Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations
Federal RegisterAug 27, 1997
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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 August 4, 1997, through August 15, 1997. The
last biweekly notice was published on August 13, 1997 (62 FR 43365).
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
[[Page 45453]]
margin of safety. The basis for this proposed determination for each
amendment request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
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 September 26, 1997, 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 at the local public
document room for the particular facility involved. 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. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for
[[Page 45454]]
amendment which is available for public inspection at the Commission's
Public Document Room, the Gelman Building, 2120 L Street, NW.,
Washington, DC, and at the local public document room for the
particular facility involved.
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: June 12, 1997
Description of amendments request: The proposed amendments would
revise the Limiting Condition for Operation (LCO) of Technical
Specification 3.6.1.6 to limit drywell average air temperature instead
of primary containment average air temperature, which is the volume-
weighted average of both drywell and wetwell atmospheres. This change
in monitored parameter is consistent with the approach taken in the
improved standard technical specifications for boiling water reactor
(BWR) plants of this type (NUREG-1433, Rev. 1, ``Standard Technical
Specifications General Electric Plants, BWR/4,'' April 1995). The
proposed amendments would additionally change the temperature limit in
this LCO from 135 deg.F (primary containment average air temperature)
to 150 deg.F (drywell average air temperature).
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 NRC has provided standards in 10 CFR 50.92 for determining
whether a significant hazards consideration exists. A proposed
amendment to an operating license for a facility involves no
significant hazards consideration if 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, (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. Carolina Power & Light Company has reviewed these proposed
license amendment requests and has concluded that their adoption
would not involve a significant hazards consideration. The basis for
this determination follows.
1. The probability of previously evaluated accidents is not a
function of the ambient drywell air temperature. The revised drywell
average air temperature limit of 150 deg.F does not affect any
instrumentation setpoints or allowable values, so [the] likelihood
of plant instrumentation initiating a plant transient or accident
has not been increased.
The design basis accidents were re-evaluated using an initial
drywell air temperature of 150 deg.F. The evaluation results
indicate that no containment design requirements are exceeded nor
are any regulatory requirements exceeded. Analyses demonstrate that
an initial drywell average air temperature of 150 deg.F will ensure
that the safety analysis remains valid by ensuring that the peak
loss-of-coolant accident drywell temperature does not result in the
drywell structure exceeding the maximum allowable temperature of
300 deg.F. Indeed, these evaluations indicate that both the peak
drywell pressure and temperature will be slightly less than the peak
drywell pressure and temperature resulting from the current
135 deg.F primary containment air temperature limit. Since the
drywell temperature and pressure associated with a postulated design
basis accident remain less than the drywell maximum design allowable
values, revised drywell average air temperature limit of 150 deg.F
does not increase the consequences of an accident previously
evaluated.
A temporary, one-time exception footnote for the Brunswick Steam
Electric Plant (BSEP), Unit No. 2 is being deleted because the
period of the footnote's applicability expired on August 15, 1985.
Deletion of this footnote is an administrative change that has no
effect on the probability or consequences of an accident previously
evaluated.
Thus, based on the above, the proposed license amendments do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated. Revising the primary containment temperature limit basis
to use the drywell average air temperature and increasing the
average air temperature limit from 135 deg.F to 150 deg.F does not
physically modify the facility nor does the proposed revision modify
the operation of any existing plant equipment. A temporary, one-time
exception footnote for BSEP Unit No. 2 is being deleted because the
period of the footnote's applicability expired on August 15, 1985.
