cycle of each turbine bypass
valve (TBV) every 92 days. The proposed change to SR 3.7.6.1 would
allow a 5 percent stroke rather than a complete (100 percent) stroke of
each TBV, and would extend the surveillance frequency from 92 days to
120 days. The complete stroke verification currently required by SR
3.7.6.1 once after each entry into MODE 4 would be retained and
renumbered SR 3.7.6.2.
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:
[[Page 64986]]
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change to Technical Specification Surveillance
Requirement (SR) 3.7.6.1 will allow a 5% stroke rather than a
complete (100%) stroke of each turbine bypass valve (TBV), and will
extend the surveillance frequency from 92 days to 120 days. The
requirement to verify one complete cycle of each TBV once after each
entry into MODE 4 will be retained.
The proposed testing requirements will provide a level of
assurance, equivalent to that which now exists, that the TBVs will
remain operable throughout the operating cycle, and that they will
be able to perform their intended safety function if called upon to
do so. Additionally, the reduction in the potential for plant
transients that can result from the current testing requirements,
will more than offset the small increase (less than one half of one
percent) in TBV failure probability per cycle with the proposed
testing regime. Thus the proposed changes will not significantly
increase the probability of an accident previously evaluated.
Fermi 2 is analyzed for the increase in reactor pressure
transient events with the assumption that the Main Turbine Bypass
System (MTBS) is out-of-service. Feedwater Controller Failure
Upscale represents the most limiting event in this analytical
category, and provides the basis for the Minimum Critical Power
Ratio (MCPR) operating limits that are applicable when the MTBS is
out of service. Because the proposed testing requirements do not
alter the assumptions for any of the increase in pressure transient
events, the radiological consequences of an accident previously
evaluated are not increased.
Therefore, this proposed amendment will not involve a
significant increase in the probability or the consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not significantly affect the assumed
performance of the TBVs, nor does it affect any other plant systems,
structures, or components. In fact, these changes reduce the
possibility of secondary plant transients and the potential for
recirculation pump runbacks during the performance of this SR while
at power. The proposed changes do not install any new plant
equipment, nor is installed plant equipment being operated in a new
or different manner. The proposed changes in test frequency and
methodology will continue to ensure that the TBVs remain capable of
performing their intended safety function. Therefore, this proposed
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The proposed change will modify the scope and the frequency of
the quarterly full stroke test of the TBVs. The operability
requirements and functional characteristics of the TBVs remain
unchanged. The proposed change to SR 3.7.6.1 from full stroke
testing to 5% stroke testing, and from 92 days to 120 days has been
evaluated to produce only a minimal increase in the failure
probability of a TBV during each cycle (less than one half of one
percent). This failure probability increase is outweighed by the
reduction in the potential for plant transients resulting from full
stroke testing during power operation. Both Alstom's sensitivity
study, and actual industry experience at Ringhals Units 1 and 2 have
shown that a partial stroke test will ensure that the valves remain
mechanically operable throughout the operating cycle. The Alstom
study further shows that a partial stroke test at 120 days, rather
than at 92 days, will ensure that the valves remain mechanically
operable throughout the operating cycle. Additionally, retaining the
requirement to full stroke test each TBV once after each entry into
MODE 4 will continue to verify that the valves are mechanically
operable prior to their first use following each startup from MODE
4. The TBV response times are used in determining the effect on the
MCPR. The surveillance test that ensures the MTBS meets the system's
response time limits (SR 3.7.6.3) is not affected by these proposed
changes and will continue to be performed at its current 18 month
frequency. Therefore, this proposed change will 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: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: October 7, 2004.
Description of amendment request: The proposed amendment would
revise the Safety Limit Minimum Critical Power Ratio in Technical
Specification 2.1.1.2 to reflect the results of cycle-specific
calculations performed for Fermi 2 operating Cycles 10 and 11.
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 basis of the Safety Limit Minimum Critical Power Ratio
(SLMCPR) is to ensure no mechanistic fuel damage is calculated to
occur if the limit is not violated. The new CPR value preserves the
existing margin to transition boiling and probability of fuel damage
is not increased. The derivation of the revised SLMCPR for Fermi 2
for incorporation into the Technical Specifications, and its use to
determine plant and cycle-specific thermal limits, have been
performed using NRC approved methods. These plant-specific
calculations are performed each operating cycle and if necessary,
will require future changes to these values based upon revised core
designs. The revised SLMCPR values do not change the method of
operating the plant and have no effect on the probability of an
accident initiating event or transient.
Therefore, this proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change results only from a specific analysis for
the Fermi 2 Cycle 10 and 11 cores. This change does not involve any
new or different methods for operating the facility. No new
initiating events or transients result from these changes.
