[Federal Register Volume 64, Number 240 (Wednesday, December 15, 1999)]
[Notices]
[Pages 70077-70098]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-32311]
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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 Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision 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 November 20, 1999, through December 3, 1999.
The last biweekly notice was published on December 1, 1999 (64 FR
67330).
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.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing 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
Review and Directives Branch, Division of Freedom of Information and
Publications 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
[[Page 70078]]
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By January 14, 2000, 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 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC, and electronically from
the ADAMS Public Library component on the NRC Web site, http://
www.nrc.gov (the Electronic Reading Room). If a request for a hearing
or petition for leave to intervene is filed by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room)
Baltimore Gas and Electric Company, Docket No. 50-317, Calvert Cliffs
Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of amendment request: November 18, 1999.
Description of amendment request: The proposed amendment revises
the Unit 1 Heatup Curve (Technical Specification Figure 3.4.3-1), Unit
1 Cooldown Curve (Technical Specification Figure 3.4.3-2), and Unit 1
Maximum Power-Operated Relief Valve (PORV) Opening Pressure vs
Temperature Curve (Technical Specification Figure 3.4.12-1) to change
fluence level from 2.61 x 10\19\ n/cm \2\ to 4.49 x 1019 n/cm \2\
(E>1MeV). This change reflects the new actual fluence level for which
these curves are valid, and is necessary to extend the
[[Page 70079]]
applicability of the curves for Unit 1 operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
In accordance with 10 CFR Part 50, Appendix G, the Calvert
Cliffs pressure/temperature (P-T) limits for material fracture
toughness requirements of the reactor coolant pressure boundary
materials were developed using the methods of linear elastic
fracture mechanics and the guidance found in the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section
III, Appendix G. The Calvert Cliffs (P-T) limits are based on
fluence level. The fluence level corresponds to the pressurized
thermal shock (PTS) screening criteria defined in 10 CFR 50.61 for
the critical elements. Methods described in the Nuclear Regulatory
Commission Regulatory Guide 1.99, Revision 2, are used to predict
the embrittlement effect of neutron irradiation on reactor vessel
materials. Regulatory Guide 1.99 defines embrittlement effect in
terms of adjusted reference temperatures (ART), which depends on the
material property of the PTS critical element.
The proposed higher fluence level for the Technical
Specification P-T limits was made possible by the identification of
a new 10 CFR 50.61 critical element for fracture toughness
requirements for protection against PTS events. The material
properties of the new critical element resulted in an increase in
fluence level from 2.61 x 10 \19\ n/cm \2\ to 4.49 x 1019 n/cm \2\
for the ART valves calculated using the material properties of the
old PTS critical element. the P-T limits analysis remain well within
the conservative acceptance limits of the ASME Boiler and Pressure
Vessel Code, Section III, Appendix G. Hence, with the new higher
fluence level, the 10 CFR Part 50, Appendix G, requirement for
adequate margin to brittle failure during normal operation,
anticipated operational occurrences, and system hydrostatic tests,
for the reactor coolant pressure boundary materials, is maintained.
Therefore the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accidents previously evaluated.
The implementation of the proposed revision has no significant
effect on either the configuration of the plant, or the manner in
which it is operated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
As discussed above, the P-T limits analysis remain well within
the conservative acceptance limits of the ASME Boiler and Pressure
Vessel Code, Section III, Appendix G. Hence, with the new higher
fluence level, the 10 CFR Part 50, Appendix G, requirement for
adequate margin to brittle failure during normal operation,
anticipated operational occurrences, and system hydrostatic tests,
for the reactor coolant pressure boundary materials, is maintained.
Therefore, this proposed modification does not significantly
reduce 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: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Acting Section Chief: Victor Nerses.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: November 19, 1999.
Description of amendments request: The amendments request approval
of changes in the Updated Final Safety Analysis Report (UFSAR) that
constitute an unreviewed safety question (USQ) as described in 10 CFR
50.59. Specifically, these changes would be an increase in the
probability of occurrence of malfunction. Additionally, these changes
were not previously evaluated in the UFSAR.
Regulations require that structures, systems, and components
important to safety be appropriately protected against the effects of
effects of missiles that might result from equipment failures. Failures
that could occur in the large turbines of the main turbine-generator
sets have the potential for producing large high-energy missiles
(hereinafter called ``turbine missiles''). Both of Baltimore Gas and
Electric Company's (BGE) turbine generator suppliers studied the
failure of the rotating elements of their turbine-generators. The UFSAR
only addresses a turbine missile hitting the Containment Building,
Control Room, Switchgear Room, and Waste Processing Area. As a result
of revising the Unit 1 and Unit 2 turbine missile analysis, BGE
determined that the discussion of turbine missiles in Section 5.3.1 of
the UFSAR was incomplete. Specifically, it did not discuss the
probability of a missile from the Unit 1 turbine-generator striking: 1)
the refueling water tanks; 2) the No. 11 Fuel Oil Storage Tank; or 3)
plant equipment through various roof slabs or through non-missile-proof
openings in the missile-proof walls. When these additional targets are
included, the total target area is increased. If the target area
increases, the probability of a turbine missile causing equipment
damage increases. It is this increase in probability that leads to a
USQ for a turbine missile from Unit 1. Note that by using methodologies
previously approved by NRC, the revised analysis concludes there is no
USQ for turbine missiles from the Unit 2 turbine-generator.
The UFSAR change is considered a USQ for Units 1 and 2 because the
results of the revised Unit 1 turbine missile analysis for the
following unprotected rooms or components show an increase in
probability of occurrence of malfunction not previously evaluated in
the UFSAR:
the Refueling Water Tanks;
the No. 11 Fuel Oil Storage Tank (non-missile-proof);
the saltwater pumps through roof hatches in the Intake Structure
roof;
the roof slabs over the refueling Water Tank Pump Room, the Control
Room Heating, Ventilation, and Air Conditioning (HVAC) Equipment Room,
the Spent Fuel Pool Area Ventilation Equipment room, and a portion of
118 level roof over the fuel cask handling area;
the Control Room HVAC Room through its non-missile-proof door; and
the Unit 1 Auxiliary Building 45 Switchgear Room through
the its non-missile-proof doors.
The probability of a missile from the Unit 1 turbine-generator
striking them is a negligible increase in the probability of occurrence
of malfunction of equipment associated with Unit 1 and 2. Upon approval
of this request, the UFSAR will be revised to reflect the proposed
turbine missile description. There is no USQ associated with the Unit 2
turbine-generator.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Regulations require that structures, systems, and components
important to safety be appropriately protected against the effects
of missiles that might result from equipment failures. Further that
could occur in the large turbines of the main turbine-generator sets
have the potential for producing large high-
[[Page 70080]]
energy missiles (hereinafter called turbine missiles). Both of our
turbine-generator suppliers studied the failure of the rotating
elements of their turbine-generators. The UFSAR only addresses
turbine missile hitting the Containment Building, Control Room,
Switchgear Room, and Waste Processing Area. As result of revising
the Unit 1 and Unit 2 turbine missile analysis, we determined that
the discussion of turbine missiles of the UFSAR was incomplete. From
the revised analysis, we determined Unit 1 and 2 USQs exist for the
following unprotected rooms or components (i.e., there is an
increase in probability of occurrence of malfunction not previously
evaluated in the UFSAR):
the Refueling Water Tanks;
the No. 11 Fuel Oil Storage Tank;
the Saltwater Pumps through roof hatches in the Intake Structure
Roof;
the roof slabs over the Refueling Water Tank Pump Room, the
Control Room Heating, Ventilation, and Air Conditioning (HVAC)
Equipment Room, Spent Fuel Pool Area Ventilation Equipment Room, and
a portion of 118' level roof over the cask handling area;
the Control Room HVAC Room through its non-missile-proof door;
and,
the Unit 1 Auxiliary Building 45' Switchgear Room through its
non-missile-proof doors.
The probability of a missile from the Unit 1 turbine-generator
striking them is a negligible, but greater than zero, increase in
the probability of occurrence of malfunction of equipment associated
with Units 1 and 2.
For Unit 1 High Trajectory Missiles (HTM), the guidance of NUREG
0800, Standard Review Plan, is used as one acceptable method for
evaluating the risk. Use of this method is not a commitment to the
Standard Review Plan and does not incorporate the Standard Review
Plan into our licensing basis. The revised analysis shows that the
total target area considered vulnerable to an HTM is less than the
Standard Review Plan limit of 10,000 ft2 for each unit.
Therefore, the risk form an HTM is insignificant. Note that all of
the Units 1, 2, and Common structures listed above are equally
vulnerable to a Unit 1 HTM. Therefore, any risk increase to the
plant structures constitutes a USQ for Units 1 and 2.
For Unit 1 Low-Trajectory Missiles (LTMs), protection for the
Auxiliary Building is provided by a 3' thick, concrete, missile-
proof wall between the Turbine Building and the Auxiliary building
(the K-line wall). This wall is 3' thick below the 69' elevation and
2' thick above the 69' for areas protecting safety-related
equipment. The revised analysis evaluates the protection of Unit 1
equipment from a Unit 1 LTM. The 69' Control Room HVAC Equipment
Room and Unit 1 Auxiliary Building 45' Switchgear Room are protected
by the missile-proof walls except for the openings at the non-
missile-proof doors. A turbine missile that hits one of these doors
is assumed to go through them, strike safety-related equipment in
the room, and cause it to fail. Recall that the Control Room HVAC
equipment is shared by both units. Therefore, any increase in risk
of failure of equipment in this room affects both Units 1 and 2.
The risk associated with a turbine missile to either of these
doors is calculated using guidance in Regulatory Guide 1.115,
Revision 1, ``Protection Against Low-Trajectory Turbine Missiles.''
This guidance states that the turbine missile hazard should be less
than 107. The missile hazard rate in the revised risk
analysis shows that the risk from LTMs from the Unit 1 General
Electric turbine-generator to the 69' Control Room HVAC Equipment
Room and Unit 1 Auxiliary Building 45' Switchgear room through these
non-missile-proof doors is less than 107.
Based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed change makes no physical changes to the plant.
Specifically, the proposed change does not add new or modify
existing plant equipment such that it could become an accident
initiator different from its current role as an accident initiator.
The only change made by this activity is the revision of the UFSAR
to include the revised turbine missile analysis. The UFSAR chapter 1
drawings correctly depict the location of plant structures and
components, including the thickness of and the openings in the
missile-proof wall between the Turbine Building and the Auxiliary
building (the K-Line Wall). Therefore, the possibility of a new or
different type of accident is not created by the proposed change.
3. Would not involve a significant reduction in a margin of
safety.
The regulations require an evaluation of turbine missiles to
ensure that structures, systems, and components important to safety
be appropriately protected from them. Revised turbine missile
analysis have been performed consistent with appropriate regulatory
guidance (Regulatory Guide 1.115 and the Standard Review Plan). The
results of the revised analysis meet the acceptance criteria of the
guidance. 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
amendments request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Acting Section Chief: Victor Nerses.
Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick
Steam Electric Plant, Unit 1, Brunswick County, North Carolina
Date of amendment request: November 17, 1999.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 2.1.1.2, ``Reactor Core Safety
Limits.'' The minimum critical power ratios (MCPR) for single and two
recirculation loop operation would be increased. In addition, the
reference in TS 5.6.5, ``Core Operating Limits Report,'' Item b.5,
would be removed.