Deletion of this footnote is an administrative change that does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety. The drywell average airspace
temperature affects the calculated containment response to
postulated Design Basis Accidents. Analyses demonstrate that an
initial drywell average air temperature of 150 deg.F will ensure
that the safety analysis remains valid by ensuring that the peak
loss-of-coolant accident drywell air temperature does not result in
the drywell structure exceeding the maximum allowable temperature of
300 deg.F. Analyses performed using an initial drywell average air
temperature of 150 deg.F also demonstrate that containment design
requirements for peak post-accident suppression pool temperature,
design basis accident related discharge loads for safety-relief
valve piping, and net positive suction head for residual heat
removal system and core spray system pumps are met. In addition,
setpoints for reactor water level instrumentation located in the
drywell have not been adversely affected, drywell equipment
environmental qualification is being maintained, and containment
performance during a postulated station blackout is not being
adversely affected. Therefore, the proposed change does not involve
a significant reduction in a margin of safety. The deletion of a
temporary, one-time exception footnote for BSEP Unit No. 2 is an
administrative change that also 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.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Gordon E. Edison (Acting)
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: July 18, 1997Description of amendments
request: The proposed amendments would revise two specifications
included in the Design Features section of the Technical Specifications
(TS). The value for primary containment suppression chamber design
temperature (TS 5.2.2.b) would be increased from 200 deg.F to
220 deg.F. The licensee has determined that the original suppression
chamber design temperature was 220 deg.F and confirmed that it is still
the correct design value. Secondly, the specification for reactor
coolant system volume (TS 5.4.2) would be redefined as the vessel
volume, rather than the vessel and recirculation system volume,
resulting in a change in the associated value from 18,670 cubic feet to
18,320 cubic feet. Additionally, the proposed amendments would correct
a typographical error in Design Features TS 5.3.2 regarding the reactor
core control rod assemblies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 45455]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
10 CFR 50.92 provides standards for determining whether a
significant hazards consideration exists. A proposed amendment to an
operating license for a facility involves no significant hazards
consideration if 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,
(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. Carolina Power & Light Company has
reviewed these proposed license amendment requests and has concluded
that their adoption would not involve a significant hazards
consideration. The basis for this determination follows.
1. The proposed license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The proposed amendments correct an inaccurate
suppression chamber design temperature to reflect the actual design
temperature used during containment analyses and pressure vessel
procurement, correct a typographical error, and update the reactor
coolant system volume to reflect a more accurate volume used in
current analyses. These changes are administrative in nature and do
not affect the probability or consequences of any accident
previously analyzed.
2. The proposed license amendments will not create the
possibility of a new or different kind of accident from any accident
previously evaluated. These changes are administrative in nature and
correct the Technical Specifications to accurately represent
information used during existing accident analyses. These changes do
not introduce a new initiating event and do not create the
possibility of a new or different kind of accident previously
evaluated.
3. The proposed license amendments do not involve a significant
reduction in a margin of safety. As stated above, these changes are
administrative in nature and correct the Technical Specifications to
accurately represent information used during existing accident
analyses. These changes document values currently used in existing
accident analyses and, therefore, do not reduce the margin of safety
already established by the analyses.
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.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Gordon E. Edison (Acting)
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: July 1, 1997
Description of amendment request: The proposed amendments would
revise Technical Specification Table 3.3.7.1-1, ``Radiation Monitoring
Instrumentation,'' to require two channels to be operable per trip
system as opposed to two per intake. This change reflects a
modification to the design of the instrument logic to satisfy single
failure requirements. The amendment would also revise the associated
action statement to clarify system logic wording.
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 because:
The proposed Technical Specification (TS) change clearly defines
the system logic and the specific actions required for system
operability. It will not change the probability of occurrence of any
accidents, because the affected radiation monitoring instrumentation
is not an accident initiator. UFSAR [Updated Final Safety Analysis
Report] Section 15.9.3.4 analyzed the effects of the loss of
ventilation from the Main Control Room in the event of a Station
Black Out (SBO). The scope of work for the design change associated
with this TS change does not affect this analysis or any of its
assumptions The consequences of an accident will not increase,
because the trip system redundancy is being restored to meet design
basis requirements. The proposed design change will eliminate the
potential of exposing main control room personnel to radiation doses
that exceed the limits specified in General Design Criteria (GDC)
19. The design change associated with this TS change will comply
with the redundancy due to two trip systems, either of which will
actuate the control room emergency makeup train as required and the
potential for spurious actuations will be reduced due to the logic
change to require two channels of one trip system to cause
actuation. The overall control logic for the remaining portions of
the CREFS [Control Room Emergency Filtration System] is not changed
by the design change.
The changes proposed to the actions are intended to clarify
system logic wording. The actions assure that automatic trip
capability is maintained and if not, then the CREFS is placed in the
pressurization mode as in the current TS. This is consistent with
the current TS.