Therefore, this proposed amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The new SLMCPR is calculated using NRC approved methods with
plant and cycle-specific parameters for the Cycle 10 and 11 core
designs. The SLMCPR value is established to ensure that greater than
99.9% of all fuel rods in the core will avoid transition boiling if
the limit is not violated, thereby preserving the fuel cladding
integrity. The operating MCPR limit is set appropriately above the
safety limit value to ensure adequate margin when the cycle-specific
transients are evaluated. Accordingly, the margin of safety is
maintained with the revised values. Therefore, this proposed
amendment does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
[[Page 64987]]
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: August 18, 2004.
Description of amendment request: The proposed amendment would
correct an inadvertent technical specification (TS) change associated
with TS Amendment 184/166 and 182/164. Licensing Amendment 182/164
deleted the safety injection steam line pressure-low (SLPL) function
and all concerned references due to redundant safety injection signals.
This amendment was approved on September 22, 1998. As part of the
conversion to standardized TS (STS), Amendment 184/166, all concerned
references to the SLPL function were not correctly deleted from STS
3.3.2. Specifically, a reference to the SLPL function was not deleted
from Footnote (c) to STS Table 3.3.2-1 and from the Basis of STS 3.3.2
Function 4.d.(1). Amendment (184/166) was approved on September 30,
1998.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Does This LAR Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated?
No. Approval and implementation of this LAR will have no effect
on accident probabilities or consequences since the proposed changes
are consistent with those previously reviewed and approved by the
NRC in TS Amendment 182/164.
Criterion 2--Does This LAR Create the Possibility of a New or Different
Kind of Accident From Any Accident Previously Evaluated?
No. This LAR does not involve any physical changes to the plant.
Therefore, no new accident causal mechanisms will be generated. The
proposed changes are consistent with those previously reviewed and
approved by the NRC in TS Amendment 182/164. Consequently, plant
accident analyses will not be affected by these changes.
Criterion 3--Does This LAR Involve a Significant Reduction in a Margin
of Safety?
No. Margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following accident conditions. These barriers include the
fuel cladding, the reactor coolant system, and the containment
system. The performance of these barriers will not be affected by
the proposed changes since they are consistent with those previously
reviewed and approved by the NRC in TS Amendment 182/164. Therefore,
the proposed changes in this license amendment will not result in a
significant reduction in the facility's margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: October 12, 2004.
Description of amendment request: The proposed license amendment
request would change the Final Safety Analysis Report (FSAR) to reflect
that the reactor core isolation cooling (RCIC) system is not required
to mitigate the consequences of the control rod drop accident (CRDA).
The FSAR revision would clarify that although the RCIC system is
designed to initiate and inject into the reactor pressure vessel (RPV)
at a low water level (L2), the additional RPV inventory is not required
to prevent the accident or to mitigate the consequences of the CRDA.
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 change clarifies, in various sections of the FSAR, that
RCIC system operation is not required in order to mitigate the
consequences of the CRDA. The proposed change involves no changes to
plant systems or accident analyses. The accident analysis for the
CRDA demonstrates that core design, the control rod pattern
controls, and the scram signal from the reactor protection system
(RPS) effectively prevent damage to the fuel rods as a result of the
dropped rod. Furthermore, based on a prescribed source term provided
from an assumed damage to less than 2% fuel in the core, the
resulting radiological consequences are not affected by RCIC
operation or failure to operate. As such, the change does not affect
initiation of analyzed events or assumed mitigation of accidents or
transients. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This change clarifies, in various sections of the FSAR, that the
RCIC system operation is not required in order to mitigate the
consequences of the CRDA. The proposed change does not involve a
physical alteration of the plant, add any new equipment, or require
any existing equipment to be operated in a manner different from the
present design. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This change clarifies, in various sections of the FSAR, that the
RCIC system operation is not required in order to mitigate the
consequences of the CRDA. The change has no effect on plant systems,
operating practices or safety analyses assumptions. For these
reasons, 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: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: September 30, 2004.
Description of amendment request: The proposed amendment would
change the existing steam generator tube surveillance program to be
consistent with that being proposed by the Technical Specifications
Task Force (TSTF) in TSTF-449, Draft Revision 2. These proposed changes
would revise the Technical Specifications and Bases for Specifications
3.4.13, RCS [Reactor Coolant System] Operational LEAKAGE, Specification
5.5.9, Steam Generator (SG) Tube Surveillance Program, and
Specification 5.6.7, Steam Generator Tube Surveillance Reports, and add
a new Specification 3.4.16 entitled Steam Generator (SG) Tube
Integrity. Also, as a result of the licensee replacing the SGs with SGs
having a new Alloy 690 thermally treated tubing design, the Technical
Specifications and Bases would be revised to reflect this replacement.
[[Page 64988]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change requires a Steam Generator Program that
includes performance criteria that will provide reasonable assurance
that the steam generator (SG) tubing will retain integrity over the
full range of design basis operating conditions (including startup,
power operation, hot standby, cooldown, anticipated transients and
postulated accidents). The SG performance criteria are based on tube
structural integrity, accident induced leakage, and operational
LEAKAGE. These criteria assure that the probability of an accident
will not be increased.