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 license amendments do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The proposed license amendment will establish MCPR Safety Limit
values of 1.10 for two recirculation loop operation and 1.11 for single
recirculation loop operation. Additionally, the proposed license
amendment replaces an expiring cycle-specific reference in the list of
analytical methods approved for determining core operating limits in
Specification 5.6.5.b with a reference to a GE [General Electric]
topical report which has been accepted by the NRC.
The methods for calculating the MCPR Safety Limit values have been
previously approved by the NRC and are described in GE's reload
licensing methodology topical report NEDE-24011-P-A. Use of these
methods ensures that the integrity of the fuel will be maintained
during normal operation and that the resulting MCPR Safety Limit values
satisfy the fuel design safety criteria that less than 0.1 percent of
the fuel rods experience boiling transition if the safety limits are
not violated. The change does not require any physical plant
modifications, physically affect any plant components, or allow the
plant to be operated any closer to fuel design limits. Therefore, the
proposed change to the MCPR Safety Limit values and to the list in
Specification 5.6.5.b of analytical methods approved for determining
core operating limits results no increase in the probability of a
previously evaluated accident.
The consequences of a previously evaluated accident are dependent
on the initial conditions assumed for the analysis, the behavior of the
fuel during the accident, the availability and successful functioning
of the equipment assumed to operate in response to the accident, and
the setpoints at which these actions are initiated.
[[Page 70081]]
The methods used for calculating the MCPR Safety Limits have been
approved by the NRC and are described in GE's reload licensing
methodology topical report NEDE-24011, ``General Electric Standard
Application for Reactor Fuel (GESTAR II).'' The proposed MCPR Safety
Limit values of 1.10 for two recirculation loop operation and 1.11 for
single recirculation loop operation will ensure that less than 0.1
percent of the fuel rods will experience boiling transition during any
plant operation if the limits are not violated. The proposed change to
the MCPR Safety Limit values does not affect the performance of any
equipment used to mitigate the consequences of a previously evaluated
accident. Also, the proposed change does not affect setpoints that
initiate protective or mitigative actions. No analysis assumptions are
violated and there are no adverse effects on the factors contributing
to offsite and onsite dose.
Based on the determination of the proposed MCPR Safety Limit values
using conservative NRC-approved methods and the operability of plant
systems designed to mitigate the consequences of accidents not being
changed, the proposed change to the MCPR Safety Limit values and to the
list in Specification 5.6.5.b of analytical methods approved for
determining core operating limits does not significantly increase the
consequences of a previously evaluated accident.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. This proposed license amendment does
not involve any physical alteration of plant systems and plant
equipment will not be operated in a different manner. As a result,
no new failure modes are being introduced. Therefore, the proposed
change to the MCPR Safety Limit values and to the list in
Specification 5.6.5.b of analytical methods approved for determining
core operating limits will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety.
The margin of safety is established through the design of the
plant structures, systems, and components; through the parameters
within which the plant is operated; through the establishment of
setpoints for actuation of equipment relied upon to respond to an
event; and through margins contained within the safety analyses.
The proposed change to the MCPR Safety Limit values and the list
in Specification 5.6.5.b of analytical methods approved for
determining core operating limits does not adversely impact the
performance of plant structures, systems, components, and setpoints
relied upon to respond to mitigate an accident. As previously
stated, the methods for calculating the MCPR Safety Limit values
have been previously approved by the NRC and are described in GE's
reload licensing methodology topical report NEDE-24011-P-A. Use of
these methods ensures that the resulting MCPR Safety Limit values
satisfy the fuel design safety criteria that less than 0.1 percent
of the fuel rods experience boiling transition if the safety limits
are not violated. As a result, the proposed changes do not
significantly impact any safety analysis assumptions or results.
Based on the assurance that the fuel design safety criteria will be
met, 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: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Richard P. Correia.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: November 19, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) for the Harris Nuclear Plant
(HNP) to incorporate American Society for Testing and Materials (ASTM)
D3803-1989, ``Standard Test Method for Nuclear-Grade Activated
Carbon,'' as the standard for testing nuclear-grade activated charcoal.
Specifically, TS 4.7.6 will be revised for the Control Room Emergency
Filtration System, TS 4.7.7 will be revised for the Reactor Auxiliary
Building Emergency Exhaust System, and TS 4.9.12 will be revised for
the Fuel Handling Building Emergency Exhaust System. These changes are
being proposed in accordance with NRC Generic Letter (GL) 99-02,
``Laboratory Testing Of Nuclear-Grade Activated Charcoal,'' dated June
3, 1999.
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 license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This proposed change to revise the standard to which activated
charcoal samples are tested will ensure that testing is accurate and
repeatable. This will help ensure that the Engineered Safety Feature
(ESF) ventilation systems are capable of performing their safety
function. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes incorporate ASTM D3803-1989 as the testing
standard for nuclear-grade activated charcoal samples. This will
ensure that testing is accurate and repeatable. Plant structures,
systems, and components will not be operated in a different manner
as a result of these proposed changes and no physical modifications
to equipment are involved. Using the improved testing protocol does
not have the potential for creating the possibility of a new or
different type of accident from any previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes do not change the manner in which
structures, systems or components are operated. Revising the
standard to which activated charcoal samples are tested will ensure
that testing is accurate and repeatable. This will help ensure that
the ESF ventilation systems are capable of performing their safety
function. Therefore, the proposed changes do not involve a 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: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Richard P. Correia.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: November 12, 1999.
Description of amendment request: The proposed change revises the
pressure-temperature limits by revising
[[Page 70082]]
the heatup, cooldown and inservice test limitations for the Reactor
Pressure Vessel to a maximum of 32 Effective Full Power Years.
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 change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
The proposed changes do not modify the reactor coolant pressure
boundary, do not make changes in operating pressure, materials or
seismic loading. The proposed changes adjust the reference
temperature for the limiting beltline material to account for
radiation effects and provide the same level of protection as
previously evaluated. The proposed changes do not adversely affect
the integrity of the reactor coolant system (RCS) such that its
function in the control of radiological consequences is affected.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not create the possibility of a new or
different kind of accident previously evaluated for Quad Cities
Nuclear Power Station. No new modes of operation are introduced by
the proposed changes. The proposed changes will not create any
failure mode not bounded by previously evaluated accidents. Use of
the revised pressure-temperature (P-T) curves will continue to
provide the same level of protection as was previously reviewed and
approved.
Further, the proposed changes to the P-T curves do not affect
any activities or equipment, and are not assumed in any safety
analysis to initiate any accident sequence for Quad Cities Nuclear
Power Station. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes reflect an update of the P-T curves to
extend the Reactor Pressure Vessel (RPV) operating limit to 32
Effective Full Power Years (EFPY). The revised curves are based on
the latest American Society of Mechanical Engineers (ASME) guidance
and actual operational data for the units. This proposed changes are
acceptable because the ASME guidance maintains the relative margin
of safety commensurate with that which existed at the time that the
ASME Section IX Appendix G was approved in 1974. Therefore, the
proposed changes do 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: November 16, 1999.
Description of amendment request: The proposed change modifies the
surveillance requirements for Functional Unit 3 on Table 4.1.A-1 due to
replacement of the Reactor Pressure Vessel Steam Dome pressure switches
with analog trip units.
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:
Does the change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
During the upcoming refueling outages at Quad Cities Nuclear
Power Station, Unit 1 and Unit 2, a design change will be
implemented that upgrades the existing Reactor Vessel Steam Dome-
High instrumentation from a pressure switch to an analog trip unit
device. Analog trip units are proven technology that are more
reliable than existing equipment. Analog trip units are used in
various applications of Quad Cities Nuclear Power Station, including
the Reactor Protection System (RPS) low water level trip function.
The proposed change adds a CHANNEL CHECK and 31-day trip unit
calibration requirement for the Reactor Vessel Steam Dome Pressure--
High RPS trip function. This requirement is not applicable to the
existing instrumentation because the Barksdale pressure switches are
non-indicating and do not employ trip units.
Technical Specification (TS) requirements that govern
operability or routine testing of plant instruments are not assumed
to be initiators of any analyzed event because these instruments are
intended to prevent, detect, or mitigate accidents. Therefore, these
changes will not involve an increase in the probability of
occurrence of an accident previously evaluated. Additionally, these
changes will not increase the consequences of an accident previously
evaluated because the proposed change does not adversely impact
structures, systems, or components (SSCs). The planned instrument
upgrade is a more reliable design than existing equipment. The
proposed change establishes requirements that ensures components are
operable when necessary for the prevention or mitigation of
accidents or transients. Furthermore, there will be no change in the
types or significant increase in the amounts of any effluents
released offsite. For these reasons, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes support a planned instrumentation upgrade
by incorporating Surveillance Requirements required to ensure
operability. The change does not adversely impact the manner in
which the instrument will operate under normal and abnormal
operating conditions. Therefore, these changes provide an equivalent
level of safety and will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The changes in methods governing normal plant operation are
consistent with the current safety analysis assumptions. Therefore,
these changes will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed change supports a planned instrumentation upgrade.
The proposed change does not affect the probability of failure or
availability of the affected instrumentation. The addition of a
CHANNEL CHECK and 31-day trip unit calibration for RPS Functional
Unit 3 (Reactor Vessel Steam Dome Pressure--High) is a conservative
change that aligns the surveillance requirements for a planned
instrumentation upgrade with that of similar instrumentation.
Therefore, it is concluded that the proposed changes will not result
in a 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: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington
Date of amendment request: October 13, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.6.1, Table 3.3.6-1, ``Primary
Containment
[[Page 70083]]
Isolation Instrumentation.'' This amendment requests that Function 5 on
Table 3.3.6-1, ``RHR SDC System Isolation,'' be modified by removing
footnote (d). Footnote (d) states, ``Only the inboard trip system is
required in Modes 1, 2, and 3, as applicable, when the outboard valve
control is transferred to the alternate remote shutdown panel and the
outboard valve is closed.'' The outboard suction valve, RHR-V-8, is no
longer used as a high/low pressure interface in the residual heat
removal (RHR) system. Valve RHR-V-9, which is in series with valve RHR-
V-8, is now used as the high/low pressure interface valve. Valve RHR-V-
9 is operable in all modes of operation and therefore, footnote (d) is
no longer needed. The current footnote (e) will be relettered as
footnote (d) for consistency.
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 involves the probability and consequences of
accidents associated with the isolation of the RHR SDC [shutdown
cooling] mode of RHR operation. Isolation is provided if high
temperatures occur in RHR pump rooms or heat exchanger areas, if
reactor vessel water level is low, or if reactor vessel pressure is
high.
FSAR [Final Safety Analysis Report] Chapter 15, ``Accident
Analysis,'' describes two events associated with the RHR system
during SDC operation. FSAR Section 15.1.6, ``Inadvertent Residual
Heat Removal Shutdown Cooling Operation,'' describes the impact of
system operation during startup or cool-down when the reactor is
near critical. The proposed change removes the exemption for the
second trip system to isolate RHR SDC operation. There will be no
change in the probability or consequences of this accident as a
result of the proposed change.
