[Federal Register Volume 64, Number 249 (Wednesday, December 29, 1999)]
[Notices]
[Pages 73083-73108]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-33684]
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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 December 4, 1999, through December 17, 1999.
The last biweekly notice was published on December 15, 1999 (64 FR
70077).
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
[[Page 73084]]
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
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 28, 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,
[[Page 73085]]
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 Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: November 22, 1999.
Description of amendments request: The proposed amendment revises
Technical Specification (TS) 5.5.11, ``Ventilation Filter Testing
Program'' for laboratory testing of charcoal in Clavert Cliffs
engineered safety feature (ESF) ventilation systems to reference the
latest charcoal testing standard (American Society for Testing and
Materials [ASTM] D3803-1989, ``Standard Test Method for Nuclear-Grade
Activated Carbon''). This TS change was requested by the Nuclear
Regulatory Commission (NRC) in Generic Letter 99-02, ``Laboratory
Testing of Nuclear-Grade Activated Charcoal,'' and is based on the
NRC's determination that testing nuclear-grade activated charcoal to
standards other than ASTM D3803-1989 does not provide assurance for
complying with the current licensing basis as it relates to the dose
limits of General Design Criterion 19 of Appendix A to Part 50 of Title
10 of the Code of Federal Regulations (10 CFR) and Subpart A of 10 CFR
Part 100. The generic letter provided a sample TS that the NRC
considers acceptable. The proposed revision to TS 5.5.11 meets the
intent of the sample TS. Specifically, the proposed change removes the
reference to testing in accordance with American National Standards
Institute N510-1975 and changes the allowable methyl iodide penetration
to an acceptance criterion that is derived from applying a safety
factor of two to the charcoal filter efficiency assumed in Calvert
Cliffs design basis dose analysis. The proposed changes will ensure
that the charcoal filters used in ESF ventilation systems will perform
in a manner that is consistent with the particular ESF charcoal
adsorption efficiencies assumed in the analyses of design basis
accidents.
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.
This proposed change makes changes to the methods, test
conditions, and acceptance criteria associated with the performance
of the laboratory tests of charcoal samples. The effected equipment
is used to mitigate the consequences of an accident and are not
accident initiators. This proposed change does not make any changes
to the method of obtaining the charcoal sample. No structural
changes or modifications are being made to the ESF ventilation
equipment. This proposed change does not make any changes to
equipment, procedures, or processes that increase the likelihood of
an accident. Therefore, this proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
The ESF ventilation systems are designed to mitigate the
consequences of accidents. The design basis analysis of the
accidents account to varying degrees for the reduction in airborne
radioactive material provided by the charcoal filters. The proposed
change will change the charcoal filter test protocol to ASTM D3803-
1989. The use of this standard will produce more accurate and
reproducible laboratory test results and provides a more
conservative estimate of charcoal filter capability. The proposed
change makes changes to the methyl iodide penetration acceptance
criteria to ensure that the charcoal filters are capable of
performing their required safety function for the expected operating
cycle. The proposed change will make it more likely that the
charcoal will meet its intended safety function as described in the
Updated Final Safety Analysis Report. Therefore, the proposed change
does not significantly increase the consequences of an accident
previously evaluated.
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 will not make any physical changes to the
plant or changes to the ESF ventilation system operation. The
proposed change is limited to the ESF ventilation system testing
protocol, test conditions, and acceptance criteria. These changes
are administrative in nature. This proposed change does not make any
changes to the method of obtaining the charcoal sample. This
proposed change does not cause any ESF ventilation equipment to be
operated in a new or different manner. No structural changes or
modifications are being made to the ESF ventilation equipment. This
proposed change does not create any new interactions between any
plant components. Therefore, the possibility of a new or different
type of accident is not created by this proposed change.
3. Would not involve a significant reduction in a margin of
safety.
The safety function of the ESF ventilation systems is to
mitigate the consequences of accidents by reducing the potential
release of radioactive material to the environment or the Control
Room following a design basis accident. The TS requirements for
laboratory testing of charcoal samples provides assurance that the
charcoal filters in these systems are capable of reducing airborne
radioactive material to within acceptable limits. The proposed
license amendment requires the use of the latest NRC-accepted
charcoal testing standard and makes changes to the charcoal testing
methyl iodide removal efficiency acceptance limits in accordance
with the formula provided by the NRC in Generic Letter 99-02. The
proposed license amendment continues to provide assurance that the
charcoal filters are capable of reducing airborne radioactive
material to within acceptable limits. 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 Section Chief: Sheri R. Peterson.
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 22, 1999.
Description of amendments request: The Baltimore Gas and Electric
Company (BGE) requests an amendment to implement a change to the
Calvert Cliffs Nuclear Power Plant (CCNPP) Updated Final Safety
Analysis Report (UFSAR) that constitutes an unreviewed safety question
as described in 10 CFR 50.59.
The change revises the information currently provided within the
UFSAR on aircraft and their flight paths for Patuxent River Naval Air
Station (Pax River NAS). The existing information is outdated and does
not reflect current conditions for aircraft utilizing Pax River NAS.
Additionally, the UFSAR will be revised to add information
[[Page 73086]]
pertaining to the corporate helipad located northwest of the plant.
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.
The probability of an aircraft crash was not quantified during
the timeframe of licensing and construction of the plant. As was
noted previously, the Directorate of Licensing at the U.S. Atomic
Energy Commission concurred with Baltimore Gas and Electric
Company's conclusion that no special design provisions were required
to be incorporated into Calvert Cliffs Nuclear Power Plant (CCNPP)
because the probability of an aircraft crash affecting the plant was
acceptably low (implies a probability of less than 10-7/
Year). Therefore, the probability of an aircraft crash affecting the
plant was acceptably low at less than 10-7/year.
The probability of an aircraft accident resulting in
radiological consequences greater than 10 CFR Part 100 exposure
guidelines was considered to still be below the Standard Review Plan
(SRP) (NUREG-0800) level of acceptability of 1.0 x 10-7
per year for CCNPP. The probability of an aircraft accident during
the timeframe of original construction and licensing of the plant
was never quantified. Since today's probability of an aircraft
accident may be higher based on the fact that, at times, aircraft
going into Patuxent River Naval Air Station fly over the plant,
where previously they came no closer than seven miles from the plant
(as described in the UFSAR), the probability of occurrence of an
accident will conservatively be considered to have increased.
However, it should be noted that the probability of an aircraft
accident resulting in radiological consequences greater than 10 CFR
Part 100 exposure guidelines is still considered to be below
1.0 x 10-7 cr/yr, which is acceptable since it is within
SRP Section 3.5.1.6 guidelines. Since the above probability of an
aircraft accident meets the criteria of SRP Section 3.5.1.6, no
additional design or procedural protection is required. Note that
the SRP criteria is only being used as one acceptable method of
evaluating risk. Use of this method is not a commitment to the SRP
and does not incorporate the SRP into our licensing basis.
Changes to the aircraft flight patterns and/or frequency
(probability) have no affect on the design or method of operating
equipment necessary to mitigate the consequences of previously
analyzed accidents. As was noted above, the aircraft hazard was
considered to be acceptable and, therefore, no additional design or
procedural protection is required for the plant. Since the aircraft
hazard is considered acceptable (where additional design features
are not required), it can be concluded that no action assumed to
occur within the accident analysis of CCNPP's Updated Final Safety
Analysis Report Chapter 14 will be degraded or prevented. Therefore,
it is concluded that the current calculated aircraft hazard will not
result in an increase of the consequences of an accident preciously
evaluated in the UFSAR.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
All possible malfunctions have been previously analyzed.
Aircraft hazard was addressed within the original design of the
plant. The frequency/probability of an aircraft crash was considered
to be so low that special design provisions to protect against
aircraft crashes did not have to be considered during construction
of CCNPP. The current calculated aircraft hazard is considered to
still be within SRP Section 3.5.1.6 guidelines. The possibility for
a malfunction of a different type than preciously evaluated in the
UFSAR is not created.
Aircraft accidents were considered within the original plant
design. The probability of an aircraft accident resulting in
radiological consequences greater than 10 CFR Part 100 exposure
guidelines is still considered to be below the level of
acceptability (per SRP Section 3.5.1.6) and no special design
provisions are required. Since an aircraft crash is not a design
basis concern, it is not plausible that the possibility of a new
accident is created that has not been previously evaluated in the
UFSAR. There are also no new challenges to safety-related equipment.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in the margin of
safety.
The probability of an aircraft crash affecting the plant, at the
time of original licensing and construction, was so low that no
special design provisions were needed in the plant for such an
event. Since aircraft hazards did not have to be considered within
the design of the plant, no margin of safety was required or
established for such a hazard. All of the plant equipment and
initial condition assumptions stipulated within the UFSAR Chapter 14
accident analysis would not be affected by such an event.
The calculated probability of an aircraft accident resulting in
radiological consequences greater than 10 CFR Part 100 exposure
guidelines, based on today's aircraft hazard, is considered to be
below the 1.0 x 10-7 per year stipulated within SRP
Section 3.5.1.6. Therefore, there is still no need for special
design provisions within the plant to guard against such an event.
All of the plant equipment and initial condition assumptions
stipulated within the UFSAR Chapter 14 accident analysis remain
unchanged. The plant will continue to operate in such a manner that
will ensure acceptable levels of protection for the health and
safety of the public.
Therefore, this proposed change 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
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 Section Chief: Sheri R. Peterson.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: November 23, 1999
Description of amendments request: The requested amendments would
change Technical Specification (TS) 5.5.7.c.1, ``Ventilation Filter
Testing.'' The testing criteria would be changed consistent with the
NRC request in Generic Letter 99-02, ``Laboratory Testing of Nuclear-
Grade Activated Charcoal.''
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.
The proposed amendment revises TS 5.5.7.c.1 to require testing
of the SGT [Standby Gas Treatment] system charcoal in accordance
with American Society for Testing and Materials (ASTM) D3803-1989,
``Standard Test Method for Nuclear-Grade Activated Carbon.'' Per the
existing TSs, the SGT system charcoal must meet an acceptance
criteria of < 1.0% penetration of methyl iodide when tested at a
relative humidity 70%. CP&L performs this testing in
accordance with the criteria of Regulatory Position C.6.a of
Regulatory Guide 1.52, Revision 1, 1976, ``Design, Testing, and
Maintenance Criteria for Engineered Safety Feature Atmosphere
Cleanup System Air Filtration and Adsorption Units of Light-Water-
Cooled Nuclear Power Plants.'' As stated in Updated Final Safety
Analysis Report, Section 6.5.1.1, the purpose of the SGT system,
along with that of the primary and secondary containment, is to
mitigate accident consequences. It is not associated with any
initiating events and, therefore, cannot affect the probability of
any accident.
ASTM D3803-1989 is an industry accepted standard for charcoal
filter testing. The conditions employed by this standard were
selected to approximate operating or accident conditions of a
nuclear reactor which would severely reduce the performance of
activated carbons. The key difference associated with the two
testing protocols is the testing temperature. Specifically, testing
to a challenge temperature of 30 deg.C per ASTM D3803-1989 versus
80 deg.C per Regulatory
[[Page 73087]]
Guide 1.52 results in a much more stringent test. Testing at a
higher temperature tends to eliminate impurities and moisture from
the sample. This creates the possibility of the charcoal achieving a
slightly higher efficiency than actual. Other parameter changes will
not significantly affect charcoal test performance and will result
in more accurate and reproducible test results.
The proposed TS change also includes a requirement that the test
be performed with a face velocity of 61 fpm. A single BSEP SGT
system train operates at a maximum flow rate of 4200 scfm which
corresponds to a face velocity of 61 fpm. In accordance with Generic
Letter (GL) 99-02, this requirement has been included in TS
5.5.7.c.1.
As recommended by GL 99-02, the proposed amendment incorporates
a safety factor of 2 into the allowed methyl iodide penetration
limit. The existing TS 5.5.7.c.1 acceptance criteria of 99% does not
account for a safety factor. In previous testing, CP&L has applied
the safety factor provided by Regulatory Guide 1.52, Revision 1,
1976, to the laboratory testing results to ensure proper charcoal
performance. The proposed changes to TS 5.5.7.c.1 require that
charcoal samples, tested in accordance with the methodology of ASTM
D3803-1989, show the methyl iodide penetration to be < 0.5%. The
0.5% penetration limit is derived by applying a safety factor of 2
to the 99% filtration efficiency assumed in the current bounding
calculations for offsite radiological dose release limits. As such,
the acceptance criteria of < 0.5% penetration of methyl iodide
ensures that 10 CFR 100 offsite dose limits are not exceeded.
Based on the more stringent testing temperature requirements,
the new face velocity testing requirement, and the acceptance
criteria of < 0.5% penetration of methyl iodide, the proposed change
will not result in an increase in the consequences of an accident
previously evaluated.
2. The proposed license amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes revise the required testing methodology for
SGT system charcoal. The SGT system is not an initiator of any
accident, and no new accident precursors are created due to the
change in the charcoal testing methodology. In addition, the change
does not alter the design, function, or operation of the SGT system.
Therefore, the proposed change to test SGT system charcoal in
accordance with ASTM D3803-1989 will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
The proposed amendment upgrades the SGT system charcoal testing
requirements to those contained in ASTM D3803-1989. The conditions
employed by ASTM D3803-1989 were selected to approximate operating
or accident conditions of a nuclear reactor which could reduce the
performance of activated carbons. The key difference between CP&L's
current testing protocol and ASTM D3803-1989 is the testing
temperature. Specifically, testing to a challenge temperature of
30 deg.C per ASTM D3803-1989 versus 80 deg.C per Regulatory Guide
1.52 results in a much more stringent test.
The proposed TS change also includes a requirement that the test
be performed with a face velocity of 61 fpm. A single BSEP SGT
system train operates at a maximum flow rate of 4200 scfm which
corresponds to a face velocity of 61 fpm. In accordance with GL 99-
02, this requirement has been included in TS 5.5.7.c.1.