The primary to secondary accident induced leakage rate for any
design basis accidents, other than an SG tube rupture, shall not
exceed the leakage rate assumed in the accident analysis in terms of
total leakage rate for all SGs and leakage rate for an individual
SG. [The primary to secondary accident induced leakage rate is
relatively inconsequential for the SG tube rupture analysis.] The
operational LEAKAGE performance criterion meets current NRC
regulations and NEI [Nuclear Energy Institute] 97-06 criteria for
reactor coolant system (RCS) operational primary to secondary
LEAKAGE through any one SG of 150 gallons per day. These criteria
assure that accident doses will stay within regulatory and licensing
basis limits.
Therefore, the proposed change does not affect the probability
or consequences of any ANO-1 [Arkansas Nuclear One, Unit 1] analyzed
accidents.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed Steam Generator
Program will not introduce any adverse changes to the plant design
basis or postulated accidents resulting from potential tube
degradation. The proposed change does not affect the design of the
SGs, their method of operation, or primary or secondary coolant
chemistry controls. The proposed change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the Steam Generator Program to manage SG
tube inspection, assessment, repair, and plugging. The requirements
established by the Steam Generator Program are consistent with those
in the applicable design codes and standards and are an improvement
over the requirements in the current technical specifications.
Therefore, the margin of safety is not changed by the proposed
change to the ANO-1 TSs.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Michael K. Webb, Acting.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: September 30, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 4.2.1, Fuel Assemblies, to permit
the use of M5 advanced alloy for fuel rod cladding and fuel assembly
structural components. Also, the proposed amendment would modify TS
2.1.1.2, Reactor Core Safety Limits, to allow the use of the high
thermal power (BHTP) correlation for departure from nucleate boiling
(DNB) calculations of reload cores containing the Mark-B-HTP fuel
design.
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 NRC approved topical reports BAW-10227P-A, Evaluation of
Advanced Cladding and Structural Material (M5) in PWR [Pressurized
Water Reactor] Reactor Fuel, and BAW-10179P-A, Safety Criteria and
Methodology for Acceptable Cycle Reload Analyses, provide the
licensing basis for the Framatome ANP (FRA-ANP) advanced cladding
and structural material, designated M5. The M5 material was shown in
these documents to have equivalent or superior properties to the
currently used Zircaloy-4 material. The cladding itself is not an
accident initiator and does not affect accident probability. The M5
cladding has been shown to meet all 10 CFR 50.46 design criteria
and, therefore, will not increase the consequences of an accident.
The proposed safety limit value ensures that fuel integrity will
be maintained during normal operations and anticipated operational
occurrences (AOOs), and that the design requirements will continue
to be met. The core operating limits will be developed in accordance
with the new methodology. The proposed safety limit value does not
affect the performance of any equipment used to mitigate the
consequences of an analyzed accident. There is no impact on the
source term or pathways assumed in accidents previously evaluated.
No analysis assumptions are violated and there are no adverse
effects on the factors that contribute to offsite or onsite dose as
the result of an accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Use of M5 clad fuel will not result in changes in the operation
or configuration of the facility. Topical report BAW-10227P-A
demonstrated that the material properties of the M5 alloy are
similar or better than those of Zircaloy-4. Therefore, M5 fuel rod
cladding and fuel assembly structural components will perform
similarly to those fabricated from Zircaloy-4, thus precluding the
possibility of the fuel becoming an accident initiator and causing a
new or different type of accident.
In addition, there will be no change in the level of controls or
methodology used for processing radioactive effluents or handling
solid radioactive waste. Since the material properties of M5 alloy
are similar or better than those of Zircaloy-4, there will be no
significant changes in the types of any effluents that may be
released off-site. There will not be a significant increase in
occupational or public radiation exposure.
The proposed safety limit value does not change the methods
governing normal plant operation, nor are the methods utilized to
respond to plant transients altered. The BHTP correlation is not an
accident / event initiator. No new initiating events or transients
result from the use of the BHTP correlation or the related safety
limit changes.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will not involve a significant reduction in
the margin of safety because it has been demonstrated that the
material properties of the M5 alloy are not
[[Page 64989]]
significantly different from those of Zircaloy-4. M5 alloy is
expected to perform similarly or better than Zircaloy-4 for all
normal operating and accident scenarios, including both loss of
coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios,
where the slight difference in M5 material properties relative to
Zircaloy-4 could have some impact on the overall accident scenario,
plant-specific LOCA analyses will be performed prior to the use of
fuel assemblies with fuel rods or fuel assembly components
containing M5. These LOCA analyses, required by the ANO-1 [Arkansas
Nuclear One, Unit 1] TSs, will demonstrate that all applicable
margins of safety will be maintained by the use of M5 alloy.