The second accident is described in FSAR Section 15.2.9,
``Failure of Residual Heat Removal Shutdown Cooling.'' It postulates
the failure of the RHR system to function in SDC mode. The
evaluation assumes a failure of the SDC mode of operation but does
not disable the remaining modes of RHR operation. The alternate SDC
paths involve the use of the safety relief valves to establish a
cooling flow path to the containment suppression pool. That
evaluated accident does not result in any fuel failure. The proposed
change will not result in an increase in the probability of fuel
failures. The evaluated accident does result in normal coolant
activity being released to the suppression pool through the safety
relief valves. The proposed activity will not result in a change in
the release of this coolant activity. The proposed change requires
the removal of the exemption for the second trip system to isolate
SDC and will have no impact on the probability or consequences of
that accident.
Therefore, the operation of WNP-2 in accordance with the
proposed amendment will 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 will not cause any new inadvertent SDC
startup, loss of water inventory or loss of coolant accidents
(LOCA). New or different inadvertent RHR SDC startup accidents are
not possible because this change is only a further restriction on
system operation. The LOCA during Mode 3 is bounded by the LOCA
defined for Modes 1 and 2. No new primary system LOCA can be
initiated because of this change.
Therefore, the operation of WNP-2 in accordance with the
proposed amendment 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 a margin of safety.
The removal of an exemption for the second trip system, as
proposed by this change, will increase the probability that leaks
and high pressure will be isolated. Therefore, operation of WNP-2 in
accordance with the proposed amendment will not decrease the margin
of safety. Therefore, the operation of WNP-2 in accordance with the
proposed amendment will not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1 (RBS), West Feliciana Parish, Louisiana
Date of amendment request: October 25, 1999.
Description of amendment request: The proposed license amendment
would revise the reactor pressure vessel (RPV) surveillance capsule
withdrawal schedule for the River Bend Station. The first surveillance
capsule would be withdrawn at 13.4 effective full power years (EFPY)
rather than 10.4 EFPY.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Pressure-temperature (P/T) limits (RBS Technical Specifications
Figure 3.4.11-1) are imposed on the reactor coolant system to ensure
that adequate safety margins against nonductile or rapidly
propagating failure exist during normal operation, anticipated
operational occurrences, and system hydrostatic tests. The P/T
limits are related to the nil-ductility reference temperature,
RTNDT, as described in ASME [American Society of
Mechanical Engineers] Section III, Appendix G. Changes in the
fracture toughness properties of RPV beltline materials, resulting
from the neutron irradiation and the thermal environment, are
monitored by a surveillance program in compliance with the
requirements of 10 CFR [Part] 50, Appendix H. The effect of neutron
fluence on the shift in the nil-ductility reference temperature of
pressure vessel steel is predicted by methods given in RG
[Regulatory Guide] 1.99, [Revision] 2.
River Bend's current P/T limits, as well as those for the
planned increase in reactor thermal power (``Power Uprate''), were
established based on adjusted reference temperatures developed in
accordance with the procedures prescribed in RG 1.99, [Revision] 2,
Regulatory Position 1. Calculation of adjusted reference temperature
by these procedures includes a margin term to ensure conservative,
upper-bound values are used for the calculation of the P/T limits.
Revision of the first capsule withdrawal schedule will not affect
the P/T limits because they will continue to be established in
accordance with Regulatory Position 1 or other NRC [Nuclear
Regulatory Commission]-approved procedures. When permitted (two or
more credible surveillance data sets available), Regulatory Position
2 (or other NRC-approved) methods for determining adjusted reference
temperature will be followed.
This change is not related to any accidents previously
evaluated. The proposed change is a revision of the first
surveillance capsule withdrawal time, identified in TRM [Technical
Requirements Manual] Table 3.4.11-1, from 10.4 EFPY to 13.4 EFPY.
This change will not affect P/T limits as given in RBS Technical
Specifications Figure 3.4.11-1 or USAR Figures 5.3-4a and 5.3-4b.
This change will not affect any plant safety limits or limiting
conditions of operation. The proposed change will not affect reactor
pressure vessel performance as no physical changes are involved and
RBS vessel P/T limits will remain conservative in accordance with RG
1.99, [Revision] 2 requirements. The proposed change will not cause
the reactor pressure vessel or interfacing systems to be operated
outside of their design or testing limits. Also, the proposed change
will not alter any assumptions previously made in evaluating the
radiological consequences of accidents. Therefore, the probability
or
[[Page 70084]]
consequences of accidents previously evaluated will not be increased
by the proposed change.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change revises the first RPV material surveillance
capsule withdrawal time in TRM Table 3.4.11-1 from 10.4 EFPY to 13.4
EFPY. This proposed change does not involve a modification of the
design of plant structures, systems, or components. The proposed
change will not impact the manner in which the plant is operated as
plant operating and testing procedures will not be affected by the
change. The proposed change will not degrade the reliability of
structures, systems, or components important to safety (ITS) as
equipment protection features will not be deleted or modified,
equipment redundancy or independence will not be reduced, supporting
system performance will not be downgraded, the frequency of
operation of ITS equipment will not be increased, and increased or
more severe testing of ITS equipment will not be imposed. No new
accident types or failure modes will be introduced as a result of
the proposed change. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from that
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
As stated in Section 5.3.2 of the River Bend Safety Evaluation
Report (NUREG-0989), ``Appendices G and H of 10 CFR [Part] 50
describe the conditions that require pressure-temperature limits and
provide the general bases for these limits. These appendices
specifically require that pressure-temperature limits must provide
safety margins at least as great as those commended in the ASME Code
[American Society of Mechanical Engineers Boiler and Pressure Vessel
Code], Section III, Appendix G. * * * Until the results from the
reactor vessel surveillance program become available, the staff will
use Regulatory Guide (RG) 1.99, Revision 1 [now Revision 2], to
predict the amount of neutron irradiation damage.* * * The use of
operating limits based on these criteria--as defined by applicable
regulations, codes, and standards--will provide reasonable assurance
that nonductile or rapidly propagating failure will not occur, and
will constitute an acceptable basis for satisfying the applicable
requirements of General Design Criteria (GDC) 31.''
Bases for RBS Technical Specification 3.4.11 states: ``The P/T
limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid encountering
pressure, temperature, and temperature rate of change conditions
that might cause undetected flaws to propagate and cause nonductile
failure of the RCPB [reactor coolant pressure boundary], a condition
that is unanalyzed. * * * Since the P/T limits are not derived from
any DBA, there are no acceptance limits related to the P/T limits.
Rather, the P/T limits are acceptance limits themselves since they
preclude operation in an unanalyzed condition.''
The proposed change will not affect any safety limits, limiting
safety system settings, or limiting conditions of operation. The
proposed change does not represent a change in initial conditions,
or in a system response time, or in any other parameter affecting
the course of an accident analysis supporting the Bases of any
Technical Specification. The proposed change does not involve
revision of the P/T limits but rather a revision of the withdrawal
time for the first surveillance capsule. The current P/T limits (and
proposed P/T limits for Power Uprate) were established based on
adjusted reference temperatures for vessel beltline materials
calculated in accordance with Regulatory Position 1 of RG 1.99,
[Revision] 2. P/T limits will continue to be revised as necessary
for changes in adjusted reference temperature due to changes in
fluence according to Regulatory Position 1 until two or more
credible surveillance data sets become available. When two or more
credible surveillance data sets become available, P/T limits will be
revised as prescribed by Regulatory Position 2 of RG 1.99,
[Revision] 2, or other NRC-approved guidance. Therefore, the
proposed changes do not involve a significant reduction in any
margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 29, 1999.
Description of amendment request: The proposed license amendment
would change the River Bend Station (RBS) Updated Safety Analysis
Report (USAR), Sections 6.2 and 15.6, to incorporate a revision to the
calculation of radiological doses following a loss-of-coolant-accident
(LOCA). The LOCA dose calculation was revised as a result of (1) an
increase in the calculated positive pressure period (PPP) to account
for a new phenomenon identified in Information Notice (IN) 88-76, (2) a
more conservative Suppression Pool water volume value, (3) an
additional and more conservative liquid leakage term identified in IN
91-56, and (4) changes to the engineered safety features (ESF) systems
liquid leakage term.
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 changes do not significantly increase the
probability or consequences of an accident previously evaluated.
The analysis changes described by this proposed change to the
USAR are not initiators to events, and therefore do not involve the
probability of an accident. These modifications reflect a revision
to the post-LOCA dose calculation. USAR Section 15.6.5.1.1 states
that ``There are no realistic, identifiable events which would
result in a pipe break inside of containment of the magnitude
required to cause an accident LOCA * * * However, since such an
accident provides an upper limit estimate to the resultant effects
for this category of pipe breaks, it is evaluated without the causes
being identified.'' The analysis itself does not identify an
initiator, nor is it the initiator, of a LOCA. There was no physical
change to the plant. The increase to the positive pressure period
(PPP) was the result of inclusion of phenomena not previously
included in the analysis documented in the SAR [safety analysis
report], and does not have any impact on accident probability. The
inclusion of an NRC [Nuclear Regulatory Commission] Information
Notice (IN) 91-56 unfiltered liquid leakage term is voluntary and
conservative in nature and does not represent an additional failure
that could be construed as an initiator to the event. Therefore,
this change does not increase the probability of occurrence of an
accident evaluated previously in the safety analysis report (SAR).
This proposed change to the USAR does increase the consequences
of an accident, but the increase is not significant. While the
calculated off-site and control room doses of a LOCA did increase in
Revision 1 to the post-LOCA dose calculation (reference 1) [of
Attachment 1 to the License Amendment request, dated October 29,
1999], the dose consequences remain below the regulatory limits of
10 CFR [Part] 100 and 10 CFR [Part] 50, Appendix A, General Design
Criteria (GDC) 19 as approved per NUREG-0989 and License Amendment
98. This change first accounts for the potential effect that
differential temperature has on the PPP assumed in the off-site dose
analysis. It also conservatively includes an additional liquid
leakage term to account for concerns documented in NRC IN 91-56.
Neither of these changes has an appreciable effect on vital area
access doses. Vital area access dose calculations were not revised
since they still conservatively reflect the expected doses discussed
in USAR Section 12.3.2.4. There is no impact on equipment
qualification associated with the proposed change since other gross
conservatisms exist in those calculations (e.g., not crediting
suppression pool scrubbing) compared to the post-LOCA dose
calculations. Reanalysis of the off-site dose calculation
demonstrates that the revised doses are increased only slightly and
remain significantly less than the regulatory
[[Page 70085]]
limits. With the IN 91-56 term excluded, the increases are within
the criteria of less than 10 [percent] of the remaining margin,
which is the criteria to be applied in the revised 10 CFR 50.59 rule
for minimal increases in consequences. With the IN 91-56 term
included, only the 30 day LPZ [low-population zone] thyroid dose
exceeds the ``minimal increase'' criterion. Note the doses
documented in Table 1 [of Attachment 1 to the License Amendment
request, dated October 29, 1999], above, are less than the values
which had been documented in the SAR prior to the implementation and
NRC approval of TS [Technical Specifications] Amendment 98.
Therefore, this change does not significantly increase the
consequences of an accident previously evaluated in the SAR.
2. The proposed changes would not create the possibility of a
new or different kind of accident from any [previously] analyzed.
This change does not represent a physical change to the plant.