As recommended by GL 99-02, the proposed amendment incorporates
a safety factor of 2 into the allowed methyl iodide penetration
limit. The existing TS 5.5.7.c.1 acceptance criteria of 99% does not
account for a safety factor. In previous testing, CP&L has applied
the safety factor provided by Regulatory Guide 1.52, Revision 1,
1976, to the laboratory testing results to ensure proper charcoal
performance. The proposed changes to TS 5.5.7.c.1 require that
charcoal samples, tested in accordance with the methodology of ASTM
D3803-1989, show the methyl iodide penetration to be < 0.5%. The
0.5% penetration limit is derived by applying a safety factor of 2
to the 99% filtration efficiency assumed in the current bounding
calculations for offsite radiological dose release limits. As such,
the acceptance criteria of < 0.5% penetration of methyl iodide
ensures that 10 CFR 100 offsite dose limits are not exceeded.
Based on the more stringent testing temperature requirements,
the new face velocity testing requirement, and the acceptance
criteria of < 0.5% penetration of methyl iodide, the proposed change
does not involve a significant [reduction] in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: November 30, 1999.
Description of amendment request: The amendment revises Technical
Specifications (TS) Section 5.5.11, Ventilation Filter Testing Program
(VFTP) testing requirements. The proposed change requires VFTP testing
be done according to ASTM D3803-1989 protocol in lieu of previous
standards.
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:
Carolina Power & Light (CP&L) Company has evaluated the proposed
Technical Specification change and has concluded that it does not
involve a significant hazards consideration. The CP&L conclusion is
in accordance with the criteria set forth in 10 CFR 50.92. The bases
for the conclusion that the proposed change does not involve a
significant hazards consideration are discussed below.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change to Technical Specification Section 5.5.11,
``Ventilation Filter Testing Program,'' does not involve any
physical alteration of plant systems, structures or components,
changes in parameters governing normal plant operation, or methods
of operation. The proposed change updates the required testing of
Engineered Safety Features (ESF) ventilation filter systems to more
recent standards accepted by the NRC and described in Generic Letter
(GL) 99-02, ``Laboratory Testing of Nuclear-Grade Activated
Charcoal.'' The NRC has found that charcoal filter test protocols
other than American Society for Testing and Materials (ASTM)
standard ASTM D3803-1989 do not assure accurate and reproducible
test results. Since this proposed change references an acceptable
testing standard and provides assurance that the current licensing
basis is met, the proposed change does not involve an increase in
the probability or consequences of an accident previously analyzed.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures or components, changes in parameters
governing normal plant operation, or methods of operation. The
proposed change does not introduce a new mode of operation or
changes in the method of normal plant operation. The proposed change
introduces a new testing standard for ESF ventilation system
charcoal samples removed for testing and does not involve
manipulation of plant systems to perform the charcoal test.
Therefore, the possibility of a new or different kind of accident
from any accident previously evaluated is not created.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change revises the required testing standard for
ESF ventilation charcoal filter systems and does not alter plant
design margins or analysis assumptions as described in the Updated
Final Safety Analysis Report. The proposed change does not affect
any limiting safety system setpoint, calibration method, or setpoint
calculation. The
[[Page 73088]]
proposed change is more restrictive with regard to testing protocol
and less restrictive with respect to the allowed penetration during
testing of the Control Room ventilation system charcoal. However,
the allowed increase in penetration is in accordance with the method
for determining the allowable penetration described in GL 99-02.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: 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.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County, Washington
Date of amendment request: November 18, 1999.
Description of amendment request: The proposed amendment requests a
revision to Technical Specification (TS) 5.5.7.c. The changes would
revise the requirements that (1) a sample of the charcoal absorber for
the standby gas treatment (SGT) system and the control room emergency
filtration (CREF) system be tested in accordance with American Society
for Testing and Materials (ASTM) D3803-1986, ``Standard Test Method for
Nuclear-Grade Activated Carbon'', (2) methyl iodide penetration be less
than a value of .175% for the SGT system and 1.0% for the CREF system,
and (3) charcoal absorber testing be conducted at a relative humidity
of greater than or equal to 70%. As requested by Generic Letter (GL)
99-02, ``Laboratory Testing of Nuclear-Grade Activated Charcoal,''
Energy Northwest proposed that TS 5.5.7.c be revised so that (1)
testing of charcoal absorber samples be in accordance with ASTM D3803-
1989 at a specified temperature of 30 deg. Centigrade (C) [86 deg.
Fahrenheit (F)], (2) methyl iodide penetration to be less than a value
of 0.5% for the SGT system and 2.5% for the CREF system, (3) testing be
performed at 70% relative humidity, and (4) a face velocity of 75 feet-
per-minute (fpm) will be specified for the SGT system. In addition, the
revision to TS 5.5.7.c will note that variations in testing parameters
are permitted in accordance with the guidance in Table 1 and Section
A5.2 of ASTM D3803-1989.
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 SGT System is designed to limit the release of airborne
radioactive contaminants from secondary containment to the
atmosphere within the guidelines of 10 CFR 100 in the event of a DBA
[design basis accident]. The CREF System provides a radiologically
controlled environment from which the plant can be safely operated
following a DBA. The proposed amendment will require that charcoal
from these two ESF [engineered safeguard feature] systems be tested
to the more conservative standards of ASTM D3803-1989. Using the
more conservative ASTM D3803-1989 testing standard will provide no
increase in the probability of an accident previously evaluated.
The staff considers ASTM D3803-1989 to be the most accurate and
most realistic protocol for testing charcoal in ESF ventilation
systems because it offers the greatest assurance of accurately and
consistently determining the capability of the charcoal. Using the
more conservative ASTM D3803-1989 testing standard will provide
greater assurance that the ESF ventilation systems will properly
perform their safety function, thus assuring no increase in the
radiological consequences of a DBA.
Therefore, 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 create a new or different kind of
accident since it only requires that charcoal from the SGT and CREF
safety-related filtration systems be tested to the more conservative
standards of ASTM D3803-1989. Using the more conservative ASTM
D3803-1989 testing standard will provide even greater assurance that
the ESF ventilation systems will properly perform their safety
function, thus helping to minimize the radiological consequences of
a DBA. The increased margin provided by the more conservative
testing standard will assure no new or different kinds of accidents
results from the proposed 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 proposed amendment requires that more conservative ESF
charcoal filter testing criteria be used to verify ESF ventilation
systems are operable. More conservative testing criteria will
provide greater assurance that the ESF ventilation systems will
properly perform their safety function, thus helping to minimize the
radiological consequences of a DBA. Using more conservative testing
criteria will result in maintaining the current 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, N.W., Washington, D.C. 20005-3502.
NRC Section Chief: Stephen Dembek.
Energy Northwest, Docket No. 50-397, WNP-2, Benton County,Washington
Date of amendment request: November 18, 1999.
Description of amendment request: The proposed amendment requests a
revision to subsection 4.3.1.2.b of Technical Specification 4.3, Fuel
Storage. The change would revise the current wording, which describes
the spacing of the fuel in the new fuel racks, with wording that would
limit the number of fuel assemblies that may be stored in the facility
and establish increased spacing limitations for storage of new fuel
assemblies in the racks.
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 does not increase the consequences of any
previously analyzed accident or transient, since the arrangement of
new nuclear fuel in storage racks maintains the effective neutron
multiplication factor much less than 0.95. The change in
configuration requirements will not increase the probability of any
previously analyzed accident, because physical constraints are
installed in the storage racks when new fuel assemblies are
inserted, assuring that only certain cells can be used for storage
of new fuel.
Therefore, 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.
[[Page 73089]]
The proposed change is consistent with a new fuel criticality
analysis performed in support of a previously implemented change to
Section 9.1 of the FSAR. A variety of accidents were considered in
that analysis, and it was determined that the effective neutron
multiplication factor was well below specified limits for any normal
or accident case.
Therefore, operation of WNP-2 in accordance with the proposed
amendment will not create the possibility of a new or different kind
of accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The current wording of Technical Specification 4.3.1.2.b was
determined to not provide sufficient margin of safety to assure that
the requirements of Technical Specification 4.3.1.2.a would be
maintained. The proposed amendment modifies the requirements for new
fuel storage configuration for Technical Specification 4.3.1.2.b, to
assure the margin of safety is maintained for optimum moderation
conditions.
Therefore, 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, N.W., Washington, D.C. 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: August 20, 1999.
Description of amendment request: The proposed amendment request is
to incorporate 17 improvements (identified by Technical Specification
Task Force (TSTF) numbers) to the Improved Standard Technical
Specifications (TSs), NUREG-1434 (for BWR/6 plants such as the Grand
Gulf plant), that was part of the basis for the current improved TSs
for Grand Gulf Nuclear Station (GGNS) that were issued in Amendment 120
dated February 21, 1995. These improvements to the improved TSs for
BWR/6 plants such as GGNS are identified by TSTF numbers and are the
following: (1) TSTF-2, relocate the 10 year sediment cleaning of the
diesel generator fuel storage tank in Surveillance Requirement (SR)
3.8.3.6 to the GGNS Updated Final Safety Analysis Report (UFSAR), (2)
TSTF-5, delete notification, reporting, and restart requirements if a
safety limit is violated in TSs Section 2.2, (3) TSTF-9, relocate the
shutdown margin values in Limiting Conditions for Operation (LCO) 3.1.1
and SR 3.1.1.1 to the Core Operating Limits Report (COLR), (4) TSTF-17,
extension of the testing frequency for the primary containment airlock
interlock mechanism from 184 days to 24 months in SR 3.6.1.2.3 and
deletion of the SR Note, (5) TSTF-18, reword and clarify SR 3.6.4.1.2
to require only one secondary containment access door per access
opening to be closed, (6) TSTF-32, move the requirement to ensure that
``slow'' and withdrawn stuck control rods are appropriately separated
from LCO 3.1.4 requirements to LCO 3.1.3 Condition A Required Actions,
(7) TSTF-33, administrative change to clarify the Completion Time for
LCO 3.1.3 Required Action A.2, (8) TSTF-38, revise and clarify the
visual surveillance in SR 3.8.4.3 for batteries to specify the
inspection is for performance degradation, (9) TSTF-45, revise SRs
3.6.1.3.2 and 3.6.1.3.3 to specify that only Primary Containment
Isolation Valves which are not locked, sealed, or otherwise secured are
required to be verified closed, (10) TSTF-60, exempt LCO 3.4.7 on
Reactor Coolant System Leakage Detection Instrumentation from LCO 3.0.4
which restricts entry into MODES, or specified conditions with required
equipment inoperable, (11) TSTF-104, relocate the discussion of
exceptions in LCO 3.0.4 to the Bases of the TSs, (12) TSTF-118, add a
sentence to the administrative controls program in TSs Administrative
Controls Section 5.5.9 that the provisions of SRs 3.0.2 and 3.0.3
applies to the specified testing frequencies of the Diesel Fuel Oil
Testing Program, (13) TSTF-153, clarify the exception Notes for LCOs
3.4.9, 3.4.10, 3.9.8, and 3.9.9 to be consistent with the requirement
being excepted, (14) TSTF-163, modify SRs 3.8.1.2, 3.8.1.12, 3.8.1.15,
and 3.8.1.20 for diesel generators to provide minimum volt/Hz limits
for the 10-second acceptance and detail the current volt/Hz range as
``steady state'' acceptance criteria, (15) TSTF-166, revise LCO 3.0.6
to explicitly require an evaluation per the Safety Function
Determination Program and delete the statement that ``additional * * *
limitations may be required,'' (16) TSTF-278, LCO 3.8.6 is revised to
require that battery cell parameters be ``within limits,'' the
reference to Table 3.8.6-1 is deleted, and a reference to the table is
added to the Actions Table for LCO 3.8.6, and (17) TSTF-279, delete the
reference to the ``applicable supports'' from the description of the
``Inservice Testing Program'' in the Administrative Controls TSs,
Section 5.5.6. The licensee is proposing the current latest revision
for each TSTF at the time of application with minor exceptions and/or
clarification in some cases.
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 (NSHC). The licensee's NSHC is divided into the following
five categories (which also list the TSTF changes in each category):
administrative changes, less restrictive changes--removed detail, less
restrictive changes--relaxation of required action, less restrictive
changes--deletion of surveillance requirement, and less restrictive
changes--relaxation of surveillance frequency. The licensee's category
NSHCs are presented below:
1. Administrative Changes
These changes involve reformatting, renumbering, and rewording
of [TSs], with no change in intent. Since they do not change the
intent of the [TSs] they are considered to be administrative in
nature. The GGNS is adopting NRC [Nuclear Regulatory Commission]
approved TSTF-5, TSTF-18, TSTF-33, TSTF-38, TSTF-104, TSTF-118,
TSTF-153, TSTF-163, TSTF-166, TSTF-278, and TSTF-279, generic
changes to the Improved Standard Technical Specifications (ISTS) as
outlined in NUREG-1434, ``Standard Technical Specifications, BWR/6
Plants.'' In accordance with the criteria set forth in 10 CFR 50.92,
EOI [Entergy Operations, Inc.] has evaluated these proposed [TSs]
changes and determined they do not represent a significant hazards
consideration. The following is provided in support of this
conclusion.
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change involves reformatting, renumbering, and
rewording the existing [TSs]. The reformatting, renumbering, and
rewording process involves no changes in intent to the [TSs]. The
proposed changes also involve [TSs] requirements, which are purely
administrative in nature. As such, this change does not [a]ffect
initiators of analyzed events or assumed mitigation of accident or
transient events. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or
[[Page 73090]]
different type of equipment will be installed) or changes in methods
governing normal plant operation. The proposed change will not
impose any new or eliminate any old requirements. Thus, this change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
c. Does this change involve a significant reduction in a margin
of safety?
The proposed change will not reduce a margin of safety because
it has no [a]ffect on any safety analyses assumptions. This change
is administrative in nature. Therefore, the change does not involve
a significant reduction in a margin of safety.
2. Less Restrictive Changes--Removed Detail
GGNS is adopting NRC approved TSTF-2, TSTF-9, and TSTF-32
generic changes to the Improved Standard Technical Specifications
(ISTS) as outlined in NUREG-1434, ``Standard Technical
Specifications, BWR/6 Plants.'' The proposed changes involve moving
details out of the [TSs] and into the [TSs] Bases, the UFSAR, or the
Core Operating Limits Report (COLR). The removal of this information
is considered to be less restrictive because it is no longer
controlled by the [TSs] change process. Typically, the information
moved is descriptive in nature and its removal conforms with NUREG-
1434 for format and content.