Based upon the above, the proposed amendment will not increase
the probability or consequences of any accident previously
evaluated.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The elimination of the electrical connection between the
redundant trip systems in a given CREFS subsystem will restore trip
system independence and eliminate the potential of a single failure
disabling the radiation monitoring instrumentation trip function.
Specifically, a single failure, resulting from a blown fuse caused
by a fault in the affected existing circuit, could remove the
control power to the isolation logic relays in both trip systems.
These relays require power in order to actuate and perform their
safety function. A loss of control power to both trip systems due to
the fault could result in exposing main control room personnel to
radiation doses that exceed GDC 19 limits.
In addition, the changes to Action Statement 70 of the
specification assure that trip capability is maintained.
Based upon the above, the proposed change will not create the
possibility of a new or different kind of accident or transient
previous evaluated.
3) Involve a significant reduction in the margin of safety
because:
The proposed TS change will not prevent the isolation logic
relays from performing their function or cause false trips. The
alarm/trip setpoints for the affected monitors (including their
measurement ranges) remain unchanged. The changes proposed to the
actions are intended to clarify system logic wording. The actions
assure that automatic trip capability is maintained and if not, then
the CREFS is placed in the pressurization mode as in the current TS.
This is consistent with the current TS.
Based on the above, the proposed TS change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document location: Jacobs Memorial Library, Illinois
Valley Community College, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra
[[Page 45456]]
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 5, 1997
Description of amendment request: The proposed amendment would
revise the Technical Specifications for the Safety Limit Minimum
Critical Power Ratio (SLMCPR) for Cycle 8 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:
The plant/cycle specific SLMCPRs have been calculated using
methods identical to those used by GE (General Electric) to assess
the SLMCPR for other BWRs (boiling water reactors). Similar methods
were used to determine the value of the SLMCPR for the previous
cycle. These methods are within the existing design and licensing
basis and cannot increase the probability or severity of an
accident. The basis of the SLMCPR calculation is to ensure that
greater that 99.9% of all fuel rods in the core avoid transition
boiling and fuel damage in the event of the occurrence of
Anticipated Operational Occurrences (AOO) or a postulated accident.
The SLMCPR is used to establish the Operating Limit Minimum
Critical Power Ratio (OLMCPR). Neither the SLMCPR nor the OLMCPR are
initiators or affect initiators of an accident previously evaluated
and therefore changes to the SLMCPR do not increase the probability
of any accident previously evaluated. The proposed changes involve
the use of an accepted methodology in calculating the SLMCPR and,
since there is no change in the definition of the SLMCPR, these
changes will not affect the consequences of any accident previously
evaluated. In addition, the proposed changes do not involve any
change in the way the plant is operated. Existing procedures will
ensure that the SLMCPR is not violated. Therefore, these changes
have no effect on the consequences of an accident.
On these bases, there will be no increase in the probability or
consequences of an accident previously analyzed as a result the
proposed changes.
The proposed changes consist of SLMCPR calculated from an
accepted method of analysis which has been used by many BWRs. These
changes do not involve any alteration of the plant and do not affect
the plant operation. Neither the SLMCPR nor the OLMCPR can initiate
an event, therefore a change to the SLMCPR does not create the
possibility of occurrence of a new or different kind of accident
from any accident previously evaluated.
The SLMCPR is a Technical Specification numerical value to
ensure that 99.9% of all fuel rods in the core will avoid transition
boiling if the limit is not violated. The proposed SLMCPR change
results from SLMCPR analysis using the accepted methods as
identified in the Attachment.
The margin of safety resides between the SLMCPR and the point at
which fuel fails. Maintaining the MCPR above the proposed SLMCPR
will maintain the margin of safety associated with GE's SLMCPR
methodology. Existing plant procedures will continue to ensure that
the SLMCPR is not violated.
Therefore, this request does not involve a reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: Government Documents Department,
Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: James W. Clifford, Acting
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: July 28, 1997
Description of amendment request: This amendment is to modify the
actions associated with Technical Specifications Table 3.3-1 for the
Reactor Protective Instrumentation and Table 3.3-3 for the Engineered
Safety Feature Actuation System Instrumentation.