The proposed safety limit value has been established in
accordance with the methodology for the BHTP correlation, to ensure
that the applicable margin of safety is maintained (i.e., there is
at least 95% probability at a 95% confidence level that the hot fuel
rod in the core does not experience DNB). The other reactor core
safety limits will continue to be met by analyzing the reload for
the mixed core using NRC approved methods, and incorporation of
resultant operating limits into the Core Operating Limits Report
(COLR).
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Michael K. Webb, Acting.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: September 1, 2004.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) 5.6.1, ``Occupational Radiation
Exposure Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated September 1, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated?
The proposed change eliminates the TS reporting requirements to
provide a monthly operating report of shutdown experience and
operating statistics if the equivalent data is submitted using an
industry electronic database. It also eliminates the Technical
Specification reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated?
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety?
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: James W. Clifford.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota; Docket No. 50-331, Duane
Arnold Energy Center, Linn County, Iowa; Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin; Docket No. 50-255,
Palisades Plant, Van Buren County, Michigan; Docket Nos. 50-266 and 50-
301, Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin; Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of amendment request: October 5, 2004.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) requirements for the licensee to
submit annual occupational radiation exposure reports and monthly
operating reports for the above nuclear plants. For the Kewaunee and
Monticello plants, the licensee is also proposing to adopt a part of
Revision 4 to TSTF-258, ``Changes to Section 5.0, Administrative
Controls,'' regarding reporting challenges to, and failures, of certain
safety/relief valves.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated October 5, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated?
The proposed change eliminates the TS reporting requirements to
provide a monthly operating report of shutdown experience and
operating statistics if the equivalent data is submitted using an
industry electronic database. It also eliminates the Technical
Specification reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
[[Page 64990]]
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Accident Previously
Evaluated?
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety?
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Southern Nuclear Operating Company (SNC), Inc., et al., Docket Nos. 50-
424 and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1 and 2,
Burke County, Georgia
Date of amendment request: August 13, 2004.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 3.7.18, ``Fuel Assembly Storage in
the Fuel Storage Pool;'' TS 4.3.1.1, the criticality design features
for fuel storage for VEGP Unit 1; and TS 4.3.1.2, the criticality
design features for fuel storage for VEGP Unit 2. The proposed
amendment would supplant the previous spent fuel rack criticality
analysis with updated criticality calculations. Editorial revisions to
TS Bases B 3.7.17, ``Fuel Storage Pool Boron Concentration,'' and B
3.7.18, ``Fuel Assembly Storage in the Fuel Storage Pool,'' are
included. In addition, Page vi of the Table of Contents will be updated
to reflect the correct page number for Figure 5.5.6-1.
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 consequence of an accident previously evaluated?
SNC has chosen to reanalyze the criticality analyses for the
VEGP Unit 1 and Unit 2 spent fuel racks. Westinghouse performed the
revised analyses using methods that address the non-conservatisms
previously identified in the current analyses. The methodologies
used for the revised analysis have been previously approved for use
by the NRC.
The analyses revised the enrichment, burnup, and Integral Fuel
Burnable Absorber (IFBA) limits required to comply with the allowed
storage configurations. The storage configurations and interface
requirements in the current Technical Specifications were retained
in the revised analyses. The boron dilution evaluation that
supported the initial amendments to permit credit for the soluble
boron at VEGP continues to remain valid. The analyses demonstrated
that Keff remains below unity for the various storage configurations
considered with zero soluble boron and that Keff remains less than
or equal to 0.95 for the entire pool with credit for soluble boron
under non-accident and accident conditions with a 95% probability at
a 95% confidence level (95/95).
Core design procedures ensure that new fuel can be stored in one
or more of the allowed storage configurations. Administrative
controls during fuel fabrication ensure that the fuel is fabricated
accordingly to ensure proper loading of the fuel in the fuel
assemblies. Administrative controls used to load fuel assemblies
into the spent fuel pool ensure that fuel assemblies are stored in
compliance with the allowed storage configurations. Fuel handling is
performed under many administrative controls and physical
limitations. These controls provide reasonable assurance that a
criticality accident, fuel fabrication error, or fuel handling
accident will not occur.
The change to the page number of Figure 5.5.6-1 on Page vi of
the Table of Contents is administrative in nature.
Therefore, based on the conclusions of the above analysis, the
proposed change does not involve 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 previously evaluated?
The types of accidents previously evaluated include fuel
fabrication errors, criticality accidents, and fuel handling
accidents. The analyses revised the enrichment, burnup, and Integral
Fuel Burnable Absorber (IFBA) limits required to comply with the
allowed storage configurations. No new or other kind of accident can
be postulated as a result of the revised analyses.
The change to the page number of Figure 5.5.6-1 on Page vi of
the Table of Contents is administrative in nature.
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 decrease in
the margin of safety?