In accordance with the criteria set forth in 10 CFR 50.92, the
EOI has evaluated these proposed [TSs] changes and determined they
do not represent a significant hazards consideration. The following
is provided in support of this conclusion.
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relocates certain details from the [TSs] to
other documents under regulatory control. The Bases and UFSAR will
be maintained in accordance with 10 CFR 50.59. In addition to 10 CFR
50.59 provisions, the [TSs] Bases are subject to the change control
provisions in the Administrative Controls Chapter of the [TSs]. The
UFSAR is subject to the change control provisions of 10 CFR
50.71(e). The COLR is controlled in accordance with TS[s] 5.6.5. The
controls of TS[s] 5.6.5 will ensure that adequate limits are
maintained and reported to the NRC. Since any changes to these
documents will be evaluated, no significant increase in the
probability or consequences of an accident previously evaluated will
be allowed. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change will not impose or eliminate any requirements, and
adequate control of the information will be maintained. Thus, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
c. Does this change involve a significant reduction in a margin
of safety?
The proposed change will not reduce a margin of safety because
it has no [a]ffect on any safety analysis assumptions. In addition,
the details to be moved from the [TSs] to other documents remain the
same as the existing [TSs]. Since any future changes to these
details will be evaluated, no significant reduction in a margin of
safety will be allowed. A significant reduction in the margin of
safety is not associated with the elimination of the 10 CFR 50.92
requirement for NRC review and approval of future changes to the
relocated details. The proposed change is consistent with the BWR/6
Standard Technical Specifications, NUREG-1434, issued by the NRC
Staff, revising the [TSs] to reflect the approved level of detail,
which indicates that there is no significant reduction in the margin
of safety.
3. Less Restrictive Changes--Relaxation of Required Action
GGNS is adopting NRC approved TSTF-60 generic changes to the
Improved Standard Technical Specifications (ISTS) as outlined in
NUREG-1434, ``Standard Technical Specifications, BWR/6 Plants.'' The
proposed changes involve relaxation of the Required Actions in the
current Technical Specifications (TS).
Upon discovery of a failure to meet an LCO, the TS specifies
Required Actions to be completed for the associated Conditions.
Required Actions of the associated Conditions are used to establish
remedial measures that must be taken in response to the degraded
conditions. These actions minimize the risk associated with
continued operation while providing time to repair inoperable
features. Some of the Required Actions are modified to place the
plant in a MODE in which the LCO does not apply. Adopting Required
Actions from this change is acceptable because the Required Actions
take into account the operability status of redundant systems of
required features, the capacity and capability of the remaining
features, and the compensatory attributes of the Required Actions as
compared to the LCO requirements. These changes have been evaluated
to not be detrimental to plant safety.
In accordance with the criteria set forth in 10 CFR 50.92, the
EOI has evaluated these proposed [TSs] changes and determined they
do not represent a significant hazards consideration. The following
is provided in support of this conclusion.
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relaxes Required Actions. Required Actions
and their associated Completion Times are not initiating conditions
for any accident previously evaluated and the accident analyses do
not assume that required equipment is out of service prior to the
analyzed event. Consequently, the relaxed Required Actions do not
significantly increase the probability of any accident previously
evaluated. The Required Actions in the change have been developed to
provide assurance that appropriate remedial actions are taken in
response to the degraded condition considering the operability
status of the redundant systems of required features, and the
capacity and capability of remaining features while minimizing the
risk associated with continued operation. As a result, the
consequences of any accident previously evaluated are not
significantly increased. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
Required Actions and associated Completion Times in the change have
been evaluated to ensure that no new accident initiators are
introduced. Thus, this change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
c. Does this change involve a significant reduction in a margin
of safety?
The relaxed Required Actions do not involve a significant
reduction in the margin of safety. As provided in the justification,
this change has been evaluated to minimize the risk of continued
operation under the specified Condition, considering the operability
status of the redundant systems of required features, the capacity
and capability of remaining features, a reasonable time for repairs
or replacement of required features, and the low probability of a
DBA [design basis accident] occurring during the repair period.
Therefore, this change does not involve a significant reduction in a
margin of safety.
4. Less Restrictive Changes--Deletion of Surveillance Requirement
GGNS is adopting NRC approved TSTF-45 which is a generic change
to the Improved Standard Technical Specifications (ISTS) as outlined
in NUREG-1434, ``Standard Technical Specifications, BWR/6 Plants.''
The proposed changes involve deletion of [SRs] in the current
Technical Specifications (TS).
The TS require safety systems to be tested and verified Operable
prior to entering applicable operating conditions. These changes
eliminate unnecessary TS [SRs] that do not contribute to
verification that the equipment used to meet the LCO can perform its
required functions. Thus, appropriate equipment continues to be
tested in a manner and at a frequency necessary to give confidence
that the equipment can perform its assumed safety function. These
changes have been evaluated to not be detrimental to plant safety.
In accordance with the criteria set forth in 10 CFR 50.92, the
EOI has evaluated these proposed [TSs] changes and determined they
do not represent a significant hazards consideration. The following
is provided in support of this conclusion.
[[Page 73091]]
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change deletes [SRs]. Surveillance's are not
initiators to any accident previously evaluated. Consequently, the
probability of an accident previously evaluated is not significantly
increased. The equipment being tested is still required to be
Operable and capable of performing the accident mitigation functions
assumed in the accident analysis. As a result, the consequences of
any accident previously evaluated are not significantly [a]ffected.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
remaining [SRs] are consistent with industry practice and are
considered to be sufficient to prevent the removal of the subject
Surveillance's from creating a new or different type of accident.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
c. Does this change involve a significant reduction in a margin
of safety?
The deleted [SRs] do not result in a significant reduction in
the margin of safety. As provided in the justification, the change
has been evaluated to ensure that the deleted [SRs] are not
necessary for verification that the equipment used to meet the LCO
can perform its required functions. Thus, appropriate equipment
continues to be tested in a manner and at a frequency necessary to
give confidence that the equipment can perform its assumed safety
function. Therefore, this change does not involve a significant
reduction in a margin of safety.
5. Less Restrictive Changes--Relaxation of Surveillance Frequency
GGNS is adopting NRC approved TSTF-17 which is a generic change
to the Improved Standard Technical Specifications (ISTS) as outlined
in NUREG-1434, ``Standard Technical Specifications, BWR/6 Plants.''
The proposed changes involve the relaxation of Surveillance
Frequencies in the current Technical Specifications (TS).
Surveillance Frequencies specify time interval requirements for
performing surveillance testing. Increasing the time interval
between Surveillance tests results in decreased equipment
unavailability due to testing which also increases equipment
availability. Reduced testing can result in a safety enhancement
because the unavailability due to testing is reduced and[,] in turn,
reliability of the [a]ffected structure, system or component should
remain constant or increase. Reduced testing is acceptable where
operating experience, industry practice or the industry standards
such as manufacturers' recommendations have shown that these
components usually pass the Surveillance when performed at the
specified interval, thus the frequency is acceptable from a
reliability standpoint. These changes have been found to be
acceptable based on a combination of the above criteria and have
been evaluated to not be detrimental to plant safety.
In accordance with the criteria set forth in 10 CFR 50.92, the
EOI has evaluated these proposed [TSs] changes and determined they
do not represent a significant hazards consideration. The following
is provided in support of this conclusion.
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change relaxes Surveillance Frequencies. The
relaxed Surveillance Frequencies have been established based on
achieving acceptable levels of equipment reliability. Consequently,
equipment which could initiate an accident previously evaluated will
continue to operate as expected and the probability of the
initiation of any accident previously evaluated will not be
significantly increased. The equipment being tested is still
required to be Operable and capable of performing any accident
mitigation functions assumed in the accident analysis. As a result,
the consequences of any accident previously evaluated are not
significantly [a]ffected. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
c. Does this change involve a significant reduction in a margin
of safety?
The relaxed Surveillance Frequencies do not result in a
significant reduction in the margin of safety. As provided in the
justification, the relaxation in the Surveillance Frequency has been
evaluated to ensure that it provides an acceptable level of
equipment reliability. Thus, appropriate equipment continues to be
tested at a Frequency that gives confidence that the equipment can
perform its assumed safety function when required. Therefore, this
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, 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: November 4, 1999. This amendment request
supercedes the licensee's application of June 10, 1999, in its
entirety. (64 FR 38025)
Description of amendment request: The proposed amendment would
remove the existing filter testing requirements of the Technical
Specifications (TSs) and replace them with a reference to the
Ventilation Filter Testing Program which is being added to the
Administrative Controls section of the Davis-Besse TS. The amendment
introduces TS 6.8.4.f, ``Ventilation Filter Testing Program,'' and
removes the specific ventilation filter testing requirements from the
surveillance requirements of TS 3/4.6.4.4, ``Hydrogen Purge System,''
TS 3/4.6.5.1, ``Shield Building Emergency Ventilation System,'' and TS
3/4.7.6.1, ``Control Room Emergency Ventilation System.'' Also included
are supporting Bases changes to TS 3/4.6.4.4, TS 3/4.6.5.1, and TS 3/
4.7.6.1
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 has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power
Station(DBNPS), Unit Number 1, in accordance with this change 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. The replacement of the specific Technical
Specification (TS) ventilation filter testing Surveillance
Requirements for the Containment Hydrogen Purge System (3/4.6.4.4),
Shield Building Emergency Ventilation System (3/4.6.5.1), and the
Control Room Emergency Ventilation System (3/4.7.6.1), with a
reference to the newly created Ventilation Filter Testing Program
contained in TS Administrative Controls Section 6.8.4.f, Ventilation
Filter Testing Program, is a removal and relocation of certain TS
details. The proposed TS 6.8.4.f will, however, add controls to
maintain similar operation, maintenance, testing and system
operability for these three ventilation systems. The TS Bases
changes reflect the use of the Ventilation Filter Testing Program.
The replacement of ASTM D 3803-1979 with ASTM D 3803-1989 for
laboratory testing of the charcoal filter samples reflects the NRC
recommendations in Generic Letter 99-02, ``Laboratory Testing of
Nuclear Grade Activated Charcoal.'' ASTM D 3803-1989 is
[[Page 73092]]
a more stringent testing standard for charcoal filter testing, than
the present standard referenced by the TS.
The increase in allowable charcoal penetration due to the use of
a safety factor of ``2'' is acceptable as a result of using this
more stringent testing standard.
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 increase in allowable
charcoal penetration due to the use of a safety factor of ``2'' is
acceptable as a result of using this more stringent testing
standard. No physical alterations of the DBNPS are involved, nor are
plant operating methods being changed. The proposed changes do not
alter the source term, containment isolation or allowable
radiological releases.
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. No new or
different types of failures or accident initiators are being
introduced by the proposed changes.
3. Not involve a significant reduction in a margin of safety
because there are no significant changes to the initial conditions
contributing to accident severity or consequences. Therefore, there
are no significant reductions in a margin of safety. Testing under
the more restrictive requirements of ASTM D 3803-1989 will continue
to ensure that the ventilation systems will perform their safety
function.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Dade County, Florida
Date of amendment request: December 1, 1999, as supplemented
December 15, 1999.
Description of amendment request: The licensee is requesting to
revise the Turkey Point Plant Physical Security Plan (PSP) to modify
the PSP requirements for compensation of a security computer failure,
and to modify the requirements of the minimum security force staffing.
The December 1, 1999, submittal supersedes two previous submittals
dated March 10 and June 8, 1999, regarding the same subject. As a
result of the proposed changes, License Conditions 3.L. for Turkey
Point Units 3 and 4 Operating Licenses will be updated to reflect the
latest revision to the Physical Security Plan dated December 1, 1999.
In addition, the phrase ``Turkey Point Plant, Units 3 and 4 Security
Plan'' was revised to ``Turkey Point Physical Security Plan.'' The
latter changes are administrative in nature.
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.
These changes will not significantly affect the ability to
detect a Protected Area intrusion. These changes do not affect the
ability of a security response to an overt attack on the plant.
These changes will not affect the ability of the security force to
respond to contingency events. Therefore, the proposed changes do
not affect the probability or consequences of accidents previously
analyzed.
(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 accident previously evaluated.
These changes do not affect the ability of the security force to
defeat the design basis threat. The composition of the response
organization is not effected by these changes.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The demonstrated level of dependability of the security system
ensures that a significant reduction in effectiveness or margin of
safety does not occur.
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. The staff has also reviewed the changes to License
Conditions 3.L. for Turkey Point Units 3 and 4 Operating Licenses, as
well as the change of the security plan title. Based on this review,
the staff finds that the changes are administrative in nature and that
they meet the three criteria discussed above. 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.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 6, 1999.
Description of amendment request: This proposed Technical
Specification TS change will revise the Cooper Nuclear Station (CNS) TS
Sections 1.0, ``Use and Application,'' 3.6, ``Containment Systems,''
Bases 3.0, ``Limiting Condition for Operation (LCO) Applicability,''
Bases 3.6, ``Containment Systems,'' and 5.5, ``Programs and Manuals,''
to adopt the implementation requirements of 10 CFR Part 50, Appendix J,
Option B, for the performance of Type A, B, and C containment leakage
rate testing. Contingent upon the Nuclear Regulatory Commission's
(NRC's) approval of the proposed TS change, the licensee is also
requesting the NRC to grant the withdrawal of two exemptions. These
exemptions were previously granted under Option A to 10 CFR Part 50,
Appendix J; however, under Option B they are no longer required.
The proposed TS change also contains line-item changes for TS
requirements addressing containment airlock interlocks, primary and
secondary containment isolation valves and power-operated automatic
valves. These changes, along with the specific change to implement
Option B, have been previously approved by the NRC through submittals
made by the Nuclear Energy Institute-sponsored TS Task Force.
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:
The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Implement 10 CFR 50 Appendix J, Option B.
There is no increase in the probability or consequences of an
accident since there is no work that would affect containment
integrity. The testing of containment isolation valves and other
containment penetration sealing devices are not postulated as an
accident precursor or initiating event.
The NRC has concluded, prior to approving Option B, that
performance-based testing would eliminate or modify prescriptive
regulatory requirements for which the burden is marginal-to-safety.
Reviews and analyses considered by the NRC are presented in NUREG-
1493, ``Performance-Based Containment Leak-Test Program, Final
Report,'' September 1995 (Attachment 2, Reference 12 [of the October
[[Page 73093]]
6, 1999, application]). The historical leakage rate test results for
Cooper and for the nuclear industry support extension of the testing
frequencies and demonstrate that structural integrity has been
maintained.