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:
An evaluation of the proposed change has been performed in
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards
considerations using the standards in 10 CFR 50.92(c). A discussion
of these standards as they relate to this amendment request follows:
1. Does Not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
The proposed change to the ANO-2 Technical Specifications (TS)
modifies the allowed outage time that a channel of the Refueling
Water Tank (RWT) Level - Low or Steam Generator differential
pressure (delta P) can be in the tripped condition from a maximum of
approximately 18 months when one channel is inoperable, and 31 days
when two channels are inoperable, to 48 hours for either of these
conditions.
If a channel of RWT Level Low is in the tripped condition and a
single failure occurs that results in one of the other three
channels of RWT Level - Low to actuate, a Recirculation Actuation
System (RAS) signal would be generated. This scenario would not be
considered severe if the condition occurred as a single event.
However, during the injection phase of a Loss of Coolant Accident
(LOCA) with a channel of RWT Level - Low in the trip condition with
the above single failure, a premature RAS actuation would be the
result. The premature RAS actuation would prevent the contents of
the RWT from being injected into the reactor coolant system and
possibly resulting in failure of both trains of Emergency Core
Cooling System (ECCS) and the Containment Spray System.
With one channel of Steam Generator delta P in the tripped
condition, as allowed by the TS, the plant is vulnerable to the
single failure of a second Steam Generator delta P channel under an
unisolable Main Steam Line Break condition. The following scenario
will result in the faulted Steam Generator being supplied feedwater
by the Emergency Feedwater System during an unisolable Main Steam
Line Break. One channel of Steam Generator delta P is in the tripped
condition as allowed by the TS and a Main Steam Line Break occurs
that is unisolable. During this event one of the remaining channels
of Steam Generator delta P fails resulting in incorrectly feeding
the faulted Steam Generator. Reducing the time that a channel of RWT
Level - Low or Steam Generator delta P can be placed in the tripped
condition will reduce the probability of these scenarios from
occurring.
The consequences of feeding the faulted Steam Generator during a
main steam line break event or a premature RAS actuation during a
LOCA are both significant. The proposed change reduces the allowed
time a channel of RWT Level - Low or Steam Generator delta P can be
in the tripped condition. Reducing the time the channel can be in
the tripped condition and thus, the exposure time to this scenario,
would not be an accident initiator or involve an increase in the
consequences of any accident previously evaluated.
The remaining proposed changes are consistent with NUREG-1432,
``Standard Technical Specifications for Combustion Engineering
Plants'' and are intended to correct the actions required by TS
Tables 3.3-1 and 3.3-3 to the current NRC approved guidance.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Does Not Create the Possibility of a New or Different Kind of
Accident from any Previously Evaluated.
The proposed change does not modify the design or configuration
of the plant. The proposed change provides a more conservative time
limit for a channel to be in the tripped condition and provides the
required actions when a channel is out of service. There has been no
physical change to plant systems, structures or components nor will
the proposed change reduce the ability of any of the safety related
equipment required to mitigate anticipated operational
[[Page 45457]]
occurrences or accidents. This change will potentially increase the
ability of safety related equipment to perform their functions. The
configuration allowed by the proposed specification is permitted by
the existing specification.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Does Not Involve a Significant Reduction in the Margin of
Safety.
The proposed change provides a more restrictive time limit for a
channel of RWT Level Low or Steam Generator delta P to be in the
tripped condition than is currently allowed by the TS. By reducing
the allowed time, the probability is reduced that a single failure
of another channel would result in a premature RAS actuation during
the injection phase of a LOCA or the feeding of a faulted Steam
Generator. By limiting the vulnerability to these events and their
consequences, the proposed change will increase the margin of
safety.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, Entergy Operations has
determined that the requested change does 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.
Local Public Document location: Tomlinson Library, Arkansas Tech
University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: James W. Clifford, Acting
Florida Power and Light Company, Docket No. 50-335, St. Lucie
Plant, Unit No. 1, St. Lucie County, Florida
Date of amendment request: July 22, 1997
Description of amendment request: The proposed amendment will
incorporate a recent evaluation of a postulated inadvertent opening of
a Main Steam Safety Valve (MSSV) into the current licensing basis for
St. Lucie Unit 1. An assessment of the potential consequences of this
specific transient is not presently contained in the Updated Final
Safety Analysis Report (UFSAR), and the proposed license amendment is
required by 10 CFR 50.59(c).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The Unit 1 UFSAR includes analyses for excess load events;
however, a stuck open MSSV is not specifically evaluated in the
UFSAR. This proposed amendment will add an evaluation of an
inadvertent opening of an MSSV to the licensing basis of the plant.