The analyses revised the enrichment, burnup, and Integral Fuel
Burnable Absorber (IFBA) limits required to comply with the allowed
storage configurations. The boron dilution evaluation that supported
the initial amendments to permit credit for soluble boron at VEGP
was shown to remain valid. The analyses demonstrated that Keff
remains below unity for the various storage configurations
considered with zero soluble boron and that Keff remains less than
or equal to 0.95 for the entire pool with credit for soluble boron
under non-accident and accident conditions with a 95% probability at
a 95% confidence level (95/95).
The change to the page number of Figure 5.5.6-1 on Page vi of
the Table of Contents is administrative in nature.
Therefore, the proposed change does not involve a significant
decrease in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Mary Jane Ross-Lee, Acting.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: July 8, 2004 (TS-427).
Description of amendment request: The proposed amendment removes
the requirement to maintain an automatic transfer capability for the
power supply to the Low Pressure Coolant Injection (LPCI) inboard
injection and recirculation pump discharge valves. In addition, the
licensee has requested to delete the references to Reactor Motor
Operator Valve Boards D and E from Limiting Condition for Operation
3.8.7, and the Actions in 3.8.7 have been requested to be revised and/
or renumbered, as appropriate.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed Technical Specification change involve a
significant increase in the probability or consequences of an
accident previously evaluated?
Response: No.
Neither Reactor Motor Operated Valve (RMOV) Boards D and E, the
equipment they power, nor the automatic power transfer
[[Page 64991]]
feature provided for these boards are precursors to any accident
previous [sic] evaluated in the Updated Final Safety Analysis Report
(UFSAR). Therefore, the probability of an evaluated accident is not
increased by modifying this equipment.
The proposed deletion of the requirement to maintain an
automatic transfer capability for the power supply to the LPCI
inboard injection and recirculation pump discharge valves does not
change the number of Emergency Core Cooling System (ECCS) subsystems
credited in the BFN licensing basis. Therefore, the proposed TS
changes will not significantly increase the consequences of an
accident previously evaluated.
2. Does the proposed Technical Specification change create the
possibility of a new or different kind of accident from any accident
previously evaluated?
Response: No.
The proposed deletion of the requirement to maintain an
automatic transfer capability for the power supply to the LPCI
inboard injection and recirculation pump discharge valves does not
introduce new equipment, which could create a new or different kind
of accident. No new external threats, release pathways, or equipment
failure modes are created. Therefore, the proposed deletion of the
requirement to maintain an automatic transfer capability for the
power supply to the LPCI inboard injection and recirculation pump
discharge valves will not create a possibility for an accident of a
new or different type than those previously evaluated.
3. Does the proposed Technical Specification change involve a
significant reduction in a margin of safety?
Response: No.
The proposed deletion of the requirement to maintain an
automatic transfer capability for the power supply to the LPCI
inboard injection and recirculation pump discharge valves does not
change the number of ECCS subsystems credited in the BFN licensing
basis. The requirements of 10 CFR 50.46 and Appendix K continue to
be met. Therefore, the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant (BFN), Unit 1, Limestone County, Alabama
Date of amendment request: August 2, 2004 (TS-435).
Description of amendment request: Modify the COMPLETION TIME for
Technical Specification Limiting Condition for Operation (LCO) 3.6.3.1,
Containment Atmosphere Dilution (CAD) System. The proposed change would
extend the current completion time of 7 days with two CAD subsystems
inoperable from existing requirement to shut down the reactor within 13
hours in accordance with LCO 3.0.3, when both CAD subsystems are
inoperable.
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 safety-related function of the CAD system is to mitigate the
effects of a loss-of-coolant-accident (LOCA) by limiting the
volumetric concentration of oxygen in the primary containment
atmosphere. The CAD System is not an event initiator, therefore, the
probability of the occurrence of an accident is not affected by this
proposed Technical Specification change. Emergency procedures
preferentially use the normal containment inerting system to provide
post accident vent and purge capability, with the CAD system only
serving in a backup role to this system. Hence, in the event of the
inoperability of both CAD subsystems, the proposed TS require the
normal containment inerting system to be verified available as an
alternate oxygen control means. Therefore, the proposed TS change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not introduce new equipment, which
could create a new or different kind of accident. This proposed
change does not result in any changes to the CAD equipment design or
capabilities or to the operation of the plant. No new external
threats, release pathways, or equipment failure modes are created.
Therefore, the implementation of the proposed change will not create
a possibility for an accident of a new or different type than those
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
As stated in GL [Generic Letter] 84-09, a Mark I type boiling
water reactor (BWR) plant does not rely upon purge/repressurization
systems such as CAD as its primary means of hydrogen control when
the unit is operated in accordance with certain technical criteria.