It does not involve initiators to any events in the SAR, nor does
the activity create the possibility for any new accidents. Rather,
this change is a result of the evaluation of the most limiting LOCA
which can occur at River Bend. Therefore, this change involves no
new system interactions and does not create the possibility of an
accident of a different type than those presently evaluated in the
SAR.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The off-site dose consequences are calculated in accordance with
regulatory guidance found in Regulatory Guide 1.3 and the SRP
[Standard Review Plan], consistent with the analyses submitted to
and approved by the NRC in support of Technical Specification
Amendment 98. It is conservatively assumed that 100 [percent] fuel
failure occurs instantaneously upon a recirculation pipe break, thus
2 of the 3 fission product barriers are immediately eliminated.
These assumptions are made without any causes for the failures being
identified. Containment is assumed to leak at its maximum allowable
leakage rate (0.26 [percent] per day) for the duration of the event.
Other leakage terms, such as engineered safety feature (ESF)
leakage, are assumed to be equal to the Technical Specification
limit. Since assumptions are made in accordance with Technical
Specification allowable values and regulatory guidance, this change
does not reduce the margin of safety as defined in the basis for any
RBS Technical Specification.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 29, 1998, as supplemented by
letters dated July 29, October 28, and November 11, 1999.
Description of amendment request: The amendment will revise
Technical Specification 6.9.1.11.1 by replacing the existing reference
to the Asea Brown Boveri-Combustion Engineering, Inc. (ABB CE), small
break loss-of-coolant (SBLOCA) accident emergency core cooling system
(ECCS) performance evaluation model with the revised model described in
the topical report CENPD-137, Supplement 2, P-A, April 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:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The SBLOCA ECCS performance evaluation is conducted to
demonstrate conformance of light water nuclear power reactors to the
ECCS acceptance criteria of 10 CFR 50.46. The proposed change is
associated with an analysis performed using the new Supplement 2
version of the ABB CE SBLOCA Model (S2M). The primary objective of
the analysis using the new model was to determine the impact of a
reduction in High Pressure Safety Injection (HPSI) pump flow rate
due to increased surveillance test measurement uncertainty. NRC
approval of the new S2M model for use in licensing applications of
CE design pressurized water reactors was obtained on December 16,
1997 (Reference 1) [of license amendment request dated July 29,
1998].
A comparison of the Waterford 3 results for the limiting SBLOCA
scenario using the new S2M model against the criteria of 10 CFR
50.46(b) is summarized below:
------------------------------------------------------------------------
Parameter Result Criterion
------------------------------------------------------------------------
Peak Cladding Temperature...... 1929 deg.F......... 2200 deg.F
Maximum Cladding Oxidation..... 8.09%.............. 17%
Core-wide Cladding Oxidation... <0.58%............. 1%
Coolable Geometry Maintained... Yes................ Yes
------------------------------------------------------------------------
These results remain within the criteria of 10 CFR 50.46. Thus,
application of the new S2M model to the ECCS at Waterford will not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No.
The proposed change will not create any new system connections
or interactions. Thus, no new modes of failure are introduced. The
revised methods used in the new SBLOCA evaluation model and their
impact has been reviewed and approved by the NRC (Reference 1) [of
license amendment request dated July 29, 1998]. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The proposed change does not alter the ability of the ECCS to
maintain compliance with 10 CFR 50.46 criteria. The revised methods
used in the new SBLOCA evaluation model and their impact has been
reviewed and approved by the NRC (Reference 1) [of license amendment
request dated July 29, 1998]. Therefore, the 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: N.S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: September 7, 1999.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 3/4.3.2.1, ``Safety
Features
[[Page 70086]]
Actuation System Instrumentation,'' Table 3.3-4, ``Safety Features
Actuation System Instrumentation Trip Setpoints,'' to remove the ``Trip
Setpoint'' values and modify the ``Allowable Values'' for Containment
Pressure-High and Containment Pressure-High-High, and would change TS
3/4.3.2, ``Reactor Protection System and Safety System
Instrumentation,'' to reflect the above change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the
proposed changes and determined that a significant hazards
consideration does not exist because operation of the Davis-Besse
Nuclear Power Station, Unit No. 1, in accordance with these changes
would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the proposed changes do not
change any accident initiator, initiating condition, or assumption.
The proposed changes would revise Technical Specification (TS)
Table 3.3-4, Safety Features Actuation System Instrumentation Trip
Setpoints, to administratively remove from TS the ``Trip Setpoint''
values for Instrument String Functional Unit ``b'', Containment
Pressure--High, and Functional Unit ``c'', Containment Pressure--
High-High, and also modify the TS ``Allowable Values'' entry for
these same Functional Units, consistent with updated calculations
using current setpoint methodology. The Trip Setpoint values removed
from TS will be maintained in DBNPS-controlled documents. The
proposed changes to Limiting Condition for Operation (LCO) 3.3.2.1
and Bases 3/4.3.1 and 3/4.3.2 are associated with these changes.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
invalidate assumptions used in evaluating the radiological
consequences of an accident, do not alter the source term or
containment isolation, and do not provide a new radiation release
path or alter radiological consequences.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not introduce a new or different accident initiator or
introduce a new or different equipment failure mode or mechanism.
3. Not involve a significant reduction in a margin of safety
because the proposed changes establish an error analysis that has
been shown to adequately preserve 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: November 2, 1999.
Description of amendment request: The proposed amendment would: (1)
relocate the Boric Acid Addition Tank System (BAAS) and Borated Water
Storage Tank requirements of Technical Specification (TS) 3/4.1.2.8,
Reactivity Control Systems--Borated Water Sources--Shutdown, in their
entirety to the Davis-Besse Nuclear Power Station Updated Safety
Analysis Report (USAR) Technical Requirements Manual (TRM); (2)
relocate the BAAS requirements of TS 3/4.1.2.9, Reactivity Control
Systems--Borated Water Sources--Operating, to the USAR TRM, except for
portions applicable to the BWST which are proposed to be deleted
because they are redundant to the existing provisions of TS 3/4.5.4,
Emergency Core Cooling Systems--Borated Water Storage Tank; (3) modify
TS 3/4.1.2.1, Reactivity Control Systems--Borated Water Sources--
Shutdown, by deleting references to TS 3.1.2.8; (4) incorporate
corresponding changes to the TS index; and (5) incorporate
corresponding changes to the TS Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no change is being made to any
accident initiator. No previously analyzed accident scenario is
changed, and initiating conditions remain as previously analyzed.
The proposed changes would relocate the Boric Acid Addition
System (BAAS) and Borated Water Storage Tank (BWST) requirements of
Technical Specification (TS) 3/4.1.2.8 in their entirety to the
Davis-Besse Nuclear Power Station (DBNPS) Updated Safety Analysis
Report (USAR) Technical Requirements Manual (TRM). The proposed
changes would also relocate the BAAS requirements of TS 3/4.1.2.9 to
the USAR TRM. The portions of TS 3/4.1.2.9 applicable to the BWST
are proposed to be deleted because they are completely redundant to
the existing provisions of TS 3/4.5.4, Emergency Core Cooling
Systems--Borated Water Storage Tank. Associated with these changes,
TS 3/4.1.2.1 is proposed to be revised to delete references to TS
3.1.2.8. The appropriate changes to the TS Index are also proposed,
as well as changes to TS Bases 3/4.1.2. The proposed changes are
also consistent with the improved ``Standard Technical
Specifications--Babcock and Wilcox Plants,'' NUREG-1430, Revision 1.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
affect accident conditions or assumptions used in evaluating the
radiological consequences of an accident. The proposed changes do
not alter the source term, containment isolation or allowable
radiological releases.
The chemical addition system, which includes the BAAS, is not
credited for mitigation of any USAR Chapter 6 or Chapter 15
accidents. The BWST is credited for mitigation of USAR Chapter 6 and
Chapter 15 accidents, as part of the Emergency Core Cooling System
(ECCS). However, the BWST's requirements concerning ECCS are
provided in separate TS 3/4.5.4, that is not proposed for change.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not change the way the plant is operated, and no new or
different failure modes have been defined for any plant system or
component important to safety. No new or different types of failures
or accident initiators are introduced by the proposed changes.
3. Not involve a significant reduction in a margin of safety
because the proposed changes are administrative in nature,
consisting of deletion and/or relocation of certain TS requirements
into licensee-controlled documents, and have no bearing on the
margin of safety which exists in the present TS or USAR.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
[[Page 70087]]
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: November 2, 1999.
Description of amendment request: The proposed amendment would: (1)
modify Technical Specification (TS) 3/4.3.2.1, Safety Features
Actuation System Instrumentation, Table 3.3-4, Safety Features
Actuation System Instrumentation Trip Setpoints, to remove ``Trip
Setpoint'' values for Instrument String Functional Unit ``f,'' Borated
Water Storage Tank (BWST) Level; (2) modify TS 3/4.3.2.1, Table 3.3-4,
Functional Unit ``f,'' Allowable Values, to make it consistent with
updated calculations using current setpoint methodology; (3) modify
Limiting Condition for Operation (LCO) 3.3.2.1, Safety Features
Actuation System Instrumentation to reflect removal of the ``Trip
Setpoint'' for this Functional Unit; (4) change the footnote associated
with TS 3/4.3.2.1, Table 3.3-4, Functional Unit ``f,'' Allowable
Values, to indicate that the Allowable Values apply to the Channel
Functional Test and no longer applies to the Channel Calibration; (5)
modify TS 3/4.1.2.9, Reactivity Control Systems--Borated Water
Sources--Operating, and TS 3/4.5.4, Emergency Core Cooling Systems--
Borated Water Storage Tank, to increase the minimum BWST water level;
and (6) make corresponding changes to TS Bases 3/4.1.2, Boration
Systems, 3/4.3.1 and 3/4.3.2, Reactor Protection System and Safety
System Instrumentation, and 3/4.5.4, Borated Water Storage Tank.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the proposed changes do not
change any accident initiator, initiating condition, or assumption.
The proposed changes would revise Technical Specification (TS)
Table 3.3.4, Safety Features Actuation System Instrumentation Trip
Setpoints, to administratively remove from the TS the ``Trip
Setpoint'' values for Instrument String Functional Unit ``f,''
Borated Water Storage Tank (BWST) Level, and also modify the TS
``Allowable values entry for this same Functional Unit, consistent
with updated calculations using current setpoint methodology. The
Trip Setpoint values removed from the TS will be maintained in
Davis-Besse Nuclear Power Station (DBNPS)-controlled documents. The
proposed changes to Limiting Condition for Operation (LCO) 3.3.2.1
and Bases 3/4.3.1 and 3/4.3.2 are associated with these changes.
Associated with the above changes, TS 3/4.1.2.9 and TS 3/
4.5.4 are proposed to be revised to increase the minimum available
BWST borated water volume requirement as specified in LCO
3.1.2.9.b.1 and LCO 3.5.4.a. The proposed changes to Bases 3/4.1.2
and Bases 3/4.5.4 are associated with these changes. These changes
are consistent with the revised setpoint analyses.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
invalidate assumptions used in evaluating the radiological
consequences of an accident, do not alter the source term or
containment isolation, and do not provide a new radiation release
path.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not introduce a new or different accident initiator or
introduce a new or different equipment failure mode or mechanism.