The probability of occurrence of an excess load event is not
increased by this amendment since the frequency of initiating events
has not changed and there is no change to the plant or plant
operation as a result of this amendment. Thus, there is no
significant increase in the probability of any accident previously
analyzed.
The radiological consequences of an excess load event other than
steam line ruptures are discussed in UFSAR Section 15.2.11.2.3, and
are based on the inadvertent opening of an Atmospheric Steam Dump
Valve (ADV). This proposed amendment revises the radiological
consequences of the UFSAR excess load event to incorporate the
results of a recent evaluation of an inadvertent opening of an MSSV.
The consequences of the postulated MSSV scenario are greater than
those of an inadvertent opening of an ADV, but the predicted two
hour site boundary doses remain a small fraction of 10 CFR 100
limits. In addition, the Unit 1 results are bounded by the St. Lucie
Unit 2 analysis results which are reported in Section 15.1.3.1.1.3
of the Unit 2 UFSAR. Therefore, operation of the facility in
accordance with the proposed amendment will not involve a
significant increase in the consequences of an accident previously
evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will add an evaluation of an inadvertent
opening of an MSSV to the licensing basis of the plant. The
evaluation addresses an anticipated operational occurrence (AOO) and
is classified as an Excess Load event under the PSL1 [Plant St.
Lucie Unit 1] accident classification criteria. Although an analysis
of this specific transient is not currently provided in the UFSAR,
analyses of Excess Load events other than steam line ruptures are
reported in UFSAR Section 15.2.11. The amendment does not change
plant design or operation and does not introduce new failure modes
or system interactions. Thus, operation of the facility with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed license amendment adds an engineering evaluation to
the licensing basis of the plant to address the consequences of a
postulated stuck open MSSV. A change is not being made to plant
design or operation. A change is not being made to any Technical
Specification Limiting Condition for Operation, Action, or
Surveillance Requirement. The evaluation demonstrates that, post-
trip, the reactor would remain subcritical throughout the transient,
and that the radiological consequences of a stuck open MSSV are a
small fraction of 10 CFR 100 limits. Therefore, operation of the
facility in accordance with the proposed amendment would not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: Indian River Community College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420
NRC Project Director: Frederick J. Hebdon
Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: August 1, 1997
Description of amendment request: The proposed amendment will
extend the semi-annual surveillance interval specified in Table 4.3-2
of the Technical Specifications for testing the Engineered Safety
Features Actuation System (ESFAS) subgroup relays to an interval
consistent with Combustion Engineering Owners Group Report CEN-403,
Revision 1-A, March 1996. The proposed surveillance interval is at
least once per 18 months, with testing to be performed on a staggered
test basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility, in accordance with the proposed
amendment, would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment revises the testing frequency of ESFAS
subgroup relays, and is based on demonstrated relay reliability.
These relays actuate the engineered safety features (ESF) equipment
which is installed to mitigate design basis accidents. ESF system
components are not considered initiators of any design basis
accident. Therefore, operation of the facility
[[Page 45458]]
with the proposed amendment would not involve a significant increase
in the probability of an accident previously evaluated.