Type A testing is capable of determining the total leakage from
both local leakage paths and gross containment leakage paths. The
Type B and C testing has consistently provided accurate leakage
rates for valves and penetrations. Administrative controls govern
maintenance and testing such that there is very low probability that
unacceptable maintenance or alignments can occur. Prior to and
following maintenance on primary containment isolation valves and
penetrations, a local leak rate test is required to be performed. As
a result, Type A testing is not required to accurately quantify the
leakage through containment penetrations.
Extension of testing frequency of containment airlock interlock
mechanism from 18 months to 24 months.
There is no increase in the probability or consequences of an
accident since there is no work that would affect containment
integrity. The testing of containment airlock interlocks, isolation
valves and other containment penetration sealing devices is not
postulated as an accident precursor or initiating event.
This changed the testing of the containment airlock interlocks
from 18 months to 24 months. This testing is only performed during
periods of reactor shut down and the primary containment is de-
inerted. Thus this change plus the allowance from SR [Surveillance
Requirement] 3.0.2, provides a total of 30 months, which corresponds
to the overall airlock leakage test frequency under Option B. In
this fashion, the interlock can be tested in a Mode where the
interlock is not required.
Clarify the Containment Isolation Valve (CIV) surveillance to
apply to only automatic isolation valves.
The Bases for SR 3.6.1.3.5 state that the isolation time test
ensures the valve will isolate in time period less than or equal to
that assumed in the safety analysis. There may be valves credited as
containment isolation valves, which are power operated, that do not
receive a containment isolation signal. These valves do not have an
isolation time as assumed in the accident analyses since they
require operator action. However, these valves are tested in
accordance with the Inservice Test Program as required. Therefore
this change reduces the potential for misinterpreting the
requirements of this SR while maintaining the assumptions of the
accident analysis.
Based on the above discussion, there is no increase in the
probability or consequences of an accident, since this change
provides clarification of the applicability of the SR and has no
affect on those automatic valves with operating times assumed in the
accident analysis.
Allow administrative means of position verification for locked
or sealed valves.
It is sufficient to assume that the initial establishment of
component status (e.g., isolation valve closed) was performed
correctly. Subsequently verification is intended to ensure the
component has not been inadvertently repositioned. Given that the
function of locking, sealing or securing components is to ensure the
same avoidance of inadvertent repositioning, the periodic re-
verification should only be a verification of the administrative
control that ensures that the component remains in the required
state. It would be inappropriate to remove the lock, seal, or other
means of securing the component solely to perform an active
verification of the required state. There is no increase in the
probability or consequences of an accident since the function of
locking, sealing, or securing components is to ensure that these
devices are not inadvertently repositioned.
Therefore, the proposed change described above does not involve
a significant increase in the probability or consequences of an
accident previously evaluated in the USAR [updated safety analysis
report].
The proposed change will not create the possibility of a new or
different kind of accident than evaluated in the USAR.
The proposed change involves individual proposed changes related
to the implementation of 10 CFR 50 Appendix J, Option B, the
extension of testing frequency of the containment airlock interlock,
clarification of the CIV surveillance to apply to only automatic
isolation valves, and the allowance of administrative means of
position verification for locked or sealed valves. The proposed
change does not result in any physical change to plant structures,
systems, or components. The proposed change does not alter the form,
fit, or function of any equipment or components credited in the
accident analyses described in the USAR. The performance history of
containment testing verifies that containment integrity has been
maintained.
The frequency changes allowed by the implementation of the
applicable proposed TS changes will not significantly decrease the
level of confidence in the ability of the containment to limit
offsite doses to allowable values. No accident or malfunction can be
the result of the allowed changes to test schedule or frequency.
Since the proposed changes will not directly impact equipment,
procedures or operations, the changes will not create the
possibility of any new or different kind of accident from any
accident previously evaluated in the USAR.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident.
The proposed change will not involve a significant reduction in
a margin of safety.
The reason for performing containment leakage rate testing is to
assure that the leakage paths are identified, and that any accident
release will be restricted to those paths assumed in the safety
analysis. The purpose for the schedule is to assure that containment
integrity is verified on a periodic basis. Implementation of Option
B to provide flexibility in the scheduled requirements does not mean
that containment integrity will be compromised.
The NRC has concluded, prior to approving Option B, that
performance-based testing would eliminate or modify prescriptive
regulatory requirements for which the burden is marginal-to-safety.
Reviews and analyses considered by the NRC are presented in NUREG-
1493, ``Performance-Based Containment Leak-Test Program, Final
Report,'' September 1995 (Attachment 2, Reference 12). The
historical leakage rate test results for CNS and for the nuclear
industry support extension of the testing frequencies and
demonstrate that structural integrity has been maintained.
Administrative controls govern position verification for locked
or sealed valves such that there is a very low probability that
unacceptable alignment can occur.
When the containment airlock interlock is opened during times
the interlock is required, the operator first verifies that one door
is completely shut before attempting to open the other door.
Therefore, the interlock is not challenged except during actual
testing of the interlock. Therefore, it should be sufficient to
ensure proper operation of the interlock by testing the interlock on
a 24 month interval.
There may be valves credited as containment isolation valves,
which are power operated, that do not receive a containment
isolation signal. These valves do not have an isolation time as
assumed in the accident analyses since they require operator action.
However, these valves are tested in accordance with the Inservice
Test Program as required and as such there will be no reduction in a
margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: December 6, 1999.
Description of amendment request: Changes are proposed to Technical
Specification (TS) Section 2.1.1.2 for the safety limit minimum
critical power ratio (SLMCPR). The proposed changes to TS 2.1.1.2
revise the SLMCPR values from 1.06 to 1.08 for two recirculation loop
operation, and from 1.07 to 1.09 for single recirculation loop
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:
[[Page 73094]]
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Evaluation: The basis for the Safety Limit Minimum Critical
Power Ratio (SLMCPR) is to ensure that at least 99.9% of all fuel
rods in the core avoid transition boiling if the SLMCPR limit is not
violated. The revised SLMCPR values preserve the existing margin to
transition boiling and thus the probability for fuel damage is not
increased. The determination of a revised SLMCPR Technical
Specification value does not affect the assumptions of accidents
previously evaluated; or initiate, or affect initiators, of
accidents previously evaluated. The proposed revisions to SLMCPR are
based on the use of methodology previously accepted by the NRC for
calculating SLMCPR and do not change the definition of SLMCPR. Thus,
the probability of an accident previously evaluated is not
increased.
The revised SLMCPR values do not affect the design or operation
of any system, structure, or component in the facility. No new or
different type of equipment is installed by this change. The
proposed revision does not change or alter the design assumptions
for systems, structures, or components used to mitigate the
consequences of an accident. Thus, he dose consequences of an
accident previously evaluated are not increased.
Therefore, the proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Evaluation: The SLMCPR ensures that at least 99.9% of all fuel
rods in the core avoid transition boiling if the SLMCPR limit is not
violated. The revised SLMCPR values preserve the existing margin to
transition boiling. The proposed revisions to SLMCPR are based on
the use of methodology previously accepted by the NRC for
calculating SLMCPR and do not change the definition of SLMCPR. The
proposed revision does not change the design or operation of any
system, structure, or component. No new or different type of plant
equipment is installed by this change. The proposed revision does
not involve a change to plant operation or allowable plant operating
modes. The calculational methodology used to determine a revised
SLMCPR Technical Specification value cannot initiate or create a new
or different type of accident.
Therefore, the proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed license amendment create a significant
reduction in the margin of safety?
Evaluation: The SLMCPR ensures that at least 99.9% of all fuel
rods in the core avoid transition boiling if the SLMCPR limit is not
violated. The revised SLMCPR values were calculated using a
methodology previously accepted by the NRC, and preserve the
existing margin to transition boiling and thus the margin of safety
to fuel failure. The proposed change does not involve a relaxation
of the criteria or basis used to establish safety limits, or a
relaxation in the criteria or bases for the limiting conditions for
operation. The assumptions and methodologies used in the plant
accident analysis remain unchanged. Therefore, the proposed change
does not create 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: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: December 16, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) Safety Limit Minimum Critical
Power Ratio (SLMCPR) values for two recirculation pump and single-loop
operation, delete cycle specific footnotes, update the single-loop
operation Average Planar Heat Generation rate limiting values, correct
a typographical error, and delete an obsolete reference to Siemens
fuel.
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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
GE [General Electric] has recently revised their single loop
operation (SLO) analysis review procedures to add an additional
requirement that the peak cladding temperature (PCT) during a LOCA
[loss-of-coolant accident] initiated while in SLO should be bounded
by the PCT for a LOCA initiated while in dual loop operation. This
desired result is enforced by revising the SLO MAPLHGR [maximum
average planar linear heat generation rate] ``multipliers'' found in
Technical Specification 3.11.A from the current value of 0.85 for
all fuel to values of 0.78 for GE10 fuel and 0.80 for GE11 and GE12
fuel. This change ensures that the condition that the Upper Bound
PCT does not exceed 1600 deg.F (as required by the NRC-approved
SAFER methodology for performing ECCS [emergency core cooling
system] LOCA calculations) is satisfied even if a LOCA were to occur
while operating in SLO. This change does not alter the method of
operating the plant and does not increase the probability of an
accident initiating event or transient. These limits are established
to preserve required margins.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
SLMCPR is a TS numerical value designed to ensure that
transition boiling does not occur in greater than 99.9% of all fuel
rods in the core during the limiting postulated transient. A change
in SLMCPR cannot create the possibility of any new type of accident.
SLMCPR values for the new fuel cycle are calculated using previously
transmitted methodology. Similarly, changes to the SLO MAPLHGR
multiplier values are designed to ensure that the PCT resulting from
a LOCA while operating in SLO are bounded by the PCT from a LOCA
while operating in dual loop operation. Thus, a change in these
multipliers cannot create the possibility of any new type of
accident. This multiplier update results from application of GE's
current standard methodology for this analysis.
The proposed changes result only from a specific analysis for
the Monticello core reload design and deletion of a cycle specific
reference for the values. These changes do not involve any new or
different method for operating the facility and do not involve any
facility modifications. No new initiating events or transients
result from these changes.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident, from any accident previously
evaluated.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
SLMCPR calculations are based on ensuring that greater than
99.9% of all fuel rods in the core avoid transition boiling if the
limit is not violated. Proposed SLMCPRs preserve required margin to
transition boiling and fuel damage in the event of a postulated
transient. Fuel licensing acceptance criteria for SLMCPR
calculations apply to Monticello Cycle 20 in the same manner as
applied in previous cycles. The revised SLMCPR values do not change
the method of operating the plant and have no effect on the
probability of an accident-initiating event or transient because
these limits are established to preserve required margin.
Fuel licensing acceptance criteria for SLMCPR calculations apply
to Monticello Cycle 20 in the same manner as previously applied.
SLMCPRs prepared by GE using methodology previously transmitted to
the NRC ensure that greater than 99.9% of all fuel rods in the core
will avoid transition boiling if the limit is not violated, thereby
preserving fuel cladding integrity. The operating MCPR limit is set
appropriately above the safety limit value to ensure
[[Page 73095]]
adequate margin when the cycle specific transients are evaluated.
Application of new SLO MAPLHGR multiplier values ensures that
SLO LOCA results are bounded by those for dual loop operation and
thus maintain or improve the margin of safety for LOCA analyses.
Therefore, the proposed TS changes 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: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
Portland General Electric Company, et al., Docket No. 50-344, Trojan
Nuclear Plant, Columbia County, Oregon
Date of amendment request: August 5, 1999.
Description of amendment request: The proposed amendment would add
a license condition denoting NRC approval of the Trojan Nuclear Plant
(TNP) License Termination Plan.
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 requested license amendment does not authorize additional
plant activities beyond those that already may be conducted under
the approved TNP Decommissioning Plan and the Defueled Safety
Analysis Report (DSAR). Accident analyses are included in the
approved TNP Decommissioning Plan and incorporated into the TNP
DSAR. No systems, structures, or components that could initiate or
be required to mitigate the consequences of an accident are affected
by the proposed change in any way not previously evaluated in the
approved TNP Decommissioning Plan and DSAR. Therefore, the proposed
change is administrative in nature and as such 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 requested license amendment does not authorize additional
plant activities beyond those that already may be conducted under
the approved TNP Decommissioning Plan and the DSAR. Accident
analyses are included in the approved TNP Decommissioning Plan and
incorporated into the DSAR. The proposed change does not affect
plant systems, structures, or components in any way not previously
evaluated in the approved TNP Decommissioning Plan and DSAR, and no
new or different failure modes will be created. Therefore, the
proposed change is administrative in nature and as such 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.
Approval of the TNP License Termination Plan by license
amendment is administrative in nature since the decommissioning and
fuel storage activities described in the TNP license Termination
Plan are consistent with those in the approved TNP Decommissioning
Plan and DSAR. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Leonard A. Girard, Esq., Portland General
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
NRC Section Chief: Michael T. Masnik.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: November 24, 1999.
Description of amendment request: The proposed amendment would
delete Section 4.7.D.1.e of Appendix A (Technical Specifications (TSs))
to the James A. FitzPatrick Operating License to eliminate the
surveillance requirement for partially stroking of the plant Main Steam
Isolation Valves (MSIVs) twice a week. The MSIVs will continue to be
fully stroked with a frequency that is in accordance with the In-
Service Testing (IST) Program per TS 4.7.D.1.d, which is consistent
with the Boiling-Water Reactor Standard Technical Specification and the
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code. The proposed changes include associated administrative changes to
Section 4.7.D.1.d, and to Bases Section 4.7.D of the TSs.
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 will not significant[ly] increase the
probability or consequences of any previously evaluated accidents.
This proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
This proposed change deletes the requirement to exercise the MSIVs
twice per week. The twice per week exercise involves partial closure
of each individual MSIV and subsequent reopening to the full open
position.