The BFN units are operated in accordance with these criteria. The
BFN Unit 1 containment is inerted with nitrogen during normal
operation, nitrogen from the containment inerting system with a
backup from the CAD system is used for pneumatically operated
components inside containment, and there are no potential sources of
oxygen generation inside containment other than the radiolytic
decomposition of water. The system preferred by the Emergency
Operating Instructions (EOIs) for oxygen control post-accident is
the normal primary containment inerting system. Because the
probability of an accident involving hydrogen and oxygen production
is small, CAD is not the primary system used to mitigate the
creation of combustible containment atmosphere mixtures, and because
the requested LCO where both CAD subsystems is inoperable is not
long, no significant reduction in the margin of safety is associated
with this 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 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 Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328,
Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: August 18, 2004.
Description of amendment request: The proposed amendment would
update the reactor coolant system (RCS) and emergency core cooling
system (ECCS) technical specifications (TSs). These changes include
deleting TS 3/4.4.2, ``Safety Valves--Shutdown'' in its entirety,
revising the action requirements for TS 3/4.4.3, ``Safety and Relief
Valves--Operating,'' and deleting surveillance requirement 4.4.3.2.1.a
for TS 3.4.3.2, ``Relief Valves--Operating.'' The proposed changes are
consistent with the Sequoyah (SQN) safety analyses provided in the SQN
Updated Final Safety Analyses Report and the improved standard
technical specifications (NUREG-1431, Revision 3).
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 64992]]
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. TVA's proposed TS revisions do not involve a significant
increase in the probability of any accidents previously evaluated.
TVA's proposed TS revisions provide improvements to the RCS and ECCS
requirements to include appropriate reference to SQN's PTLR
[PressureTemperature Limits Report] requirements. The proposed
revision is a TS improvement that remains consistent with the
improved standard TS requirements for Pressurized Water Reactors
(PWRs) (NUREG-1431, Revision 3). TVA's proposed revision to delete
SQN TS 3/4.4.2.1, ``Reactor Coolant Safety Valves--Shutdown,'' does
not involve a significant increase in the probability of any
accident previously evaluated. Pressurizer code safety valve
requirements are not applicable for plant shutdown conditions (i.e.,
modes 4 and 5) because the valves do not perform a safety function
in these modes. The pressurizer code safety valves are not used as
inputs to initiating events or accidents previously evaluated.
Protection of the RCS against an overpressure condition in modes 4
and 5 is provided by the LTOP [low temperature overpressure
protection] system which is governed by SQN TS 3.4.12. The setpoint
for the pressurizer code safety valves is sufficiently high such
that the safety valves do not afford protection to the RCS during
low temperature operation. Accordingly, there is no impact on the
consequences previously evaluated for the proposed change.
The proposed revisions are not the result of changes to plant
equipment, test methods or operating practices. The proposed changes
do not contribute to the generation or assumptions for postulated
accidents. The proposed changes do not affect the design basis
accidents or their assumptions. The revisions to SQN TSs continue to
support SQN's required safety functions.
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 revisions are not the result of changes to
plant equipment or plant design. The proposed revisions adopt
standard TS requirements that are consistent with SQN's safety
analysis and design and provide improvements over the existing
requirements. The safety functions of the RCS and ECCS remain
unchanged and do not affect any assumptions in SQN's accident
analyses.
TVA's proposed change to delete the mode 4 and mode 5 TS
requirements for pressurizer safety valves is consistent with the
Policy Criterion of 10 CFR 50.36. The pressurizer code safety valves
are not assumed to function for any safety analysis in modes 4 and 5
and consequently, the proposed changes do not create the possibility
of a new or different kind of accident.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed TS change does not involve a significant
reduction in a margin of safety. TVA's proposed revisions will not
result in changes to system design features or plant features that
could be precursors to accidents or potential degradation of
accident mitigation systems. The proposed changes to the RCS and
ECCS requirements remain consistent with the current TS requirements
for equipment operability. Therefore, the proposed changes do not
involve a significant reduction in the margin of safety.
TVA's proposed change that removes the requirement for a
pressurizer safety valve in modes 4 and 5 does not affect any margin
of safety because the lift setting of the pressurizer code safety
valves (2485 pounds per square inch gauge [psig] 3
percent) is well above the limit needed to protect the RCS during
low temperature operation and would not provide any safety function
for overpressure protection in the lower modes. The TS requirements
associated with low temperature operation are governed by SQN TS 3/
4.4.12, LTOP system. The LTOP system provides the necessary
overpressure protection for SQN's RCS in modes 4 and 5.
Accordingly, TVA's proposed deletion of operability requirements
for SQN's pressurizer code safety valves for modes 4 and 5 will not
affect the margin of safety.
The United States Nuclear Regulatory Commission (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 Section Chief: Michael L. Marshall, Jr.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: October 12, 2004.
Brief description of amendment request: The proposed amendment
would approve an engineering evaluation performed in accordance with
Pilgrim Nuclear Power Station Technical Specification (TS) 3.6.D.3 to
justify continued power operation with a safety relief valve discharge
pipe temperature exceeding 212 degrees Fahrenheit for greater than 24
hours as required by TS 3.6.D.4.