3. Not involve a significant reduction in a margin of safety
because the proposed changes establish an error analysis that has
been shown to adequately preserve the margin of safety, and the trip
setpoint values removed from the TS will be maintained in the DBNPS
Updated Safety Analysis Report, with proposed changes subject to the
regulatory requirements of 10 CFR 50.59.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: November 8, 1999.
Description of amendment request: The proposed amendment would
relocate Technical Specification (TS) 6.5.1, Station Review Board, and
TS 6.5.2, Company Nuclear Review Board, to Davis-Besse Updated Safety
Analysis Report Chapter 17.2, Quality Assurance During the Operations
Phase, also known as the Quality Assurance Program. The proposed
changes are consistent with the recommendations in NRC Administrative
Letter 95-06, ``Relocation of Technical Specification Administrative
Controls Related to Quality Assurance.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions or assumptions are affected by the proposed changes to
Section 6.0, Administrative Controls, of the Technical
Specifications (TS).
The proposed changes to relocate the detailed listings of TS
Section 6.5.1, Station Review Board (SRB), and TS 6.5.2, Company
Nuclear Review Board (CNRB), to the Davis-Besse Nuclear Power
Station (DBNPS) Quality Assurance Program in Chapter 17 of the
Updated Safety Analysis Report are consistent with the NRC's
guidance in NUREG-1430, ``Standard Technical Specifications--Babcock
and Wilcox Plants,'' Revision 1 and NRC Administrative Letter 95-06,
``Relocation of Technical Specification Administrative Controls
Related to Quality Assurance,'' dated December 12, 1995. These TS
being relocated will remain subject to the controls of other NRC
regulations (e.g., 10 CFR 50.54(a)). The proposed changes to the TS
Index reflect the relocation of TS 6.5.1 and TS 6.5.2. These are
administrative changes that do not reduce the duties or
responsibilities of the SRB and CNRB in ensuring the safe operation
of the DBNPS.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no accident conditions or
assumptions are affected by the proposed changes. As described
above, these changes are consistent with the improved ``Standard
Technical Specifications--Babcock and Wilcox Plants'' (NUREG-1430
Revision 1) and Administrative Letter 95-06, and are administrative
changes. The proposed changes do not alter the source term,
containment isolation, or allowable releases. The proposed changes,
therefore, will not increase the radiological consequences of a
previously evaluated accident.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new
accident initiators or assumptions are introduced by the proposed
changes, which involve the administrative location for listing SRB
and CNRB responsibilities. The proposed changes do not alter any
accident scenarios.
3. Not involve a significant reduction in a margin of safety
because the proposed changes are administrative and do not reduce or
adversely affect the capabilities of any plant structures, systems
or components to perform their nuclear safety functions.
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
[[Page 70088]]
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: November 1, 1999
Description of amendment request: The proposed license amendment is
prescribed by the requested actions of Generic Letter 99-02,
``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The
proposed amendment will modify the existing Ventilation Filter Testing
Program contained in Technical Specification 5.5.7.c by replacing the
reference to ASTM D3803-1986, the standard for charcoal filter testing
for ESF ventilation systems, with ASTM D3803-1989. The proposed
amendment will also incorporate the suggested safety factor for
charcoal filter efficiency regarding methyl iodide penetration.
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 change to reference American Society for Testing
and Materials (ASTM) D3803-1989, ``Standard Test Method for Nuclear-
Grade Activated Carbon,'' for laboratory testing of Engineered
Safety Features (ESF) ventilation systems in lieu of ASTM D3803-1986
is prescribed by the requested actions of Generic Letter (GL) 99-02,
``Laboratory Testing of Nuclear-Grade Activated Charcoal.'' The use
of ASTM D3803-1989 allows for increased accuracy in monitoring the
degradation of ESF ventilation system activated carbon (charcoal)
over time and is a reproducible method for determining the realistic
capability of charcoal. The 1989 standard is endorsed by the NRC and
is considered to be more stringent regarding testing criteria than
the previous referenced standard (1986). GL 99-02 encourages
addressees, if necessary, to amend their Technical Specifications
(TS) to reference ASTM D3803-1989 for charcoal filter laboratory
testing for ESF ventilation systems. In response to the referenced
GL, the proposed change modifies the existing Perry Nuclear Power
Plant (PNPP) Ventilation Filter Testing Program (VFTP) contained in
the PNPP TS to reference ASTM D3803-1989 as the standard for
charcoal filter laboratory testing for ESF ventilation systems. In
addition, the proposed change incorporates the safety factor
suggested within GL 99-02 for charcoal filter efficiency with
respect to methyl iodide penetration. The proposed change provides
assurance for compliance with the current licensing basis regarding
dose limits of General Design Criteria (GDC) 19 of Appendix A to 10
CFR 50 and 10 CFR 100. The proposed change ensures originally stated
design criteria are met and therefore does not affect the precursors
for accidents or transients analyzed in Chapter 15 of the PNPP
Updated Safety Analysis Report (USAR). With the proposed change, the
radiological consequences are the same as previously stated in the
USAR. Therefore, the implementation of 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.
The proposed change to reference ASTM D3803-1989 for the
laboratory testing of charcoal filters of ESF ventilation systems in
lieu of ASTM D3803-1986 is prescribed by the requested actions of GL
99-02. ASTM D3803-1989 is endorsed by the NRC and is considered a
more stringent testing standard than the previous referenced
standard, ASTM D3803-1986. In addition, the proposed change
incorporates the safety factor suggested within GL 99-02 for
charcoal filter efficiency with respect to methyl iodide
penetration. The proposed change provides assurance for compliance
with the current licensing basis regarding dose limits of GDC 19 of
Appendix A to 10 CFR 50 and 10 CFR 100. The proposed change does not
change the assumptions used in any accident analysis and no new or
different kind of accident is created. The proposed change ensures
originally stated design criteria are met and therefore does not
affect the precursors for accidents or transients analyzed in
Chapter 15 of the PNPP USAR. Therefore, the implementation of 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.
The proposed change is prescribed by the requested actions of GL
99-02. The use of ASTM D3803-1989 allows for increased accuracy in
monitoring the degradation of ESF ventilation systems charcoal over
time and is a very accurate and reproducible method for determining
the realistic capability of charcoal. ASTM D3803-1989 is considered
a more stringent testing standard than the previous referenced
standard, ASTM D3803-1986. Additionally, as specified in GL 99-02, a
safety factor of 2 has been utilized in the calculation of the
revised allowable penetration based upon the credited efficiency
approved by the NRC. The proposed change provides assurance for
compliance with the current licensing basis regarding dose limits of
GDC 19 of Appendix A to 10 CFR 50 and 10 CFR 100. Therefore, the
implementation of the proposed change does not involve a 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: November 1, 1999
Description of amendment request: Technical Specification
Surveillance Requirement (SR) 3.6.1.7.4 requires that each containment
spray nozzle be verified unobstructed on a 10-year frequency. The
proposed amendment would revise the frequency for SR 3.6.1.7.4 from
once every 10 years to only those conditions when maintenance is
performed which could result in nozzle blockage.
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 would not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change revises the surveillance frequency from
every 10 years to following maintenance that could result in nozzle
blockage. Analyzed events are initiated by the failure of plant
structures, systems or components. The containment spray system is
not considered as an initiator of any analyzed event. The proposed
change does not have a detrimental impact on the integrity of any
plant structure, system or component that initiates an analyzed
event. The proposed change will not alter the operation of, or
otherwise increase the failure probability of any plant equipment
that initiates an analyzed accident. As a result, the probability of
any accident previously evaluated, is not significantly increased.
The proposed change revises the Surveillance Frequency. Reduced
testing is acceptable where operating experience has shown that
these components usually pass the Surveillance when performed at the
specified interval, thus the frequency is acceptable from a
reliability standpoint. The proposed containment spray nozzle
Surveillance Frequency has been established based on achieving
acceptable levels of equipment
[[Page 70089]]
reliability. This change does not affect the plant design. Due to
the plant design, the spray header is maintained dry and alarmed on
water intrusion. Formation of significant corrosion products is
unlikely. Due to its location at the top of the containment,
introduction of foreign material from exterior to the header is
unlikely. Since maintenance that could introduce foreign material is
the most likely cause for obstruction, testing or inspection
following such maintenance would verify the nozzle(s) being
unobstructed, and the system would be capable of performing its
safety function. As a result, the consequences of any accident
previously evaluated are not significantly affected.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed change would not create the possibility of a new
of different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed change would not involve a significant reduction
in a margin of safety.
The margin of safety for this system is based on the capacity of
the spray headers. Since the system is not susceptible to corrosion
induced obstruction or obstruction from external to the system, and
performance of maintenance on the system would require evaluation of
the potential for nozzle blockage and the need for a test or
inspection, the spray header nozzles will not become blocked in the
event that the safety function is required. Therefore, the capacity
of the system would remain unaffected. Hence, this 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: November 17, 1999.
Description of amendment request: The proposed amendments for St.
Lucie, Units 1 and 2, will revise the current 72-hour action completion
allowed outage time (AOT) specified in Technical Specification (TS)
3.8.1.1, Action ``b,'' to allow 14 days to restore an inoperable
emergency diesel generator set to operable status. The proposed AOT is
based on an integrated review and assessment of plant operations,
deterministic design basis factors, and an evaluation of overall plant
risk using probabilistic safety assessment techniques.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments for St. Lucie Unit 1 and Unit 2 will
extend the action completion/allowed outage time (AOT) for a single
inoperable Emergency Diesel Generator (EDG) from 72 hours to 14
days. The EDGs are designed as backup AC power sources for essential
safety systems in the event of a loss of offsite power. As such, the
EDGs are not accident initiators, and an extended AOT to restore
operability of an inoperable diesel generator would not
significantly increase the probability of occurrence of accidents
previously analyzed.
The proposed technical specification revisions involve the AOT
for a single inoperable EDG, and do not change the conditions,
operating configuration, or minimum amount of operating equipment
assumed in the plant safety analyses for accident mitigation. Plant
defense-in-depth capabilities will be maintained with the proposed
AOT, and the design basis for electric power systems will continue
to conform with 10 CFR 50, Appendix A, General Design Criterion 17.
In addition, a Probability Safety Assessment (PSA) was performed to
quantitatively assess the risk-impact of the proposed amendment for
each unit. The impact on the early radiological release probability
for design basis events was also evaluated and it is concluded that
the risk contribution from this proposed AOT is small and consistent
with regulatory risk-assessment acceptance guidelines. Therefore,
operation of either facility in accordance with its proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments will not change the physical plant or
the modes of operation defined in either facility license. The
changes do not involve the addition of new equipment or the
modification of existing equipment, nor do they alter the design of
St. Lucie plant systems. Therefore, operation of either facility in
accordance with its proposed amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendments are designed to improve EDG reliability
by providing flexibility in the scheduling and performance of
preventive and corrective maintenance activities. The surveillance
intervals or the operability requirements are not changed by the
proposal; only the AOT for a single inoperable EDG will be extended.
The proposed changes do not alter the basis for any technical
specification that is related to the establishment of, or the
maintenance of, a nuclear safety margin, and design defense-in-depth
capabilities are maintained. An integrated assessment of the risk
impact of extending the AOT for a single inoperable EDG has
determined that the risk contribution is small and is within
regulatory guidelines for an acceptable TS change. Therefore,
operation of either facility in accordance with its proposed
amendment would 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida
Date of amendment request: November 23, 1999.