The proposed amendment does not alter the design or operation of
ESF systems. The mean time between failures demonstrated by the
ESFAS subgroup relays is significantly greater than the proposed
surveillance interval, and testing will be performed on a staggered
test basis. This, in addition to ESF redundancy, provides assurance
that these systems will continue to function as evaluated to
mitigate design basis accidents. Therefore, operation of the
facility, in accordance with the proposed amendment, would not
involve a significant increase in the consequences of an accident
previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. The changes do
not involve the addition of new equipment or the modification of
existing equipment, nor do they alter the design of St. Lucie plant
systems. Therefore, operation of the facility, in accordance with
the proposed amendment, would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment revises the surveillance interval for
testing the ESFAS subgroup relays consistent with the Combustion
Engineering Owners Group topical report CEN-403, Revision 1-A, and
conforms to criteria specified in the associated safety evaluation
issued by the NRC staff. The St. Lucie Unit 2 subgroup relay mean
time between failures is significantly greater than the proposed
surveillance interval, and testing will be performed on a staggered
test basis. ESFAS setpoints, system operation, and plant
configuration will not be changed, and the subgroup relays are not
subject to time-related instrument drift. Accident analyses
assumptions, initial conditions, and conclusions reported in the
Updated Final Safety Analysis Report are not changed by the revised
surveillance interval. Therefore, operation of the facility in
accordance with the proposed amendment would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420
NRC Project Director: Frederick J. Hebdon
GPU Nuclear (GPUN) Corporation, et al., Docket No. 50-289, Three
Mile Island Nuclear Station, Unit No. 1, Dauphin County,
Pennsylvania
Date of amendment request: July 30, 1997
Description of amendment request: The purpose of this Technical
Specification change request (TSCR) is to incorporate additional system
leakage limits and leak test requirements for systems outside
containment which were not previously contained in Technical
Specification 4.5.4 nor considered in the TMI-1 Updated Final Safety
Analysis Report (UFSAR) design basis accident (DBA) analysis dose
calculations for 2568 MWt. This TSCR also revises the Technical
Specification 3.15.3 Bases for the Auxiliary and Fuel Handling Building
Ventilation System (AFHBVS). The revisions to Technical Specification
3.15.3 Bases for the AFHBVS serve to clarify system design requirements
and accident analysis considerations. The revision states that the
AFHBVS is not credited in reducing off-site dose for the Maximum
Hypothetical Accident (MHA) or the Waste Gas Tank Rupture (WGTR)
accident analysis dose calculations.
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 (SHC), which is presented below:
GPUN has determined that this TSCR poses no significant hazards
consideration as defined by 10 CFR 50.92.
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. No physical modifications which would change
structures, systems, or components are being made or proposed by
this TSCR. This change has no [effect] on the LOCA [loss-of-coolant
accident] safety analysis for ECCS [emergency core cooling system]
performance. The results of revised MHA dose calculation are less
than that previously evaluated in the UFSAR for the exclusion area
boundary (EAB). In addition the doses are below the 10 CFR 100
guideline limits for both the EAB and low population zone (LPZ) ...,
and below the 10 CFR 50 Appendix A, GDC [General Design Criteria]-19
limits for the control room. The LPZ increases in dose consequence
are the result of using more conservative assumptions in the revised
analyses and the new values remain a small fraction of the 10 CFR
100 limits. The WGTR dose calculation is not affected by this TSCR.
The proposed Technical Specification changes ensure that the MHA and
WGTR accident analysis parameters remain bounded during plant
operation.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. This TSCR does not
involve any physical modifications which would affect structures,
systems, or components, nor does it involve any changes in plant
operation. The only changes resulting from this TSCR are revisions
to leakage limits and testing requirements necessary to reflect the
revised MHA analysis and to correct discrepancies identified by the
NRC .... Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. This TSCR does not involve changes to Technical
Specification defined Safety Limits, Limiting Conditions for
Operation, and does not involve any change to safety system
setpoints for operation. Therefore, the proposed change does not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Ronald B. Eaton (Acting)
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, (TMI-1) Dauphin County,
Pennsylvania
Date of amendments request: August 12, 1997
Description of amendments request: The amendment requests changes
to the Surveillance Specification of the Technical Specification (TS)
for the once through steam generator (OTSG) inservice inspection for
TMI-1 Cycle 12 Refueling (12R) examinations applicable to TMI-1 Cycle
12 operation. These proposed changes impose axial and circumferential
extent sizing limitations in addition to TS requirements for
[[Page 45459]]
inside diameter (ID) initiated degradation where bobbin coil eddy
current test (ECT) signal amplitudes do not permit reliable through
wall sizing. Editorial changes are being made to improve consistency of
format, to the Bases which relate to the requested changes in Section
4.19 of the TS, and to the reporting requirements in Section 4.19.5 of
the TS.