The safety function of the MSIV is to isolate the main steam
line in case of a steam line break, Control Rod Drop Accident or
Loss of Coolant Accident in order to limit the loss of reactor
coolant and/or the release of radioactive materials. The MSIVs
perform a safety function which mitigates the consequences of
accidents: however, an event can be initiated by the inadvertent
closure of MSIVs. Therefore, eliminating excessive operation of the
MSIVs reduces the probability of an inadvertent closure. Also, the
surveillance which is being deleted does not test the safety
function of the MSIVs. The safety function is tested during the full
stroke fast closure test. Since deleting the twice per week exercise
of the valves is not considered to have any effect on the
reliability of the MSIVs to perform there safety function, there is
no increase in the consequences of any postulated accidents.
Therefore, deleting the requirement for twice per week exercising of
the MSIVs does not significantly increase the probability or
consequences of any previously evaluated accidents.
(2) The proposed change will not create the possibility of a new
or different kind of accident.
The safety function of the MSIV is to isolate the main steam
line in case of a steam line break, Control Rod Drop Accident, or
Loss of Coolant Accident in order to limit the loss of reactor
coolant and/or the release of radioactive materials. The MSIVs
perform a safety function which mitigates the consequences of
accidents: however, an event can be initiated by the inadvertent
closure of MSIVs. The inadvertent closure of the MSIVs event has
been previously evaluated in Chapter 14 of the James A. FitzPatrick
Final Safety Evaluation Report (FSAR). The surveillance which is
being deleted does not test the safety function of the MSIVs. The
safety function is tested during the full stroke fast closure test.
Since the MSIVs perform a mitigating safety function, and the MSIV
full stroke fast closure test adequately tests the safety function,
elimination of the twice per week exercise will not create any new
or different kind of accident.
(3) The proposed change will not involve a significant reduction
in a margin of safety.
The safety function of the MSIV is not tested during the twice
per week exercise. The ability of the MSIVs to perform their safety
function is tested during the MSIV full stroke fast closure test in
accordance with the IST Program. Therefore, deletion of the
requirement does not reduce the margin of safety. The exercising of
the MSIVs was
[[Page 73096]]
originally specified in order to detect binding of the pilot valve.
The type of pilot valve that was susceptible to binding was replaced
and there is no longer any need for frequent exercising of the
MSIVs. The full closure test of the MSIVs in accordance with the IST
Program adequately demonstrates that the MSIVs and their pilot
valves are not binding and that the MSIVs will perform their safety
function. Additionally, reducing the frequency of MSIV operation
reduces the probability of inadvertent scrams and transients, and
challenges to relief valves, providing a net addition to the margin
of safety. The full stroke fast closure test is considered to be
sufficient. It is the only test required by the ASME Boiler and
Pressure Vessel Code and the BWR Standard Technical Specifications.
Therefore, eliminating the twice per week exercise of the MSIVs does
not significantly reduce any margin of safety.
The proposed change will not increase the probability or
consequences of any previously analyzed accident, introduce any new
or different kind of accident previously evaluated, or reduce
existing margin to safety. Therefore, the proposed license amendment
will not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: Alexander W. Dromerick (Acting Section Chief).
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: November 24, 1999.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to implement Filtration,
Recirculation, and Ventilation System (FRVS) and Control Room Emergency
Filtration (CREF) System charcoal filter testing requirements that are
consistent with the U. S. Nuclear Regulatory Commission requirements
delineated in Generic Letter 99-02, ``Laboratory Testing of Nuclear-
Grade Activated Charcoal.''
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.
The proposed TS change does not involve any physical changes to
plant structures, systems or components (SSC). The CREF and FRVS
systems will continue to function as designed. The CREF and FRVS
systems are designed to mitigate the consequences of an accident,
and therefore, can not contribute to the initiation of any accident.
The proposed TS surveillance requirement changes implement testing
methods that more appropriately demonstrate charcoal filter
capability and establish acceptance criteria, which ensure that Hope
Creek's licensing and design basis assumptions are met.
In addition, this proposed TS change will not increase the
probability of occurrence of a malfunction of any plant equipment
important to safety, since the manner in which the CREF and FRVS
systems are operated is not affected by these proposed changes. The
proposed surveillance requirement acceptance criteria ensure that
the FRVS and CREF safety functions will be accomplished. Therefore,
the proposed TS changes would not result in the increase of the
consequences of an accident previously evaluated, nor do they
involve an increase in the probability 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 TS changes do not involve any physical changes to
the design of any plant SSC. The design and operation of the CREF
and FRVS systems are not changed from that currently described in
Hope Creek's licensing basis. The CREF and FRVS systems will
continue to function as designed to mitigate the consequences of an
accident. Implementing the proposed charcoal filter testing methods
and acceptance criteria does not result in plant operation in a
configuration that would create a different type of malfunction to
the CREF and FRVS systems than any previously evaluated. In
addition, the proposed TS changes do not alter the conclusions
described in Hope Creek's licensing basis regarding the safety
related functions of these systems.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes contained in this submittal would implement
TS requirements that: (1) Are consistent with the requirements
delineated in Generic Letter 99-02; (2) implement testing methods
that adequately demonstrate charcoal filter capability; and (3)
establish acceptance criteria consistent with Hope Creek's licensing
basis. The ability of CREF and FRVS to perform their safety
functions is not adversely affected by these proposed changes.
Therefore, the proposed TS change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Southern California Edison Company, et al., Docket Nos. 50-206, 50-361,
and 50-362, San Onofre Nuclear Generating Station, Units 1, 2, and 3,
San Diego County, California
Date of amendment requests: December 2, 1999 (Unit 1--PCN 267,
Units 2 and 3--PCN 506).
Description of amendment requests: This amendment application is a
request to revise the Unit 1 Technical Specifications Section D6,
Administrative Controls, to be consistent with the San Onofre Units 2
and 3 Technical Specification Section 5.0, Administrative Controls, and
incorporate changes related to certified fuel handlers and 10 CFR
50.54(x), administrative control of working hours and working hour
deviation approvals, position titles and responsibilities and
organizational description reference, qualifications for a multi-
discipline supervisor, quality assurance program control of review and
audit and record retention procedures, high radiation area controls,
description of the plant configuration for environmental protection,
and environmental protection related document reporting.
This amendment application also requests to revise the Unit 2 and
Unit 3 Technical Specifications, Section 5.0, Administrative Controls,
to incorporate changes related to the operating organization, working
hours deviation approvals, qualifications for a multi-discipline
supervisor, the schedule for submitting Technical Specification Bases
changes, reference to American Society of Mechanical Engineers (ASME)
code class components, steam generator inspection reporting, Core
Operating Limits Report references, high radiation area controls,
offsite dose calculation manual change control reference, and
environmental protection related document reporting.
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?
[[Page 73097]]
No. This proposed change is to revise the administrative
controls section of the San Onofre Units 1, 2 and 3 technical
specifications. To the extent practicable, the San Onofre Unit 1
technical specification Section D6, Administrative Controls, is made
consistent with the San Onofre Units 2 and 3 technical specification
Section 5.0, Administrative Controls. This change allows the
handling of key administrative controls to be consistent on site.
Certain position titles have been revised, and the cognizant Vice
President has been included as an approver of deviations from the
work hours and reviewer of overtime hours. The Vice President--
Business and Financial Services is identified to be responsible for
Unit 1 decommissioning. The specification allowing the certified
fuel handlers to implement 10 CFR 50.54(x) is removed since this is
now included in the regulations. The qualification requirements for
a multi-discipline supervisor consistent with the American National
Standards Institute [ANSI] standard have been added to the staff
qualifications section. The schedule for submitting technical
specification Bases changes is revised to be consistent with the NRC
approved exemption to 10 CFR 50.71(e) for submitting Updated Final
Safety Analysis Report (UFSAR) updates. A reference to Class 1, 2,
and 3 ASME code components is removed from the technical
specifications and maintained in the Licensee Controlled
Specifications (LCS) and the inservice inspection and testing
program. The Units 2 and 3 steam generator inspection reporting
requirements are revised to refer to the technical specification
requirement. The Core Operating Limits Report (COLR) section is
revised to include references to 2 topical reports related to the
reload analysis technology transfer and the NRC's evaluation of the
technology transfer. The sections on high radiation are revised to
be consistent with Regulatory Guide 8.38 which provides an
acceptable method for controlling access to high radiation areas.
The environmental protection section of the San Onofre Unit 1
technical specifications is revised to reflect the current status of
the discharge system. The environmental protection sections for Unit
1 and Units 2/3 are further revised by including a 30 day timeframe
for providing the NRC copies of reports related to unusual or
important environmental events and deleting the requirement to
provide the NRC copies of proposed changes and renewal applications
for NPDES permits.
All of these changes are being made to provide consistency and
flexibility in the handling of site programs, and update and clarify
the administrative controls. There are no equipment changes or
modifications to the plant associated with these changes that would
affect the probability or consequences of accidents at all three
units.
Therefore, this change does not affect the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different type of accident
from any accident previously evaluated?
No. This proposed change is to revise the administrative
controls sections of the San Onofre Units 1, 2, and 3 technical
specifications. The changes provide consistency and flexibility in
the handling of site programs, and update and clarify the
administrative controls. There is no administrative change being
made that could create a new or different accident at any of the
three units and there is no plant or equipment modification
associated with this change.
Therefore, this change does not create the possibility of a new
or different type of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety?
No. This change revises the administrative controls sections of
the San Onofre Units 1, 2, and 3 technical specifications. The
changes provide consistency and flexibility in the handling of site
programs, and update and clarify the administrative controls. There
is no change to plant equipment associated with this change. This
change does not affect any margin of safety.
Therefore, this change does not involve a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chiefs: Michael Masnik (Unit 1); Stephen Dembek (Units
2 and 3).
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: November 24, 1999 (PCN-274).
Description of amendment requests: The licensee proposes to revise
Technical Specification (TS) 3.3.11, ``Post Accident Monitoring
Instrumentation (PAMI).'' Specifically, the proposed change would
extend the PAMI channel calibration surveillance frequency from 18
months to 24 months to accommodate a 24-month fuel cycle. All PAMI
instruments would then be on a 24-month calibration interval, which
removes the need for Surveillance Requirement (SR) 3.3.11.5. Therefore,
the licensee also proposes to delete SR 3.3.11.5.
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?
Response: No.
The proposed license amendment[s] to extend the calibration
surveillance frequency of Post Accident Monitoring Instrumentation
(PAMI) instrumentation [are] being made to support plant operation
with a 24-month fuel cycle.
Increasing the calibration intervals for PAMI instrumentation to
30 months [24 months plus the 25% surveillance interval extension
allowed by SR 3.0.2] does not affect the initiation or probability
of any previously analyzed accident. Increasing the calibration
interval will not affect the integrity of any of the principal
barriers against radiation release (fuel cladding, reactor vessel,
and containment building). The ability of the plant to mitigate the
consequences of any previously analyzed accidents is not adversely
affected.
PAMI instrumentation provides to the operators both qualitative
and quantitative information used in accident mitigation and for the
safe shutdown of the plant. Instrumentation which provides
qualitative information is unaffected by a change in instrument
accuracy induced by drift due to the increased surveillance interval
because no explicit value is required by the Emergency Operating
Instructions (EOIs). Instrumentation that provides quantitative
information (i.e., decision points) in the EOIs have been evaluated.
This evaluation resulted in no changes to any operating
instructions. This evaluation of the proposed change to the
surveillance interval demonstrates that licensing basis safety
analyses acceptance criteria and San Onofre Nuclear Generating
Station (SONGS) Units 2 and 3 EOI criteria will continue to be met.
The proposed new surveillance frequency for these instrument
channels was evaluated using the guidance of Generic Letter 91-04.
The basis for the change includes a quantitative evaluation of
instrument drift for PAMI instrumentation providing quantitative
information to the EOIs. Also, loop accuracy/setpoint calculations
for these instruments were updated to accommodate the extended
surveillance period. Analyses and evaluations completed to assess
the proposed increase in the surveillance interval demonstrate that
the effectiveness of these instruments in fulfilling their
respective functions is maintained. Technical Specifications Channel
Checks and Channel Functional Checks for the subject channels, will
continue to be performed to provide assurance of instrument channel
OPERABILITY.
Therefore, the proposed amendment[s do] not involve a
significant increase in the probability or consequences of any
previously analyzed accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated?
Response: No.
The increased calibration surveillance interval for PAMI
instrumentation is justified based on evaluation of past equipment
[[Page 73098]]
performance and does not require any plant hardware changes or
changes in normal system operation. Changing the calibration
interval for this instrumentation has no means of creating the
possibility of a new or different kind of accident. There are no new
decision points or operator responses required to support existing
accident mitigation strategies.
Therefore, there are no new failure modes introduced as a result
of extending these surveillance intervals, and the proposed
amendment[s do] not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety?
Response: No.
The proposed change to the calibration surveillance interval was
evaluated using the criteria of 95% probability/95% confidence level
for process sensor drift.
PAMI instrumentation are used to provide indication following
certain hypothetical accident conditions and are used in EOIs for
trending and to initiate operator action at certain decision points.
Instrument uncertainty calculations have been updated for PAMI
instrumentation used for EOI decision points as appropriate. Updated
calculations show that the total loop uncertainty for PAMI evaluated
either decreased or remained the same. These updated calculations
demonstrate that applicable accuracy requirements for SONGS 2 and 3
are satisfied with the proposed new surveillance intervals.
Changing the calibration interval for these channels does not
affect the margin of safety for previously analyzed accidents.
Therefore, the proposed amendment[s do] not involve a significant
reduction in a margin of safety.
Based on the responses to these three criteria, Southern
California Edison (SCE) has concluded that the proposed amendment[s
involve] no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: December 13, 1999 (PCN-507).
Description of amendment requests: San Onofre Nuclear Generating
Station (SONGS) Units 2 and 3 are currently licensed for operation for
40 years commencing with issuance of their construction permits. The
licensee proposes to amend the SONGS Units 2 and 3 operating licenses
to revise the expiration dates of these licenses to 40 years from the
date of issuance of the operating licenses. Thus, these amendment
applications request that the SONGS Unit 2 operating license expiration
date be changed from October 18, 2013, to February 16, 2022, and the
SONGS Unit 3 operating license expiration date be changed from October
18, 2013, to November 15, 2022.
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?