Description of amendment request: The proposed license amendments
are submitted in response to Generic Letter (GL) 99-02, Laboratory
Testing of Nuclear-Grade Activated Charcoal, which requires that
American Society for Testing and Materials (ASTM) D3803-1989 be used
for testing both new and used charcoal in engineered safety feature
applications. The proposed amendments would modify Technical
Specification (TS) 3/4.6.3, EMERGENCY CONTAINMENT FILTERING SYSTEM, TS
3/4.6.6, POST ACCIDENT CONTAINMENT VENT SYSTEM, and TS 3/4.7.5, CONTROL
ROOM EMERGENCY VENTILATION SYSTEM.
[[Page 70090]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The probability of occurrence of an accident previously
evaluated for Turkey Point is not altered by the proposed TS changes
because no physical modifications are being made to the plant.
The proposed change requires that new and used charcoal in the
plant engineered safety feature (ESF) ventilation systems be tested
in accordance with ASTM D3803-1989, at a temperature of 30 deg.C
and a relative humidity of 95%. The use of a new or different test
standard to satisfy the charcoal surveillance test requirement does
not change the radiological consequences of any previously evaluated
accident. The adoption of the ASTM standard will, however, require
that future charcoal samples from the emergency containment filters
be tested for methyl iodide removal rather than elemental iodine
removal as permitted by previous test protocols. The revised test
method will provide a more uniform test program for the ESF filters,
and will not adversely affect the filters affinity for elemental
iodine removal. The adoption of the ASTM standard for laboratory
analysis of the ESF charcoal does not impact the design bases of the
ESF systems, alter post-accident source terms, or modify the removal
efficiencies credited in the facility dose calculations.
The ASTM standard is very stringent and has been shown to
provide a more reliable measure of the ability of charcoal to
fulfill its intended design function, i.e., to remove radioiodine in
any chemical form from the attendant plant gas stream, than previous
test protocols. Consequently, the adoption of the ASTM standard for
laboratory analysis of the ESF charcoal will ensure that Turkey
Point is operated in a manner consistent with the licensing basis of
the facility as it relates to the protection of the public and the
control room operators during radiological accidents.
Based on the above, it is concluded that the proposed amendment
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any previously evaluated.
The proposed change does not create a new or different type of
accident for Turkey Point because no physical plant changes are
being made, and no compensatory measures are imposed that would
create a new failure scenario. The proposed change only imposes a
more stringent surveillance requirement for both new and used
charcoal in the plant ESF ventilation systems. Since no new failure
modes are associated with the proposed changes, the activity does
not create the possibility of a new or different kind of accident
from any previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed license amendment adopts a more stringent standard
for performing laboratory surveillance tests on both new and used
charcoal in the ESF ventilation systems. Given the increased
accuracy of the proposed test standard, the amendment also supports
the adoption of revised acceptance criteria having a lower safety
factor to the plant safety analysis limits. The composite change
does not impact the design bases of the ESF systems, alter post-
accident source terms, or modify the removal efficiencies credited
in the facility dose calculations
The margin of safety associated with operation of the ESF
ventilation systems is established by the facility dose calculations
and the acceptance criteria for system performance defined in 10 CFR
100 and Criterion 19 of Appendix A to 10 CFR 50. The proposed
amendments will not change this acceptance criteria nor the
calculated dose limits used to establish the current plant-licensing
basis.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Corriea.
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: October 12, 1999.
Brief description of amendment: The proposed amendment would revise
the Appendix B Environmental Protection Plan of the Crystal River Unit
3 (CR-3) Operating License. The changes incorporate requirements from a
biological opinion (BO) issued by the National Marine Fisheries Service
(NMFS). The BO reviews the effects of the cooling water intake system
on species of sea turtles protected by the Endangered Species Act
(ESA). Additionally, other administrative changes are proposed to
Appendix B.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
The proposed changes to the CR-3 EPP are administrative in
nature and reflect the information provided in the NMFS BO. These
changes do not affect the initial conditions, assumptions, or
conclusions of the CR-3 accident analyses. In addition, the proposed
changes do not affect the operation or performance of any equipment
assumed in the accident analyses. Therefore, the proposed changes
would not significantly increase the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from previously evaluated accidents?
The proposed changes are administrative in nature and reflect
information provided by the NMFS BO regarding the incidental taking
of species of sea turtles protected by the ESA. These changes do not
impact or alter the configuration or operation of the facilities and
do not create any new modes of operation. Therefore, the proposed
changes would not create the possibility of a new or different kind
of accident.
3. Involve a significant reduction in a margin of safety?
As indicated above, the proposed changes do not change the
configuration or operation of the plant and do not affect the CR-3
accident analyses. The proposed changes are administrative in nature
and do not affect any margin of safety for CR-3. Therefore, the
proposed changes would not result in 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: R. Alexander Glenn, General Counsel (MAC-
BT15A), Florida Power Corporation, P. O. Box 14042, St. Petersburg,
Florida 33733-4042.
NRC Section Chief: Richard Correia.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: October 29, 1999.
Description of amendment request: The proposed license amendment
would modify the Technical Specifications (TSs) to: (1) Add operating
limits for make-up tank (MUT) level and pressure in a new figure 3.3.1;
(2) add surveillance requirements for the MUT pressure instrument
channel; (3) change the frequency of calibration for the MUT level
instrument from F (every 24 months) to R (refueling interval); (4)
change the frequency of calibration for
[[Page 70091]]
the high pressure injection (HPI) and low pressure injection (LPI) flow
instruments; and (5) make minor editorial changes.
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 changes do not represent a significant increase
in the probability or consequences of an accident previously
evaluated.
The changes included in this LCA [License Change Application]
impose new requirements for MU/HPI system operation and testing and
extension of calibration frequencies for the MUT level, HPI flow and
LPI flow instruments. These changes could not result in initiation
of any accident previously evaluated. Therefore, the probability of
an accident could not be affected by changes to the MU/HPI system.
As described in the list of benefits for operation with the MU/
HPI cross-connect valves open, listed in Section III.B above
[Section III.B of the October 29, 1999 application], the purpose of
changing the operation of the MU/HPI system was to preclude the
possibility of HPI pump damage. The addition of surveillance
requirements for the MUT pressure instrument and the addition of LCO
[limiting conditions for operation] limits on MUT level and pressure
along with an appropriate action statement and AOT [allowed outage
time] will ensure that gas entrainment of the MUT does not occur.
The proposed change in instrument calibration frequencies will
continue to maintain the required accuracy of the MUT level, HPI
flow, and LPI flow instruments.
Minor editorial changes are included in this request to improve
clarity and readability of the T.S. and could not adversely affect
plant operation.
Therefore, the proposed changes will not adversely impact the
reliability of the MU/HPI system and could not represent a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This LCA does not involve the addition of any new hardware.
Along with minor editorial changes, the requested changes involve
MU/HPI system operation and testing, which could only affect RCS
[reactor coolant system] coolant inventory changes during operation
and the ability to provide protection in the event of a Loss of
Coolant Accident (LOCA). The full spectrum of LOCAs has been
evaluated in the FSAR [Final Safety Analysis Report]. Therefore, no
new accident scenarios have been created.
The additional controls on MUT level and pressure provided by
this LCA will ensure that a malfunction of a different type, gas
entrainment of the MU/HPI pumps, will not occur. These limits on MUT
level and pressure ensure that the initial conditions assumed for
ECCS [emergency core cooling system] operation are maintained. The
T.S. limits maintain the accident analysis initial conditions such
that no operator action is required to meet NPSH [net positive
suction head] or to avoid gas entrainment during ECCS operation with
the postulated single failure as required by the TMI-1 licensing
basis (Reference 14) [of the October 29, 1999, application].
Extension of the calibration frequencies for the HPI level, HPI
flow, and LPI flow will continue to maintain the accuracy of these
instruments and could not create the potential for any new accident
that has not been evaluated.
Minor editorial changes are included in this request to improve
the clarity and readability of the T.S. and could not adversely
affect plant operation.
Therefore, these changes do not create the potential for any
accident different from those that have been evaluated.
3. These proposed changes do not involve a significant reduction
in a margin of safety.
This LCA includes changes to the MU/HPI system operation and
testing and an extension of the calibration frequency for certain
instrument[s]. The requested changes will serve to maintain the
proper system initial conditions to ensure the ability of the MU/HPI
system to provide protection in the event of a Loss of Coolant
Accident (LOCA) and maintain the required instrument accuracy for
the instruments where changes to a refueling interval frequency are
being requested. NRC guidance for addressing the effect on increased
surveillance intervals on instrument drift and safety analysis
assumptions presented in GL [generic letter] 91-04 has been
addressed in enclosure 1A above [of the October 29, 1999,
application].
Minor editorial changes are included in this request to improve
the clarity and readability of the T.S. and could not adversely
affect plant operation.
These changes, which are consistent with the TMI-1 licensing and
design basis requirements, do not result in a degradation of safety
related equipment, and therefore, do 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 request involves no significant hazards consideration.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Sheri R. Peterson.
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: November 17, 1999.
Description of amendment request: The proposed amendments would
revise the Technical Specification (TS) values for methyl iodide
penetration for the main control room environmental control system and
the standby gas treatment system. Also, editorial revisions are being
made to portions of TS Section 5.0 to reference the correct sections of
Regulatory Guide 1.52.
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 changes do not involve a significant increase in
the probability or the consequences of a previously evaluated
accident.
This proposed revision makes changes to Technical Specification
(TS) Section 5.5.7, ``Ventilation Filter Testing Program'' (VFTP).
The references to sections in the Regulatory Guide 1.52, Revision 2
for VFTP are being corrected. Additionally, the proposed revision
also changes the allowable methyl iodide penetration percent for the
carbon in the Standby Gas Treatment (SGT) and the Main Control Room
Environmental Control (MCREC) systems when tested in accordance with
ASTM DS3803-1989. This is based on the values that would be derived
using a factor of safety of 2 between the credited and tested carbon
efficiencies. This safety factor is contained in the Generic Letter
99-02. The Generic Letter allows the reduction of the factor of
safety between the credited and tested carbon efficiencies from 5
(for systems with heaters) and 7 (for systems without heaters) to 2
(for systems with or without heaters) when tested per ASTM D-3803-
1989. Since the factor of safety of 2 is maintained, this change
does not involve a significant increase in the probability or the
consequences of a previously evaluated event. The changes in the
section references to Regulatory Guide 1.52 Revision 2 for the
Ventilation Filter Testing Program (VFTP) are considered to be
editorial corrections.
2. The change does not involve a significant increase in the
probability of or the consequences of an event not previously
analyzed.
This proposed revision makes changes to TS Section 5.5.7,
``Ventilation Filter Testing Program'' (VFTP). The section
references to Regulatory Guide 1.52 Revision 2 for the Ventilation
Filter Testing Program (VFTP) are being corrected. The change in the
allowable methyl iodide penetration percent is based
[[Page 70092]]
on the values that would be derived using the safety factor of 2
contained in Generic Letter 99-02. The Generic Letter will reduce
the factor of safety between the credited and tested carbon
efficiencies from 5 (for systems with heaters) and 7 (for systems
without heaters) to 2 if tested per ASTM D-3803-1989. Since the
credited carbon efficiencies in the dose calculations are not being
compromised, this change will not involve a significant increase in
the probability of, or the consequences of an event not previously
analyzed.