Date of publication of individual notice in Federal Register:
October 20, 2004 (69 FR 61695).
Expiration date of individual notice: December 19, 2004.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these
[[Page 64993]]
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]. (Note: Public access to ADAMS has
been temporarily suspended so that security reviews of publicly
available documents may be performed and potentially sensitive
information removed. Please check the NRC Web site for updates on the
resumption of ADAMS access.)
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: July 30, 2004.
Brief description of amendment: The proposed amendment would (1)
add License Condition 2.C.(22) requiring an integrated tracer gas test
of the control room envelope using methods described in American
Society for Testing and Materials E741-00, ``Standard Test Method for
Determining Air Change in a Single Zone by Means of a Tracer Gas
Dilution,'' and (2) delete Surveillance Requirement 3.7.3.6, which
requires verification that unfiltered inleakage from control room
emergency filtration system duct work outside the control room envelope
is within limits.
Date of issuance: October 25, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 162.
Facility Operating License No. NPF-43: Amendment adds a license
condition and revises the Technical Specifications.
Date of initial notice in Federal Register: August 13, 2004 (69 FR
50217)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 25, 2004.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 21, 2003, as supplemented by
letters dated February 10, 2004, and August 24, 2004.
Brief description of amendment: Modifies the Technical
Specifications (TSs) to delete TS 3.6.4.4, ``Shield Building Annulus
Mixing System'' and a reference to TS 3.6.4.4 within TS 3.10.1,
``Inservice Leak and Hydrostatic Testing Operation,'' and revise TS
Surveillance Requirement 3.6.1.3.10, main steam isolation valve leakage
limits.
Date of issuance: October 15, 2004.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 143.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 2004 (69 FR
29764). The supplement dated August 24, 2004, 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 October 15, 2004.
No significant hazards consideration comments received: No.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: February 16, 2004, as supplemented by
letters dated June 8 and August 26, 2004.
Brief description of amendment: Modifies the Technical
Specifications (TSs) to change Surveillance Requirement 3.6.5.1.3 of TS
3.6.5.1, ``Drywell,'' to allow a one-time extension of the test
interval for the next drywell bypass leakage rate test from 10 years to
15 years.
Date of issuance: October 15, 2004.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 144.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 2004 (69 FR
29765). The supplements dated June 8 and August 26, 2004, 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 October 15, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of application for amendment: January 29, 2004, as
supplemented on April 12, June 16, June 30, July 16, August 3, August
12, and September 24, 2004.
Brief description of amendment: The amendment revises the operating
license and Technical Specifications to authorize an increase in the
maximum steady-state reactor core power level from 3114.4 megawatt
thermal (MWt) to 3216 MWt. This represents a nominal increase of 3.26%
rated thermal power.
Date of issuance: October 27, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 241.
Facility Operating License No. DPR-26: Amendment revised the
Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: March 2, 2004 (69 FR
9859). The April 12, June 16, July 16, August 3, August 12, and
September 24, 2004, supplements provided additional information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the NRC 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 October 27, 2004.
No significant hazards consideration comments received: No.
[[Page 64994]]
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: January 15, 2004, and
supplemented on July 19, 2004.
Brief description of amendments: The amendments provide for an
alternative means of testing the main steam Electromatic relief valves
and the dual function Target Rock safety/relief valves.
Date of issuance: October 19, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 211/203, 222/217.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 16, 2004.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 19, 2004.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: August 19, 2003.
Brief description of amendments: The amendments modify Technical
Specification (TS) 5.5.13, ``Primary Containment Leakage Rate Testing
Program,'' to allow an exception to the testing guidance contained in
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test
Program.'' Specifically, the TS change will allow potential valve
atmospheric leakage paths (e.g., valve stem packing) that are not
exposed to test pressure during reverse-direction Type B or C tests
(local leakage rate tests) to instead be tested during regularly
scheduled Type A tests (integrated leakage rate tests).
Date of issuance: October 14, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 168/154.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 23, 2003 (68
FR 74266).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 14, 2004.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334,
Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County,
Pennsylvania
Date of application for amendment: June 28, 2004, as supplemented
September 3, 2004.
Brief description of amendment: The amendment revised the BVPS-1
Technical Specifications (TSs) surveillance requirements (SRs)
4.4.5.4.a.6, 4.4.5.4.a.8, and 4.4.5.5.d.1 and added SRs 4.4.5.4.a.11
and 4.4.5.5.e for Cycle 17 operation only. The change revised the
definition of steam generator tube inspection scope in SR 4.4.5.4.a.8
to exclude the portion of the tube within the tubesheet below the W*
distance, tube to tubesheet weld and tube-end extension by crediting
the Westinghouse W* methodology as described in Topical Report WCAP-
14797, Revision 2.
Date of issuance: October 15, 2004.