Response: The proposed change does not involve any changes to
the design or operation of the San Onofre Nuclear Generating Station
(SONGS) 2 and 3 which may affect the probability or consequences of
an accident evaluated in the Updated Final Safety Analysis Report
(UFSAR). SONGS 2 and 3 were designed and constructed on the basis of
a forty (40) year life. The accidents analyzed in the UFSAR were
postulated on the basis of a 40 year life. No changes will be made
that could alter the design, construction, or postulated scenarios
regarding accident initiation and/or response. Existing
surveillance, inspection, testing and maintenance practices and
procedures ensure that degradation in plant equipment, structures,
and components will be identified and corrected throughout the life
of the plant. The effect of aging of electrical equipment, in
accordance with 10 CFR50.49, has been incorporated into the plant
maintenance and surveillance procedures. Therefore, the probability
or consequences of a postulated accident previously evaluated in the
UFSAR are not increased as a result of the proposed change.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: The proposed change does not involve any changes to
the physical structures, components, or systems of SONGS 2 and 3.
Existing surveillance, inspection, testing, and maintenance
practices and procedures will assure full operability for the
plant's design lifetime of 40 years. Continued operation of SONGS 2
and 3 in accordance with these approved procedures and practices
will not create a new or different kind of accident.
(3) Involve a significant reduction in a margin of safety?
Response: There are no changes in the design, design basis, or
operation of SONGS 2 and 3 associated with the proposed change.
Existing surveillance, inspection, testing, and maintenance
practices and procedures provide assurance that any degradation of
equipment, structures, or components will be identified and
corrected throughout the lifetime of the plant. These measures
together with the continued operation of SONGS 2 and 3 in accordance
with the Technical Specifications assure an adequate margin of
safety is preserved on a continuous basis. Therefore, the proposed
change does 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Stephen Dembek.
Southern Nuclear Operating Company, Inc., 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 30, 1999.
Description of amendment request: The proposed amendments would
change Technical Specification Surveillance Requirement 3.8.1.12 to
remove the restriction which prevents performance of the diesel
generator 24-hour run while operating in either Mode 1 or Mode 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or the consequences of a previously evaluated
event for the following reasons:
The primary function of the diesel generators is to supply
emergency power to the safety-related equipment necessary to safely
shut down the plant in case of a design basis event, such as a loss
of coolant accident (LOCA) concurrent with a loss of offsite power
(LOSP). The diesels are not designed to prevent such an event.
Accordingly, the probability of a LOCA/LOSP event is not increased
by allowing the performance of the 24-hour run with the reactor
operating.
It is possible that, with a diesel generator connected to its
bus, an electrical disturbance will travel through the system and
affect the other busses. This is most likely to happen when
initially connecting the diesel to the bus. However, the
surveillance procedures require that diesel generator output voltage
be synchronized with the bus prior to the diesel output breaker
being closed in, thus reducing the chance of an electrical
distribution system disturbance.
[[Page 73099]]
If a LOCA occurred concurrent with an LOSP while a diesel
generator is connected to the bus in its 24-hour run, the diesel
logic automatically realigns itself to the Standby mode of
operation, allowing the diesel to supply power to the emergency bus.
A Technical Specifications surveillance requirement tests this
feature. Also, the proposed specification prevents the test from
being performed unless the other two diesel generators are operable;
this includes suspending the surveillance if one of the other
available diesels becomes inoperable during the actual test. This
restriction will ensure that two diesels are available to safely
shut down the plant if necessary.
Additionally, this amendment request does not affect any other
system or piece of equipment necessary to prevent or mitigate the
consequences of previously evaluated events. As a result, the
consequences of a LOCA/LOSP event are not increased.
2. The proposed changes do not create the possibility of an
accident of a new or different kind from any previously evaluated
based upon the following:
This proposed modification to SR 3.8.1.12 does not introduce any
new modes of operation or testing. In fact, each diesel generator is
already connected to its respective bus during operation to satisfy
SR 3.8.1.2, the monthly test. In the monthly test, the diesel is run
loaded for 1 hour, connected to the grid, with the unit in
operation. Therefore, allowing the 24 hour test to be performed for
the diesels introduces nothing new with respect to diesel testing,
and as a result, the possibility of a new type of event is not
created.
3. The change does not significantly reduce the margin of safety
for the following reasons:
The probability of an electrical disturbance affecting plant
operation while connecting the diesel to the bus is minimized by the
fact that the diesel's output voltage and phase angle are
synchronized with those of the grid prior to being tied to the
emergency bus. Based on engineering judgement, with the diesel
synchronized and running connected to the grid, the likelihood of an
electrical disturbance being transferred through the system and
causing a plant transient is very small. Furthermore, since only one
diesel will be tied to the bus in either Mode 1 or Mode 2, neither
of the other two diesel generators will be affected by the
disturbance.
If a LOCA/LOSP occurred during the 24-hour run, the diesel
generator's auto-logic would take the diesel out of the test mode.
This feature is tested once per 18 months per Technical
Specifications. With the diesel no longer in test, it would be free
to once again tie itself to the bus. Additionally, only one diesel
will be tied to the line during a 24-run performed with the reactor
operating, with other diesel generators available to supply power to
their respective emergency busses. This ensures two diesels are
available to shut down the plant and maintain it in a safe
condition.
Other precautions will also be placed into plant procedures;
specifically, the 24-hour run will not be performed on line during
periods of severe weather or during grid instabilities.
For the above reasons, the proposed Technical Specifications
change will 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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Richard L. Emch, Jr.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: November 18, 1999.
Description of amendment request: The proposed amendments would
revise technical specification surveillance requirements 4.7.7, 4.7.8,
and 4.9.12, on the control room makeup and cleanup filtration system
and the fuel handling building exhaust air system, from a requirement
that laboratory analysis of charcoal filter samples meets the
laboratory testing criteria of Regulatory Position C.6.a of Regulatory
Guide 1.52, ``Design, Testing, and Maintenance Criteria for
Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air
Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power
Plants,'' Revision 2, March 1978, to a requirement that the analysis
meets the laboratory testing criteria of American Society for Testing
and Materials ASTM D3803-1989, ``Standard Test Method for Nuclear-Grade
Activated Carbon.''
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:
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change revises the test protocol for Engineered
Safety Feature charcoal filters from ASTM D3803-1979 to ASTM D3803-
1989. The change in protocol is a conservative change in that the
revised test conditions will more accurately reflect the
functionality of the charcoal filters under accident conditions.
There is no change in plant configuration or components. The tests
are conducted under laboratory conditions, so that change in
protocol has no effect on plant operation. There is no change in how
samples are taken to be used in analyses.
Based on the above, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change revises the test protocol for Engineered
Safety Feature charcoal filters from ASTM D3803-1979 to ASTM D3803-
1989. The change in protocol is a conservative change in that the
revised test conditions will more accurately reflect the
functionality of the charcoal filters under accident conditions.
There is no change in plant configuration or components. The tests
are conducted under laboratory conditions, so that change in
protocol has no effect on plant operation. There is no change in how
samples are taken to be used in analyses.
Based on the above, this change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a significant reduction in
a margin of safety.
The proposed change revises the test protocol for Engineered
Safety Feature charcoal filters from ASTM D3803-1979 to ASTM D3803-
1989. The change in protocol is a conservative change in that the
revised test conditions will more accurately reflect the
functionality of the charcoal filters under accident conditions.
There is no change in plant configuration or components. The tests
are conducted under laboratory conditions, so that change in
protocol has no effect on plant operation. There is no change in how
samples are taken to be used in analyses.
Based on the above, the margin of safety is not significantly
reduced by this change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: December 6, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification Definition 1.9, ``Core Alterations,'' to
explicitly define core alterations as the movement of any fuel,
sources, or reactivity control components within the reactor vessel
with the vessel head removed and fuel in the vessel.
[[Page 73100]]
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 or consequences of an accident previously evaluated?
The proposed change does not involve an increase in the
probability or consequences of an accident previously evaluated. The
proposed change does not involve any physical changes to the
facility. The change to the definition of core alterations is
consistent with that used in NUREG-1431, Revision 1, ``Improved
Standard Technical Specifications for Westinghouse Plants.'' The
proposed revision to the definition of core alterations will not
affect the Technical Specifications Section 3/4.9, ``Refueling
Operations'', requirements which ensure the core remains
subcritical, nor will any Limiting Condition for Operation required
for core alterations or the movement of fuel be changed. The
proposed change will not affect any safety margin or safety limit
applicable to the facility. Therefore, the proposed change does not
involve an increase in the probability or consequences of any
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 change does not affect any previously evaluated
accident scenario, nor does it create any new accident scenarios.
The proposed change is a clarifying revision to the definition of
core alterations only, and will not alter any of the currently
approved refueling operation activities, nor will it create any new
refueling operation activities.
Since the proposed change does not impact operation of the
facility as presently approved, no possibility exists for a new or
different kind of accident from those previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
South Texas Project Technical Specification 3/4.9.1, ``Boron
Concentration'', ensures that the reactor will remain subcritical
(Keff 0.95) during core alterations and that
uniform boron concentration is maintained for reactivity control in
the water volume having direct access with the reactor vessel. The
proposed change in the definition of core alterations will allow
``non-reactive'' components, such as cameras, lights, tools, movable
incore detector thimbles, etc., to be moved or manipulated in the
vessel, with fuel in the vessel and the vessel head removed, without
constituting a core alteration. This is acceptable because these
types of components will have negligible effect on core reactivity,
and will not affect reactor coolant system boron concentration.
Therefore, operations using these types of components will not
adversely affect Keff or the shutdown margin.
Additionally, reactor subcriticality status is continuously
monitored in the control room during Operating Mode 6, as specified
in Specification 3/4.9.2, ``Instrumentation''. Thus, there will be
no reduction in a margin of safety resulting from the proposed
change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: October 14, 1999 (TS 99-12).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Operating Licenses DPR-77 (Unit 1) and DPR-79
(Unit 2) by revising the Technical Specification (TS) surveillance
requirements for steam generator tube integrity by incorporating an
alternate repair criteria for axial primary water stress corrosion
cracking at dented tube support plate intersections.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), 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.
Operation of Sequoyah Units 1 and 2, in accordance with the
proposed license amendment, does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Examination of crack morphology for primary water stress
corrosion cracking (PWSCC) at dented intersections has been found to
show one or two microcracks well aligned with only a few uncorroded
ligaments and little or no other inside diameter axial cracking at
the intersection. This relatively simple morphology is conducive to
obtaining good accuracy in Non-destructive Examination (NDE) sizing
of these indications. Accordingly, alternate repair criteria is
established based on crack length and average and maximum depth
within the thickness of the tube support plate (TSP) or limited
extension outside the thickness of the TSP.
The application of the alternate repair criteria (ARC) requires
a condition monitoring assessment. If all indications satisfy the
structural limits with regard to bounding lengths and average
depths, the condition monitoring burst pressure requirements are
satisfied.
In addition, an operational assessment is performed to determine
the length/depth repair bases. The crack profiles are projected to
the end of the operating cycle for comparison with acceptance limits
(i.e., length limit and average depth limit). Burst pressures are
calculated from the depth profiles by searching the total crack
length for the partial length that results in the lowest burst
pressure. Because the burst pressure can be lower than that for the
longest acceptable crack length at its average depth, a fixed repair
limit is not established. The repair bases is obtained by projecting
the crack profile to the end of the next operating cycle and
determining if the burst pressure for the projected profile meets
the burst pressure margin requirements defined by [Westinghouse
Topical Report] WCAP-15128, Revision 1, dated August 1999. If the
projected end-of-cycle (EOC) burst margin requirements are
satisfied, the indication is left in service. Thus, the repair limit
relative to length and average depth assures that the operational
assessment requirements are satisfied.
Crack length limits are established in the WCAP to assure that
crack extension and growth outside of the TSP provides adequate
margin against burst for the free-span crack (i.e., 3DPNO
burst capability is maintained) in addition to the total crack
length. A repair limit is also established in the WCAP for maximum
depth to provide a high confidence that the indication will not
progress through the wall at the end of an operating cycle.
Based on the above, the proposed amendment does not result in
any increase in the probability or consequences of an accident
previously evaluated within the Sequoyah FSAR [Final Safety Analysis
Report].
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Implementation of the proposed S/G [steam generator] tube ARC
does not introduce any significant changes to the plant design
basis. A single or multiple tube rupture event would not be expected
in a S/G in which the plugging criteria has been applied. Both
condition monitoring and operational assessments are completed as
part of the implementation of ARC to determine that structural and
leakage margin exists prior to returning S/Gs to service following
inspections. If the condition monitoring requirements are not
satisfied for burst or leakage, the causal factors for EOC
indications exceeding the expected values will be evaluated. The
methodology and application of this ARC will continue to ensure that
tube integrity is maintained during all plant conditions consistent
with the requirements of draft RG [Regulatory Guide] 1.121 and
Revision 1 of RG 1.83.
A S/G tube rupture event is one of a number of design basis
accidents that are analyzed as part of a plant's licensing basis. In
the analysis of a S/G tube rupture event, a bounding primary-to-
secondary leakage rate equal to the operational leakage limits in
the TSs, plus the leak rate associated with the double ended rupture
of a single tube, is
[[Page 73101]]
assumed. For other design basis accidents such as a main steam line
break and loss of alternating current power, the tubes are assumed
to retain their structural integrity and exhibit primary-to-
secondary leakage within the limits assumed in Final Safety Analysis
Report (FSAR) accident analyses. The proposed ARC does not result in
an accident leakage rate in excess of that assumed or calculated in
SQN's current accident analyses.
Even under severe accident conditions, the potential for
significant leakage would be expected to be small and not
significantly different than for other degradation mechanisms
repaired to 40 percent depth limits. It is concluded that
application of the proposed ARC for PWSCC at dented TSP locations
results in a negligible difference from current 40-percent repair
limits.
TVA continues to implement a maximum operating condition leak
rate limit of 150 gallons per day (0.1 gallons per minute) per S/G
to preclude the potential for excessive leakage during all plant
conditions.
The possibility of a new or different kind of accident from any
previously evaluated is not created because S/G tube integrity is
maintained by inservice inspection and effective primary-to-
secondary leakage monitoring.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
Tube repair limits provide reasonable assurance that tubes
accepted for continued service without plugging or repair will
exhibit adequate tube structural and leakage integrity during
subsequent plant operation. The implementation of the proposed ARC
is demonstrated to maintain S/G tube integrity consistent with the
criteria of draft NRC Regulatory Guide 1.121. The guidelines of RG
1.121 describe a method acceptable to the NRC staff for meeting
General Design Criteria (GDC) 2, 4, 14, 15, 31, and 32 by ensuring
the probability or the consequences of S/G tube rupture remain
within acceptable limits. This is accomplished by determining the
limiting conditions of degradation of S/G tubing, for which tubes
with unacceptable cracking should be removed from service.