The changes in the section references to Reg. Guide 1.52 are
editorial and thus do not significantly increase the probability of,
or the consequences of a previously unanalyzed event.
3. The change does not significantly reduce the margin of
safety.
The change in the allowable methyl iodide penetration percent
implements the Generic Letter's carbon efficiency safety factor of 2
between the credited and the tested carbon efficiencies. Per the
generic letter, it is acceptable to use this new safety factor since
the new standard is more accurate and demanding than previous ones.
Therefore, the proposed revision will not significantly reduce the
margin of safety. The changes in the section references for
Regulatory Guide 1.52 Revision 2 are considered to be editorial
corrections.
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: Richard L. Emch, Jr.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: November 15, 1999 (TS 99-016).
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) for Watts Bar Unit 1 to: (1)
revise the Watts Bar TS and associated TS Bases for TS 3.6.11.5 to
change the methodology and frequency for sampling the ice condenser ice
bed (stored ice) and (2) add a new TS 3.6.11.7 and associated TS Bases
to address sampling requirements for all ice additions to the ice bed.
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:
A. The Proposed Change Does Not Involve a Significant Increase
in the Probability or Consequences of an Accident Previously
Evaluated.
The only analyzed accidents of possible consideration in regards
to changes potentially affecting the ice condenser are a loss of
coolant accident (LOCA) and a main steam line break (MSLB) inside
containment. However, the ice condenser is not postulated as being
the initiator of any LOCA or MSLB. This is because it is designed to
remain functional following a design basis earthquake, and the ice
condenser does not interconnect or interact with any systems that
interconnect or interact with the reactor coolant or main steam
systems. Since the proposed changes to the TS and TS Bases are
solely to revise and provide clarification of the ice sampling and
chemical analysis requirements, and are not the result of or require
any physical change to the ice condenser, then there can be no
change in the probability of an accident previously evaluated in the
Safety Analysis Report (SAR).
In order for the consequences of any previously evaluated event
to be changed, there would have to be a change in the ice
condenser's physical operation during a LOCA or MSLB, or in the
chemical composition of the stored ice. The proposed changes do not
alter either from existing requirements, except to add an upper
limit on boron concentration, which is the bounding value for the
Hot Leg Switchover timing calculation. Though the frequency of the
existing surveillance requirement for sampling the stored ice is
changed from once every 18 months to once every 54 months, the
sampling requirements are strengthened overall with (1) the
requirement to obtain one randomly selected sample from each ice
condenser bay (24 total samples) rather than nine ``representative''
samples, and (2) the addition of a new surveillance requirement to
verify each addition of ice meets the existing requirements for
boron concentration and pH value. The only other change is to
clarify that each sample of stored ice is individually analyzed for
boron concentration and pH, but that the acceptance criteria for
each parameter is based on the average values obtained for the 24
samples. This is consistent with the bases for the boron
concentration of the ice, which is to ensure the accident analysis
assumptions for containment sump pH and boron concentration are not
altered following complete melting of the ice condenser.
Historically, chemical analysis of the stored ice has had a very
limited number of instances where an individual sample did not meet
the boron or pH requirements, with all subsequent evaluations
(follow up sampling) showing the ice condenser as a whole was well
within these requirements. Requiring chemical analysis of each
sample is provided to preclude the practice of melting all samples
together before performing the analysis, and to ensure the licensee
is alerted to any localized anomalies for investigation and
resolution without the burden of entering a 24 hour ACTION
Condition, provided the averaged results are acceptable. Thus, based
on the above, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
B. The Proposed Change Does Not Create The Possibility Of A New
Or Different Kind Of Accident From Any Accident Previously
Evaluated.
Because the TS and TS Bases changes do not involve any physical
changes to the ice condenser, any physical or chemical changes to
the ice contained therein, or make any changes in the operational or
maintenance aspects of the ice condenser as required by the Tech
Specs, there can be no new accidents created from those already
identified and evaluated.
C. The Proposed Change Does Not Involve A Significant Reduction
In A Margin Of Safety.
The ice condenser Technical Specifications ensure that during a
LOCA or SLB the ice condenser will initially pass sufficient air and
steam mass to preclude over pressurizing lower containment, that it
will absorb sufficient heat energy initially and over a prescribed
time period to assist in precluding containment vessel failure, and
that it will not alter the bulk containment sump pH and boron
concentration assumed in the accident analysis. Since the proposed
changes do not physically alter the ice condenser, but rather only
serve to strengthen and clarify ice sampling and analysis
requirements, the only area of potential concern is the effect these
changes could have on bulk containment sump pH and boron
concentration following ice melt. However, this is not affected
because there is no change in the existing requirements for pH and
boron concentration, except to add an upper limit on boron
concentration. This upper limit is the bounding value for the Hot
Leg Switchover timing calculation. Averaging the pH and boron values
obtained from analysis of the individual samples taken is not a new
practice, just one that was not consistently used by all ice
condenser plants. Using the averaged values provides an equivalent
bulk value for the ice condenser, which is consistent with the
accident analysis for the bulk pH and boron concentration of the
containment sump following ice melt. Changing the performance
frequency for sampling the stored ice does not reduce any margin of
safety because (1) the newly proposed surveillance (SR 3.6.15.7)
ensures ice additions meet the existing boron concentration and pH
requirements, (2) there are no normal operating mechanisms,
including sublimation, that reduce the ice condenser bulk pH and
boron concentration, and (3) the number of required samples has been
increased from nine to 24 (one randomly selected ice basket per
bay), which is approximately the same number of samples that would
have been taken in the same time period under the existing
requirements. Thus, it can be concluded that the proposed TS and TS
Bases changes do 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
[[Page 70093]]
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 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard Correia.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: November 20, 1998 and July 19, 1999
(TS99-014).
Description of amendment request: The proposed amendment would
revise the Watts Bar Nuclear plant Unit 1 Technical Specifications (TS)
and associated TS Bases to alter the acceptance criteria in
Surveillance Requirement (SR) 3.6.11.4 and to revise the Bases for TS
3.6.12. The changes would replace the current visual inspection
requirement that uses a 0.38 inch ice/frost buildup criterion with a
visual surveillance program that provides an increased confidence level
that flow blockage in ice condenser baskets does not exceed the 15
percent assumed in the accident analyses. The proposed amendment dated
July 19, 1999 is considered to supercede and replace entirely a
proposed amendment dated November 20, 1998 on this same subject.
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:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Neither the TS amendment nor the TS Bases changes can increase
the probability of occurrence of any analyzed accident because they
are not the result or cause of any physical modification to ice
condenser structures, and for the current design of the ice
condenser, there is no correlation between any credible failure of
it and the initiation of any previously analyzed event.
Regarding the consequences of analyzed accidents, the ice
condenser is an engineered safety feature designed, in part, to
limit the containment subcompartment and steel containment vessel
pressures immediately following the initiation of a LOCA [loss-of-
coolant accident] or HELB [high energy line break]. Conservative
subcompartment pressure analysis shows this criteria will be met if
the reduction in the flow area per bay provided for ice condenser
air/steam flow channels is less than or equal to 15 percent, or if
the total flow area blocked within each lumped analysis section is
less than or equal to the 15 percent assumed in the safety analysis.
The present 0.38 inch frost/ice buildup surveillance criteria only
addresses the acceptability of any given flow channel, and has no
direct correlation between flow channels exceeding this criteria and
percent of total flow channel blockage. In fact, it was never the
intent of the current SR to make such a correlation. If problems
were encountered in meeting the 0.38 inch criteria, it was expected
that additional inspection and analysis, such as provided in the
proposed amendment, would be performed to make such a determination.
Verifying an ice bed is left with less than or equal to 15
percent flow channel blockage at the conclusion of a refueling
outage assures the ice bed will remain in an acceptable condition
for the duration of the operating cycle. During the operating cycle,
a certain amount of ice sublimates and reforms as frost on the
colder surfaces in the Ice Condenser. However, frost does not
degrade flow channel area. The surveillance will effectively
demonstrate operability for an allowed 18 month surveillance period.
Therefore, limiting ice bed flow channel blockage to less than or
equal to 15 percent ensures operation is consistent with the
assumptions of the design basis accident (DBA) analyses. Thus, the
proposed amendment for flow blockage determination provides the
necessary assurance that flow channel requirements are met without
additional evaluations, and thus will not increase the consequences
of a LOCA or HELB.
In regard to the TS 3.6.12 Bases change, clarifying that
Condition B does not apply when personnel are standing on or opening
doors for a short duration to perform surveillances or minor
maintenance activities, such as ice removal, does not increase
analyzed accident consequences. These are not new or additional
actions to those performed previously, the probability of an
accident versus the time to perform these actions is small, the
number of personnel involved is small, and their duration is
generally much less than the four hour frequency of Required Action
B.1 (monitor maximum ice condenser temperature). Therefore, these
activities do not adversely affect ice bed sublimation, melting, or
ice condenser flow paths. However, if during these activities any
door is determined to be restrained, not fully closed from a
previous activity, or otherwise not operable, then separate entry
into Condition B is required for each door so identified.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
For such a possibility to exist, there would have to be either a
physical change to the ice condenser, or some change in how it is
operated or physically maintained. None of the above is true for the
proposed TS amendment and TS Bases change.
There is no change to the existing design requirements or
inputs/results of any accident analysis calculations.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
Design Basis Accident analyses have shown that with 85 percent
of the total flow area available (uniformly distributed), the ice
condenser will perform its intended function. Thus, the safety limit
for ice condenser operability is a maximum 15 percent blockage of
flow channels. SR 3.6.11.4 currently uses a specific value of 0.38
inch buildup to determine if unacceptable frost/ice blockage exists
in the ice condenser. However, this specific value does not have a
direct correlation to the safety limit for blockage of ice condenser
flow area. The proposed TS amendment requires more extensive visual
inspection (33 percent of the flow area/bay) than is currently
described (2 flow channels/bay) in the TS Bases for SR 3.6.11.4,
thus providing greater reliability and a direct relationship to the
analytical safety limits. Changing the TS to implement a
surveillance program that is more reliable and uses acceptance
criteria of less than or equal to 15 percent flow blockage, as
allowed by the TMD [transient mass distribution] analysis, will not
reduce the margin of safety of any TS.
Additionally, verifying an ice bed is left with less than or
equal to 15 percent flow channel blockage at the conclusion of a
refueling outage assures the ice bed will remain in an acceptable
condition for the duration of the operating cycle. During the
operating cycle, a certain amount of ice sublimates and reforms as
frost on the colder surfaces in the Ice Condenser. However, frost
has been determined to not degrade flow channel flow area. Thus,
design limits for the continued safe function of containment
subcompartment walls and the steel containment vessel are not
exceeded due to this change.
The change made to TS 3.6.12 Bases does not affect the margin of
safety as defined in any TS as it does not involve design
specifications or acceptance criteria. This change only adds a
clarifying note that entry into Condition B is not required solely
because of actions (standing on and opening intermediate/upper deck
doors) necessary for the performance of required ice condenser
surveillances, maintenance, or routine activities. This does not
preclude entry into Condition B during performance of these
activities should an intermediate deck door or upper deck door
otherwise be determined inoperable.