Effective date: This license amendment is effective as of its date
of issuance and shall be implemented within 60 days.
Amendment No.: 262.
Facility Operating License No. DPR-66: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 2004 (69 FR
46584). The supplement dated September 3, 2004, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the Nuclear
Regulatory Commission staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 15, 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: December 9, 2003, as
supplemented September 16, 2004.
Brief description of amendment: The amendment allows a one-time
increase in the completion time for restoring an inoperable emergency
feedwater (EFW) system train to operable status to allow the
realignment of the diesel-driven EFW pump during power operations.
Date of issuance: October 21, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 214.
Facility Operating License No. DPR-72: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 2004 (69 FR
16620). The September 16, 2004, supplemental letter provided additional
information that clarified the application, but did not expand the
scope of the application as originally noticed and did not change the
U.S. Nuclear Regulatory Commission staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 21, 2004.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: February 3, 2004.
Description of amendment request: This amendment revised a footnote
to clarify a surveillance requirement and associated bases for
emergency diesel generator testing.
Date of issuance: October 25, 2004.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 98.
Facility Operating License No. NPF-86: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 16, 2004 (69 FR
12371).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 25, 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: July 15, 2004, as supplement by letters
dated September 28 and October 14, 2004.
Brief description of amendment: The amendment revises the Technical
Specification (TS) Section 3.8.1, AC Sources--Operating, Condition B,
to provide a one-time extension of the allowed outage time for one
Diesel Generator (DG) inoperable from 7 days
[[Page 64995]]
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.
Date of issuance: October 15, 2004.
Effective date: As of the date of issuance and shall be implemented
on or before October 22, 2004.
Amendment No.: 207.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 2004 (69 FR
46586). The supplements dated September 28 and October 14, 2004,
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 October 15, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: December 23, 2003, as
supplemented by letter dated August 16, 2004.
Brief description of amendments: The amendments modify technical
specification (TS) requirements to adopt the provisions of Industry/TS
Task Force (TSTF) change TSTF-359, ``Increased Flexibility in Mode
Restraints.''
Date of issuance: October 20, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 167, 157.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 16, 2004 (69
FR 55844) The supplement dated August 16, 2004, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 20, 2004.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: December 1, 2003, as supplemented by
letter dated July 2, 2004.
Brief description of amendment: The amendment changes the Fort
Calhoun Station, Unit No. 1 Technical Specifications (TS) 2.7,
``Electrical Systems, TS Table 3-5, ``Minimum Frequencies for Equipment
Tests,'' and TS 5.0, ``Administrative Controls,'' to modify the
requirements for the diesel generator (DG) fuel oil for consistency
with the Improved Standard Technical Specifications. The amendment also
adds requirements for the DG lubricating oil and DG starting air.
Date of issuance: October 21, 2004.
Effective date: October 21, 2004, and shall be implemented within
120 days from the date of its issuance.
Amendment No.: 229.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7526). The additional information provided in the supplemental
letter dated July 2, 2004, did not expand the scope of the application
as noticed and did not change the NRC staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated October 21, 2004.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: October 23, 2003, as
supplemented by letters dated June 24, 2004 and August 26, 2004.
Brief description of amendment: The amendment revised Technical
Specifications to delete the Surveillance Requirement associated with
the emergency diesel generator lockout features.
Date of issuance: October 22, 2004.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 155.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68671). The June 24, 2004, and August 26, 2004, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the
application beyond the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 22, 2004.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: December 12, 2003.
Brief description of amendment: The amendment revised the operating
conditions for which Technical Specification (TS) 3/4.3.7.1,
``Radiation Monitoring Instrumentation,'' requires the control room
ventilation radiation monitor to be operable. Additionally, the
amendment revised the operating conditions for which TS 3/4.7.2,
``Control Room Emergency Filtration System,'' is applicable.
Date of issuance: October 28, 2004.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 156.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7527).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 28, 2004.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: December 24, 2003, as
supplemented by letter dated June 8, 2004.
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to allow the use of GE14 fuel in reload cycle 13.
Specifically, the change modified the TSs to reflect the use of General
Electric (GE) core reload analysis methodology. The change revised the
limiting conditions for operation for the recirculation loops to modify
and add action statements to provide further thermal limit control
during single-loop operation to be consistent with the GE methodology
specified in the core operating limits report. The change also
[[Page 64996]]
modified the TS definitions and TS requirements for average planar
linear heat generation rate. Additionally, TS Section 6.9.1.9 is
revised to correct an error from a previous amendment that
inadvertently removed a reference.
Date of issuance: October 20, 2004.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 154.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7528). The June 8, 2004 letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination or expand the application beyond the scope
of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 20, 2004.
No significant hazards consideration comments received: No.
Dated in Rockville, Maryland, this 1st day of November 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-24804 Filed 11-8-04; 8:45 am]
BILLING CODE 7590-01-P