Upon implementation of the proposed ARC, even under the worst-
case conditions, the occurrence of PWSCC at the tube support plate
elevations is not expected to lead to a S/G rupture event during
normal or faulted plant conditions. All tubes are shown to retain
the margins of safety against burst consistent with the safety
factor margins implicit in the stress limit criteria of Section III
of the American Society of Mechanical Engineers [ASME] Code, for all
service loading conditions. In addition, all tubes have been shown
to retain a margin of safety against gross failure or burst
consistent with the stress limits of [Paragraph] NB-3225 of Section
III of the ASME Code under postulated accident conditions concurrent
with a safe shutdown earthquake.
In addressing the combined effects of loss-of-coolant accident
plus safe shutdown earthquake on the S/G component (as required by
GDC 2), it has been determined that tube collapse will not occur in
the Sequoyah S/Gs. This analysis is discussed in WCAP 13990, dated
May 1994. No tubes are excluded from the application of the proposed
ARC.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin with respect to the plant safety analyses as defined in the
FSAR or TSs.
The NRC 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.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: November 8, 1999.
Brief description of amendments: The proposed amendments would
change Technical Specification 5.5.11, ``Ventilation Filter Testing
Program (VFTP)'' to include the requirement for laboratory testing of
Engineered Safety Feature (ESF) Ventilation System charcoal samples per
American Society for Testing and Materials (ASTM) D3803-1989 and the
application of a safety factor of 2.0 to the charcoal filter efficiency
assumed in the plant design-basis dose analyses.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Do the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed changes only involve the laboratory testing
methodology performed on activated charcoal to help determine
whether the charcoal in the filtration units can remain in place or
[if it] require[s] replacement.
Generic Letter 99-02 intends to standardize the way nuclear-
grade activated charcoal is tested throughout the industry in order
to provide conservative filtration results as well as uniform and
repeatable tests. The purpose is to ensure the filtration systems
protect the Operators in the Control Room (GDC [General Design
Criterion] 19) as well as the public (10CFR100), in the event of a
radiological accident scenario.
The charcoal adsorber sample laboratory testing per ASTM D3803-
1989 is more stringent than the current testing practice and more
accurately demonstrates the required performance of the adsorbers
following a design ba[s]is LOCA [loss of coolant accident]. No
Licensing Basis Accidents or mitigation capability will be affected
by incorporation of these changes.
Therefore, this change will not result in a significant increase
in the probability or consequences of an accident previously
evaluated.
(2) Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Plant procedures are only altered to the extent that the revised
specification will allow different reference standards for testing
activated charcoal. These changes ensure continued support of the
safety related ESF filtration equipment and do not affect their
failure or failure modes.
Therefore, this change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
(3) Do the proposed changes involve a significant reduction in a
margin of safety?
None of the changes being proposed alter the environmental
conditions maintained in the areas supported by the ESF filtration
systems during normal operations and following an accident. Also
these changes will not cause an increase in radiological releases
through the Primary Plant Ventilation Exhaust System. As a result,
the margin of safety for these functions remains the same.
Therefore, this change does not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: December 3, 1999 (ULNRC-04158).
Description of amendment request: The proposed amendment requested
changes to Section 5.6.6, ``Reactor Coolant System (RCS) Pressure and
Temperature Limits Report (PTLR),'' of the improved Technical
Specifications (ITS) that were issued on May 28, 1999, in Amendment No.
133. The current Technical Specifications (CTS) remain in effect until
the ITS are implemented on or before April 30, 2000. The proposed
changes to the ITS would approve the use of the PTLR by the licensee to
make changes to the plant pressure temperature limits and low
temperature overpressure protection
[[Page 73102]]
limits without prior NRC staff approval in accordance with Generic
Letter 96-03, ``Relocation of the Pressure Temperature Limit Curves and
Low Temperature Overpressure Protection System Limits,'' dated January
31, 1996. The proposed changes are: (1) Add the word criticality to ITS
Subsection 5.6.6.a as one of the reactor conditions for which RCS
pressure and temperature limits will be determined, (2) add the phrase
``and COMS PORV,'' where COMS PORV stands for cold overpressure
mitigation system power operated relief valve, to the the introductory
paragraph of ITS subsection 5.6.6.b to show that the analytical methods
listed in the subsection are also for the COMS PORV, and (3) replace
the two documents listed in ITS subsection 5.6.6.b by the reference to
the future NRC letter that approves the use of the PTLR and the
Westinghouse Topical Report, WCAP-14040-NP-A, Revision 2, ``Methodology
Used to Develop Cold Overpressure Mitigating System Setpoints and RCS
Heatup and Cooldown Limit Curves,'' dated January 1996, that provides
the methodology that will be used by the licensee in using the PTLR
report. The current plant pressure temperature limits and low
temperature overpressure protection limits are in the CTS and were
approved in Amendment No. 124, which was issued April 2, 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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change submits the PTLR, which contains the
relocated CTS heatup and cooldown, and COMS PORV limits and the
methodology used to calculate them, and the added references into
ITS 5.6.6. The proposed change is administrative in nature since it
is a movement of information from the CTS to a licensee controlled
document, and has prior NRC staff approval. The PTLR contains the
limit curves and the ITS requires more restrictive actions to be
taken when the limiting conditions for operation are not met than is
currently required by the CTS. The heatup and cooldown, and COMS
PORV limits within the PTLR will be implemented and controlled per
Callaway Plant programs and procedures and changes to the PTLR will
be performed per requirements of 10 CFR 50.59 to ensure that change
to these limits in the future 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.
As stated earlier, the movement of the heatup and cooldown, and
COMS PORV limits from the CTS to the PTLR has no influence or
impact, nor does it contribute in any way to the probability or
consequences of an accident. No safety-related equipment, safety
function, or plant operations will be altered as a result of this
proposed change. The proposed change is administrative in nature
since it is a movement of requirements from the CTS to a licensee
controlled document, the PTLR, and the change adds references into
the ITS incorporating the licensee controlled document. Therefore,
the possibility of a new or different kind of accident from any
accident previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not affect the acceptance criteria for
an analyzed event. The margin of safety presently provided by the
CTS remains unchanged. There will be no effect on the manner in
which safety limits or limiting safety system settings are
determined nor will there be any effect on those plant systems
necessary to assure the accomplishment of protective functions.
Therefore, the proposed change is administrative in nature and does
not impact the operation of Callaway Plant in a manner that involves
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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Section Chief: Stephen Dembek.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: November 5, 1999, as supplemented on
December 3, 1999.
Description of amendment request: This proposed change revises the
applicability for the reactor power distribution limits and the Average
Power Range Monitor (APRM) gain adjustments. The applicability is
proposed to be revised to operation at 25% Rated Thermal
Power (RTP).
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 operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change does not involve an increase in the
probability or consequences of an accident previously evaluated
because the revisions standardize and make consistent the
applicability and actions for the reactor power distribution limits
in the current Technical Specifications. Since reactor operation
with these revised Specifications is fundamentally unchanged, no
design or analytical acceptance criteria will be exceeded. As such,
this change does not impact initiators of analyzed events or assumed
mitigation of accident or transient events. The structural and
functional integrity of plant systems is unaffected. Therefore, the
proposed change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not affect any parameters or conditions
that could contribute to the initiation of any accident. No new
accident modes are created. No safety-related equipment or safety
functions are altered as a result of these changes. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
At thermal power levels < 25% RTP, the reactor is operating with
substantial margin to the reactor power distribution limits [and
this margin is unchanged]. The proposed change does not impact
operation at power levels 25% RTP and has no effect on
any safety analysis assumption or initial condition. Thus, the
margin of safety required for safety analyses [is] maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: James W. Clifford.
[[Page 73103]]
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant (PBNP), Units 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: November 15, 1999 (TSCR 202).
Description of amendment request: The proposed amendments would
change the Technical Specifications (TSs) in order to extend the
required frequency of the control rod exercise test (TS 15.4.1, Table
15.4.1-2, Item 10) from the current frequency of every 2 weeks to
quarterly.
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 Point Beach Nuclear Plant in accordance with
the proposed amendments will not result in a significant increase in
the probability or consequences of an accident previously evaluated.
Relaxing the frequency of performance for a surveillance does
not result in any hardware changes, nor does it significantly
increase the probability of occurrence for initiation of any
analyzed events since the function of the equipment has remained
unchanged. The proposed frequency has been determined to be adequate
based on industry operating data as supported by the conclusions
reached in NUREG 1366 and NRC GL [Generic Letter] 93-05.
Surveillance tests are intended to provide assurance of
continued component operability. The frequency of performance of a
surveillance does not significantly increase the consequences of an
accident, as a change in frequency does not change the response of
the equipment in performing its specified function (i.e. the overall
functional capabilities of the rod control system will not be
modified). Increasing the interval of control rod exercise testing
will reduce the possibility of inadvertent testing related [to]
reactor trips and dropped rods, and resulting in fewer challenges to
safety systems and resultant plant transients.
This change does not involve a significant increase in the
consequences of an accident or event previously evaluated because
the source term, containment isolation or radiological releases are
not being changed by the proposed revision. Existing system and
component redundancy and operation is not being changed by the
proposed change. The assumptions used in evaluating the radiological
consequences in the PBNP Final Safety Analysis Report are not
invalidated. Therefore, this change does not affect the consequences
of previously evaluated accidents.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This change does not introduce nor increase the number of
failure mechanisms of a new or different type of accident than those
previously evaluated since there are no physical changes being made
to the facility. The design and design basis of the facility remain
unchanged. The plant safety analyses remain unchanged. All equipment
important to safety will continue to operate as designed. Component
integrity is not challenged. The changes do not result in any event
previously deemed incredible being made credible. The changes do not
result in more adverse conditions nor result in any increase in
challenges to safety systems. Therefore, operation of the Point
Beach Nuclear Plant in accordance with the proposed amendment will
not create the possibility of a new or different type of accident
from any accident previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not involve a significant reduction in
a margin of safety.
The proposed change does not involve a significant reduction in
the margin of safety because existing component redundancy is not
being changed by this proposed change. There are no changes to
initial conditions contributing to accident severity or
consequences. The proposed surveillance frequency, as supported by
past test results, continues to provide the required assurance of
operability, such that safety margins established through the design
and facility license, including the Technical Specifications, remain
unchanged. Therefore, there are no significant reductions in a
margin of safety introduced by this proposed amendment.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Claudia M. Craig.
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 notice was previously published as a separate
individual notice. The notice content was the same as above. It was
published as an individual notice either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. It is 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.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: October 20, 1999.
Brief description of amendment: The amendment changed the footnote
to the Improved Technical Specifications associated with the Design
Features Fuel Storage Specification 4.3.1.1.b which required that 2300
ppm boron be maintained in the Spent Fuel Pool.
Date of publication of individual notice in Federal Register:
November 19, 1999 (64 FR 63346).
Expiration date of individual notice: December 20, 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
[[Page 73104]]
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).
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: September 14, 1999.
Brief description of amendments: The amendments approve the
administrative changes to PVNGS TS 5.5.2, Primary Coolant Sources
Outside Containment, to delete the references to the post-accident
sampling return piping of the radioactive waste gas system and the
liquid radwaste system, and TS 5.6.2, Annual Radiological Environmental
Operating Report, to delete the administrative requirement to include
in the report certain TLD [thermoluminescence dosimeter] results that
are no longer available.
Date of issuance: November 24, 1999.
Effective date: November 24, 1999, to be implemented within 60
days.
Amendment Nos.: Unit 1--122, Unit 2--121, Unit 3--121.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 20, 1999 (64 FR
56528).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 24, 1999.
No significant hazards consideration comments received: No.
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 application for amendment: October 21, 1999.
Brief description of amendment: This amendment revises Technical
Specifications (TS) for the Shearon Harris Nuclear Power Plant by
implementing selected improvements described in NRC Generic Letter (GL)
93-05, ``Line-Item Technical Specifications To Reduce Surveillance
Requirements For Testing During Power Operation,'' dated September 27,
1993.
Date of issuance: December 17, 1999.
Effective date: December 17, 1999.
Amendment No: 93.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 17, 1999 (64
FR 62705).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 17, 1999.
No significant hazards consideration comments received: No.
CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz
Mill, Pennsylvania
Date of application for amendment: September 15, 1999, as
supplemented on October 4, 1999.
Brief description of amendment: This amendment changes the
decommissioning Technical Specifications dealing with controls for
ingress, egress, and equipment removal from containment.
Date of issuance: December 7, 1999.
Effective Date: December 7, 1999.
Amendment No: 11.
Facility License No. TR-2: This amendment changes the
decommissioning Technical Specifications.
Date of initial notice in Federal Register: November 3, 1999 (64 FR
59798).
The Commission has issued a Safety Evaluation for this amendment
dated December 7, 1999.
No significant hazards consideration comments received: No.
Consolidated Edison Company of New York, Inc., Docket No. 50-003,
Indian Point Nuclear Generating Station, Unit 1, Buchanan, New York
Date of application for amendment: July 20, 1999.
Brief description of amendment: The amendment would revise the
Technical Specifications to change the senior license requirements for
the Operations Manager.
Date of issuance: December 15, 1999.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment No: 46.
Facility Operating License No. DPR-5: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 2, 1999 (64
FR 49027).
The July 20, 1999, letter providing clarifying information that did
not change the scope of the original application and proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated December 15, 1999.
No significant hazards consideration comments received: No.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania
Date of application for amendments: May 5, 1999, as supplemented
June 22 and July 30, 1999.
Brief description of amendments: These amendments conform the
licenses to reflect the transfer of Operating Licenses Nos. DPR-66 and
NPF-73 for the Beaver Valley Power Station Unit Nos. 1 and 2, to the
extent held by Duquesne Light Company (DLC) to the Pennsylvania Power
Company, and the operating authority under the licenses from DLC to
FirstEnergy Nuclear Operating Company as previously approved by an
Order dated September 30, 1999.
Date of issuance: December 3, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 226 and 104.
Facility Operating License Nos. DPR-66 and NPF-73: These amendments
revised the Operating Licenses.