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 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
[[Page 70094]]
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.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: November 8, 1999.
Description of amendment request: The amendment changed action
statements, definitions, and footnotes pertaining to the Technical
Specifications for primary containment leakage and primary containment
purge system to allow an alternative approach to the existing
requirement.
Date of publication of individual notice in Federal Register:
November 16, 1999 (64 FR 62228).
Expiration date of individual notice: December 16, 1999.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and electronically from the ADAMS Public
Library component on the NRC Web site, http://www.nrc.gov (the
Electronic Reading Room).
CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz
Mill, Pennsylvania
Date of application for amendment: September 7, 1999, as
supplemented on October 1, 1999.
Brief description of amendment: This amendment reassigns the
responsibilities of the Site Manager, who works for the Westinghouse
Electric Company (a contractor to CBS), to the TR-2 Decommissioning
Project Director, who works for CBS.
Date of issuance: November 23, 1999.
Effective Date: November 23, 1999.
Amendment No: 10.
Facility License No. TR-2: This amendment changes the
decommissioning plan.
Date of initial notice in Federal Register: October 20, 1999 (64 FR
56529).
The Commission has issued a Safety Evaluation for this amendment
dated November 23, 1999.
No significant hazards consideration comments received: No.
Commonwealth Edison Company, Docket No. 50-254, Quad Cities Nuclear
Power Station, Unit 1, Rock Island County, Illinois
Date of application for amendment: March 30, 1999.
Brief description of amendment: The amendment revises the Technical
Specifications by changing Surveillance Requirement 4.6.E.2 to allow a
one-time extension of the 18-month requirement to pressure set test or
replace one half of the Main Steam Safety Valves to an interval of 24
months.
Date of issuance: November 30, 1999.
Effective date: Immediately, to be implemented within 60 days.
Amendment No.: 191.
Facility Operating License No. DPR-29: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 5, 1999 (64 FR
24194).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 30, 1999.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: April 6, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) to expand the allowable values for
Interlocks P-6 (Intermediate Range Neutron Flux) and P-10 (Power Range
Neutron Flux) in TS 3.3.1, Table 3.3.1-1, Function 16, Reactor Trip
System Interlocks, as recommended by Westinghouse.
Date of issuance: November 30, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-189; Unit 2-170.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 19, 1999 (64 FR
27319).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 30, 1999.
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 application for amendment: May 5, 1999.
Brief description of amendment: This amendment conforms the license
to reflect the transfer of Operating License NPF-58 for the Perry
Nuclear Power Plant, Unit 1, to the extent held by Duquesne Light
Company, to the Cleveland Electric Illuminating Company as previously
approved by an Order dated September 30, 1999.
Date of issuance: December 3, 1999.
Effective date: December 3, 1999.
[[Page 70095]]
Amendment No.: 108.
Facility Operating License No. NPF-58: This amendment revised the
operating license.
Date of initial notice in Federal Register: June 14, 1999 (64 FR
31879).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 30, 1999.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida
Date of application for amendments: July 27, 1999, as supplemented
October 4, 1999.
Brief description of amendments: Revises the Technical
Specifications (TS) to extend the allowed outage time, on a one-time
basis, for an inoperable emergency diesel generator from 72 hours to 7
days, to replace the Unit 3 diesel engine radiators prior to April
2000. The revision applies to Turkey Point Unit 3 only, however, Unit 4
is included administratively because the TS are combined for both
Units.
Date of issuance: November 19, 1999.
Effective date: As of date of issuance, to be implemented prior to
April 2000.
Amendment Nos.: 202 and 196.
Facility Operating License Nos. DPR-31 and DPR-41: Amendments
revised the TS.
Date of initial notice in Federal Register: August 25, 1999 (64 FR
46441). The supplemental letter of October 4, 1999, provided
clarification information that did not change the original no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 19, 1999.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: October 8, 1998.
Brief description of amendments: The proposed amendments would
change the Technical Specifications for both units to place tighter
restrictions on the allowed outage time for the refueling water storage
tank water level instrumentation.
Date of issuance: November 30, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 232 and 215.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 31, 1999 (64 FR
47532). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 30, 1999.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: September 10, 1999.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3/4.4.7 so that the surveillance requirement does
not need to be performed when the reactor is defueled with no forced
circulation. The revision to TS 3/4.4.7 also includes changes to Tables
3.4-1 and 4.4-3. TS Table 4.4-3 is revised to change the reactor
coolant system (RCS) chemistry sampling frequency from three times per
7 days with a maximum interval of 72 hours to a frequency of at least
once per 72 hours. An editorial change to Unit 1 Tables 3.4-1 and 4.4-3
relocates the asterisk for the footnote to a position adjacent to the
parameter ``dissolved oxygen,'' from its current position next to the
allowable chemistry limit in Table 3.4-1 and the analysis frequency in
Table 4.4-3. An editorial change also corrects the footnote for Table
3.4-1 for Unit 1 and Unit 2 by making the word ``limit'' plural, as it
applies to both the steady-state and transient limits. Surveillance
Requirement 4.11.2.2 is revised to delete the phrase ``by analysis of
the Reactor Coolant System noble gases.''
Date of issuance: November 19, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 231 and 214.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 6, 1999 (64 FR
54376).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 19, 1999.
No significant hazards consideration comments received: No.
PECO Energy Company, Docket No. 50-352, Limerick Generating Station,
Unit 1, Montgomery County, Pennsylvania
Date of application for amendment: June 7, 1999.
Brief description of amendment: The amendment revised the technical
specifications (TSs) to reflect the permanent deactivation in the
closed position of the ``wet'' instrument reference leg isolation valve
HV-61-102. Specifically, TS Table 3.6.3.1, ``Primary Containment
Isolation Valve,'' and its associated notations were revised to reflect
this current plant configuration.
Date of issuance: November 18, 1999.
Effective date: As of its date of issuance and shall be implemented
within 30 days.
Amendment No.: 138.
Facility Operating License No. NPF-39. This amendment revised the
TSs.
Date of initial notice in Federal Register: October 6, 1999 (64 FR
54380).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 18, 1999.
No significant hazards consideration comments received: No.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: January 15, 1999, as
supplemented January 18 and October 22, 1999.
Brief description of amendment: The amendment provides a revision
to the Technical Specifications for the FitzPatrick Nuclear Power Plant
by modifying the description of what constitutes an acceptable Local
Power Range Monitor calibration.
Date of issuance: November 22, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 257.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 10, 1999 (64 FR
11965).
The January 18, 1999, and October 22, 1999, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 22, 1999.
No significant hazards consideration comments received: No.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: June 22, 1999.
[[Page 70096]]
Brief description of amendment: This amendment changes the
Technical Specifications by extending the pressure-temperature (P-T)
limit curves to 24 effective full-power years (EFPY) and 32 EFPY. The
current P-T limit curves are valid through 16 EFPY.
Date of issuance: November 29, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 258.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43775).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 29, 1999.
No significant hazards consideration comments received: No.
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 amendments request: June 30 1997, as supplemented by
letters of February 22, March 19, June 30, and October 4, 1999.
Brief Description of amendments: The amendments change the
Technical Specifications (TS) to clarify surveillance requirements for
the control room emergency filtration system, penetration room
filtration system, and related storage pool ventilation system. The
changes also revised the required number of radiation monitoring
instrumentation channels, and deleted the containment purge exhaust
filter TS.
Date of issuance: November 23, 1999.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 145 and 136.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: September 1, 1999 (64
FR 47870).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 23, 1999.
No significant hazards consideration comments received: No.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: September 21, 1999.
Brief description of amendment: The amendment increases the
required volume of stored fuel in the diesel fuel oil storage tank as a
result of a conservative recalculation of diesel generator fuel
consumption.
Date of Issuance: November 22, 1999.
Effective date: As of its date of issuance, and shall be
implemented within 30 days.
Amendment No.: 180.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 20, 1999 (64 FR
56537). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated November 22, 1999.
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: September 21, 1999, as supplemented by
letter dated November 5, 1999.
Brief description of amendment: The amendment extended the
effective full implementation date by six months, from December 31,
1999, to June 30, 2000, for Amendment No. 120 issued March 22, 1999,
that approved a modification to increase the storage capacity of spent
fuel assemblies at the site. The extension is due to delays fabricating
and installing the new fuel storage racks.
Date of issuance: November 30, 1999.
Effective date: November 30, 1999, to be implemented by June 30,
2000.
Amendment No.: 129.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 20, 1999 (64 FR
56538). The supplemental letter of November 5, 1999, provided
additional clarifying information, 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
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 30, 1999.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
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 for the
amendment 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.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an
[[Page 70097]]
opportunity for public comment. If comments have been requested, it is
so stated. In either event, the State has been consulted by telephone
whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective 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
application for amendment, (2) the amendment to Facility Operating
License, 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, the Gelman Building, 2120 L Street, NW., Washington, DC, and
electronically from the ADAMS Public Library component on the NRC Web
site, http://www.nrc.gov (the Electronic Reading Room).
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By January 14, 1999, 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 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and electronically from the ADAMS Public Library
component on the NRC Web site, http://www.nrc.gov (the Electronic
Reading Room). If a request for a hearing or petition for leave to
intervene is filed by the above date, the Commission or an Atomic
Safety and Licensing Board, designated by the Commission or by the
Chairman of the Atomic Safety and Licensing Board Panel, will rule on
the request and/or petition; and the Secretary or the designated Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW, Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: November 17, 1999.
Brief description of amendments: The amendments revised the
Technical Specifications to modify the definition
[[Page 70098]]
of steam generator repair limit for axial tube imperfections detected
between the primary side surface of the tube sheet clad and the end of
the tube.
Date of Issuance: December 3, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1-308; Unit 2-308; Unit 3-308.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. The NRC published a public notice of the proposed
amendments, issued a proposed finding of no significant hazards
consideration and requested that any comments on the proposed no
significant hazards consideration be provided to the staff by the close
of business on December 2, 1999. The notice was published in the
``Greenville News,'' Greenville, SC; and the ``Anderson Independent-
Mail,'' Anderson, SC, on November 24, 1999. No comments have been
received.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, consultation with the State of South Carolina,
and final no significant hazards consideration determination are
contained in a Safety Evaluation dated December 3, 1999.
Attorney for licensee: Richard W. Blackburn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington DC 20005.
NRC Section Chief: Richard L. Emch, Jr.
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 application for amendments: November 10, 1999 (PCN-510).
Brief description of amendments: The amendments modify the
Technical Specification Limiting Condition for Operation 3.4.9.b to
delete the phrase stating that two groups of pressurizer heaters be
``capable of being powered from an emergency power supply.
Date of issuance: November 22, 1999.
Effective date: November 22, 1999.
Amendment Nos.: Unit 2-161; Unit 3-152.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. The NRC published a public notice of the proposed
amendments, issued a proposed finding of no significant hazards
consideration, and requested that any comments on the proposed no
significant hazards consideration be provided to the staff by close of
business November 19 , 1999. The notice was published in the ORANGE
COUNTY REGISTER on November 15-16, 1999. No public comments were
received.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated
November 22, 1999.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Dated at Rockville, Maryland, this 8th day of December 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-32311 Filed 12-14-99; 8:45 am]
BILLING CODE 7590-01-P