Date of initial notice in Federal Register: June 14, 1999 (64 FR
31880).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 30, 1999. The June 22 and July
30, 1999, supplements were within the scope of the initial application
as originally noticed.
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: September 14, 1999.
Brief description of amendment: This amendment eliminates License
Condition 2.C.10 of the Operating License regarding controls over the
containment air locks during plant outages and modifies License
Condition 2.F of the Operating License regarding reporting requirements
for violations of the Technical Specifications and the Environmental
Protection Plan.
Date of issuance: December 15, 1999.
Effective date: December 15, 1999.
Amendment No.: 109.
Facility Operating License No. NPF-58: This amendment revised the
Operating License.
Date of initial notice in Federal Register: November 3, 1999 (64 FR
59803).
The Commission's related evaluation of the amendment is contained
in a
[[Page 73105]]
Safety Evaluation dated December 15, 1999.
No significant hazards consideration comments received: No.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: June 29, 1999, as supplemented
August 27, October 29, and November 3, 1999.
Brief description of amendment: The amendment clarifies the
authority to possess certain types of radioactive materials and
components at either Unit 1 or Unit 2. Following the transfer of the
Three Mile Island, Unit 1 (TMI-1), operating license to AmerGen, these
items, under the amendment, may continue to be moved between the TMI-1
and TMI-2 units as they currently are.
Date of issuance: December 9, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 217.
Facility Operating License No. DPR-50: Amendment revised the
License.
Date of initial notice in Federal Register: July 12, 1999 (64 FR
37572). The August 27, October 29, and November 3, 1999, letters
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination or expand
the amendment beyond the scope of the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 9, 1999.
No significant hazards consideration comments received: No.
GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear
Station, Unit 2, (TMI-2) Middletown, Pennsylvania
Date of application for amendment: June 29, 1999, as supplemented
by letters dated August 27, October 29, and November 3, 1999.
Brief description of amendment: The amendment adds a provision to
the license conditions to ensure that the storage of certain types of
radioactive materials and components at Three Mile Island (TMI), Unit
2, pursuant to the TMI, Unit 1 license, does not result in a source
term that would exceed the limits in the TMI, Unit 2 Post-Defueling
Monitored Storage Safety Analysis Report.
Date of issuance: December 14, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 53.
Facility Operating License No. DPR-73: Amendment revised the
License.
Date of initial notice in Federal Register: July 12, 1999 (64 FR
37572). The August 27, October 29, and November 3, 1999, supplements
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination or expand
the amendment beyond the scope of the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 14, 1999.
No significant hazards consideration comments received: No.
Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit
1, DeWitt County, Illinois
Date of application for amendment: July 23, 1999, as supplemented
July 30, August 9, August 20, October 7, and October 11, 1999.
Brief description of amendment: The amendment replaces references
to Illinois Power Company in the Operating License with references to
AmerGen Energy Company, LLC, to reflect the transfer of the license as
approved by an Order dated November 24, 1999.
Date of issuance: December 15, 1999.
Effective date: December 15, 1999.
Amendment No.: 123.
Facility Operating License No. NPF-62: The amendment revised the
Operating License.
Date of initial notice in Federal Register: August 19, 1999 (64 FR
45290).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 24, 1999.
Comments received: Yes. Comments received from The Environmental
Law and Policy Center of the Midwest were addressed in the staff's
safety evaluation.
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 23, 1999, as
supplemented October 11 and November 10, 1999.
Brief description of amendments: The amendments provide approval to
move steam generator sections through the auxiliary building and to
disengage crane travel interlocks, and provide relief from performance
of Technical Specification Surveillance Requirement 4.9.7.1. The loads
to be moved are in support of the Unit 1 Steam Generator Replacement
Project.
Date of issuance: December 7, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 233 and 216.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 26, 1999 (64 FR
57665). The October 11, 1999, submittal provided corrected TS pages.
The November 10, 1999, submittal was in response to a NRC request for
additional information dated October 26, 1999, and provided clarifying
information to the original submittal. This information was within the
scope of the original Federal Register notice and did not change the
staff's initial proposed no significant hazards considerations
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 7, 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 1, 1999, as
supplemented November 19, 1999.
Brief description of amendments: The amendments involve the
resolution of an unreviewed safety question related to certain small-
break loss-of-coolant accident scenarios for which there may not be
sufficient containment recirculation sump water inventory to support
continued operation of the emergency core cooling system and
containment spray system pumps during and following switchover to cold
leg recirculation. Resolution of this issue consists of a combination
of physical plant modifications, new analyses of containment
recirculation sump inventory, and resultant changes to the accident
analyses to ensure sufficient water inventory in the containment
recirculation sump. The amendments would also change the Technical
Specifications dealing with the refueling water storage tank inventory
and temperature, the required amount of ice in each ice basket in the
containment, and the delay to start the containment air recirculation/
hydrogen skimmer fans.
Date of issuance: December 13, 1999.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 234 and 217.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
[[Page 73106]]
Date of initial notice in Federal Register: October 29, 1999 (64 FR
58458).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 13, 1999.
No significant hazards consideration comments received: No.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: February 12, 1999.
Brief description of amendment: The amendment changes the Technical
Specifications to (1) allow reactor vessel hydrostatic and leakage
tests when reactor coolant temperature is above 212 deg.F without
maintaining primary containment integrity and (2) establish a limit and
a surveillance requirement on reactor coolant activity when reactor
coolant temperature is above 212 deg.F, the reactor is not critical,
and primary containment has not been established.
Date of issuance: November 24, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 107.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 24, 1999 (64 FR
14283).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 24, 1999.
No significant hazards consideration comments received: No.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: September 30, 1999.
Brief description of amendment: The amendment changes the Technical
Specification surveillance periodicity requirements for the control
room emergency filtration system.
Date of issuance: December 8, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 108.
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 3, 1999 (64 FR
59805).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 8, 1999.
No significant hazards consideration comments received: No.
PECO Energy Company, Public Service Electric and Gas Company,Delmarva
Power and Light Company, and Atlantic City Electric Company, Docket No.
50-278, Peach Bottom Atomic Power Station, Unit No. 3, York County,
Pennsylvania
Date of application for amendment: March 1, 1999, as supplemented
June 14, October 1 and October 6, 1999.
Brief description of amendment: The amendment supports the
installation of a digital Power Range Neutron Monitoring system and the
incorporation of the long-term thermal-hydraulic stability solution
hardware.
Date of issuance: October 14, 1999.
Effective date: Effective as of date of issuance and shall be
implemented prior to restart from the Peach Bottom Atomic Power
Station, Unit 3, October 1999 refueling outage.
Amendment No.: 234.
Facility Operating License No. DPR-56: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 2, 1999 (64 FR
29711). The June 14, October 1 and October 6, 1999, 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 October 14, 1999.
No significant hazards consideration comments received: No.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: April 6, 1999.
Brief description of amendment: The amendment changes the Technical
Specifications by removing the words ``three individual underground''
and ``underground'' from the limiting conditions for operation when
referring to the emergency diesel generator fuel oil storage tanks in
Sections 3.7.A.5 and 3.7.F.4.
Date of issuance: December 7, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 198.
Facility Operating License No. DPR-64: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 2, 1999 (64 FR
29713).
No significant hazards consideration comments received: No.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 7, 1999.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., (SNC) Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Dates of amendments request: March 12, 1998, as supplemented by
letters of April 24, 1998, August 20, 1998, November 20, 1998, February
3, 1999, February 20, 1999, April 30, 1999 (two letters), May 28, 1999,
June 30, 1999, July 27, 1999, August 19, 1999, August 30, 1999,
September 15, 1999, and September 23, 1999.
Brief description of amendments: The amendments fully convert SNC's
Current TS (CTS) to Improved TS (ITS) based on NUREG-1431, ``Standard
Technical Specifications, Westinghouse Plants,'' Revision 1, of April
1995. The amendments add two new Additional Conditions to Appendix C of
the Unit 1 and Unit 2 Facility Operating Licenses. The first new
Additional Condition authorizes SNC to relocate certain CTS
requirements to SNC-controlled documents. The second new condition
addresses the schedule for performing new and revised ITS
surveillances.
Date of issuance: November 30, 1999.
Effective date: As of the date of issuance and shall be implemented
no later than March 31, 2000.
Amendment Nos.: 146 and 137.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments fully
convert SNC's CTS to ITS.
Dates of initial notices in Federal Register: May 25, 1999 (64 FR
28218) and August 25, 1999 (64 FR 46443). The supplemental letters
dated April 24, 1998, August 20, 1998, November 20, 1998, February 3,
1999, February 20, 1999, April 30, 1999 (two letters), May 28, 1999,
June 30, 1999, July 27, 1999, August 19, 1999, August 30, 1999,
September 15, 1999, and September 23, 1999, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determinations.
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.
[[Page 73107]]
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: April 28, 1999.
Brief description of amendments: The amendments revised Vogtle's
operating licenses to allow the licensee to establish containment
hydrogen monitoring within 90 minutes of initiation of a safety
injection following a loss-of-coolant accident, compared to the current
30 minute requirement.
Date of issuance: December 8, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 110 and 88.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Operating Licenses.
Date of initial notice in Federal Register: August 11, 1999 (64 FR
43779).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 8, 1999.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: July 28, 1998, as supplemented by
letters dated May 31 and October 21 (2 letters), 1999.
Brief description of amendments: The amendments authorize the
revision of the South Texas Project updated final safety analysis
report (UFSAR) to allow the use of operator action to reduce the steam
generator power-operated relief valve setpoint consistent with the
revised small-break loss-of-coolant accident analysis for the
replacement Delta 94 SGs.
Date of issuance: December 14, 1999.
Effective date: December 14, 1999. Revisions will be incorporated
into the next UFSAR update in accordance with the schedule in 10 CFR
50.71(e).
Amendment Nos.: Unit 1--119, Unit 2--107.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
authorize revision of the UFSAR.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48268).
The May 31 and October 21 (2 letters), 1999, supplements provided
additional clarifying information. One of the October 21, 1999,
supplements also provided a revised UFSAR pages. This information was
within the scope of the original application and Federal Register
notice and did not change the staff's initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 14, 1999.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant , Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: September 30, 1999.
Description of amendment request: The amendments revise the
operating licenses to remove license conditions that have become
outdated, are no longer applicable, or are redundant, and to
consolidate license conditions which currently exist in two locations
in each units license.
Date of issuance: December 16, 1999.
Effective date: December 16, 1999.
Amendment Nos.: 237, 262, and 222.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the licenses.
Date of initial notice in Federal Register: November 3, 1999 (64 FR
59807).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 1999.
No significant hazards consideration comments received: No.
TXU Electric, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: February 27, 1998, as supplemented by
letters dated June 10, 1998, and October 22, 1999.
Brief description of amendments: The amendments change the
refueling water storage tank (RWST) low-low level setpoints in
Technical Specification Table 3.3.2-1, ``Engineered Safety Feature
Actuation System Instrumentation,'' to increase the volume of water
available to containment spray pumps when the containment spray system
switches to the recirculation mode of operation.
Date of issuance: December 8, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 73 and 73.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38205). The October 22, 1999, supplement provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the
application beyond the scope described in the initial notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 8, 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: August 18, 1999.
Brief description of amendment: The amendment revises the reactor
core spiral reloading pattern such that it begins around a source range
monitor. The offloading pattern is the reverse sequence.
Date of Issuance: December 14, 1999.
Effective date: As of its date of issuance, and shall be
implemented within 30 days.
Amendment No.: 181.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 8, 1999 (64
FR 48867).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated December 14, 1999.
No significant hazards consideration comments received: No.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: September 23, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) by deleting the test requirements for
snubbers from the TSs. These requirements are already included in the
Point Beach Nuclear Plant In-Service Inspection Program.
Date of issuance: December 6, 1999.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 191 and 196.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
[[Page 73108]]
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71977).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 6, 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: December 29, 1998, as supplemented by
letters dated July 29 and October 21, 1999.
Brief description of amendment: The amendment revised (1) the
reactor coolant system (RCS) heatup and cooldown limit curves in
Figures 3.4-2 and 3.4-3 and cold overpressure mitigation system power-
operated relief valve setpoint limit curve in Figure 3.4-4 of the
current TSs, and (2) the list of references in Section 5.6.6 on the RCS
pressure temperature limits report (PTLR) in the improved TSs. The
improved TSs were issued in Amendment No. 123, dated March 31, 1999, to
replace the current TSs, but have not yet been implemented. The
revision to Section 5.6.6 of the improved TSs replaced the previous
references to NRC documents giving criteria for the above limit curves
in the current TSs by the references to (1) the NRC letter of December
2, 1999, that approved the use of the PTLR of Generic Letter 96-03,
``Relocation of the Pressure Temperature Limit Curves and Low
Temperature Overpressure Protection System Limits,'' dated January 31,
1996, for WCGS, and (2) WCAP-14040-NP-A, ``Methodology Used to Develop
Cold Overpressure Mitigation System Setpoints and RCS Heatup and
Cooldown Limit Curves.'' The PTLR will provide the methodology for the
licensee to revise the heatup and cooldown and setpoint limit curves
for WCGS in the future without prior staff approval, after the improved
TSs are implemented and have replaced the current TSs. The improved TSs
are to be implemented by December 31, 1999.
Date of issuance: December 7, 1999.
Effective date: December 7, 1999, to be implemented by December 31,
1999.
Amendment No.: 130.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9023) and September 8, 1999 (64 FR 48869). The October 21, 1999,
supplemental letter 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 7, 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: November 8, 1999.
Brief description of amendment: The amendment corrects 15 errors in
the improved Technical Specifications that was issued in Amendment No.
123 on March 31, 1999. In addition, four corrections to Table LG,
``Details Relocated from Current Technical Specifications [CTS],'' that
was attached to the safety evaluation dated March 31, 1999, issued with
Amendment No. 123 were made.
Date of issuance: December 16, 1999.
Effective date: December 16, 1999, to be implemented December 31,
1999.
Amendment No.: 131.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 16, 1999 (64
FR 62231).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 1999.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 8th day of December 1999.
For the Nuclear Regulatory Commission.
Suzanne C. Black,
Deputy Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 99-33684 Filed 12-28-99; 8:45 am]
BILLING CODE 7590-01-P