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:
GPU Nuclear has determined that this TSCR [Technical
Specification Change Request] poses no significant hazards
consideration as defined by 10 CFR 50.92.
A. These proposed changes do not represent a significant
increase in the probability of occurrence or consequences of an
accident previously evaluated. The only accidents previously
evaluated that could be significantly affected by changes to the
OTSG tube inservice inspection requirements are the steam generator
tube rupture (STGR) and the main steam line break (MSLB) accidents.
The proposed flaw disposition strategy based on measurable eddy
current parameters of axial and circumferential extent for Inside
Diameter (ID) Initiated Inter-Granular Attack (IGA) will provide
high confidence that unacceptable flaws that do not have the
required structural integrity to withstand the MSLB are removed from
service. The proposed axial and circumferential length limits for
eddy current inside diameter degradation indications meet the RG
[Regulatory Guide] 1.121 acceptance criteria for margin to failure
for MSLB applied differential pressure and axial tube loads. The
capability for detection of flaws is unaffected and the
identification of tubes which should be repaired or removed from
service is maintained or improved. The operation of the OTSG or
related structures, systems, or components is otherwise unaffected.
Therefore, neither the probability nor consequences of a SGTR is
significantly increased either during normal operation or due to the
limiting loads of [an] MSLB accident.
Neither the editorial changes in format, punctuation, or grammar
nor the administrative changes or changes in reporting requirements,
as described above, could significantly affect the probability of
occurrence or consequences of any accident previously evaluated.
B. These proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
because there are no hardware changes involved nor changes to any
operating practices. These changes involve only the OTSG tube
inservice inspection surveillance requirements, which could only
affect the potential for OTSG primary-to-secondary leakage. The
proposed changes impose additional flaw length limits for ID IGA
that go beyond existing requirements to assure tube structural and
leakage integrity.
In addition, neither the editorial changes in format,
punctuation, or grammar nor the administrative changes, as described
above, could possibly create the possibility of an accident of a new
or different type from any previously evaluated. These changes are
included only to improve the clarity and readability of the
Technical Specifications and comply with the NRC's desire to obtain
the results of the inspections as soon as practical.
Therefore, these changes do not create the potential for single
or multiple tube ruptures or any other kind of accident different
from those that have been evaluated.
C. Those proposed changes do not involve a significant reduction
in a margin of safety because the changes are more restrictive than
the current technical specification and the margins of safety
defined in R.G. 1.121 are retained. The probability of detecting
degradation is unchanged since the bobbin coil eddy current methods
will continue to be the primary means of initial detection and the
probability of leakage from any indications left in service remains
acceptable small. The strategy for dispositioning ID initiated IGA
will continue to provide a high level of confidence that tubes
exceeding the allowable limits for tube integrity are repaired or
removed from service.
In addition, neither the editorial changes in format,
punctuation, or grammar nor the administrative changes or changes in
reporting requirements, as described above, could significantly
affect a margin of safety and are included only to improve the
clarity and readability of the Technical Specifications and comply
with the NRC's desire to obtain the results from tube inspections as
soon as practical.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Ronald B. Eaton, Acting
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, (TMI-1) Dauphin County,
Pennsylvania
Date of amendment request: August 14, 1997
Description of amendment request: The proposed license amendment,
if approved, would revise the TMI-1 Updated Final Safety Analysis
Report (UFSAR) Section 14.1.2.9-Steam Line Break analysis to include
the environmental dose consequences associated with postulated
accident-induced steam generator tube leakage not previously analyzed.
The revised environmental dose consequences for the TMI-1 Steam Line
Break analysis would be increased above the values previously reviewed
by the NRC.
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:
GPU Nuclear has determined that this License Amendment Request
poses no significant hazards as defined by 10 CFR 50.92.
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. This change has no effect on structures,
systems or components prior to the postulated steam line break
accident or any other accident. OTSG [once through steam generator]
tube loads resulting from other postulated accidents are bounded by
the calculated steam line break accident tube loads. Other TMI-1
design basis accidents, which could result in OTSG tube loads and
environmental dose consequences, involve releases within the reactor
building. These events generally result in rapid depressurization of
the primary system which minimizes the differential pressure needed
to establish a significant primary-to-secondary leak rate and the
OTSG is isolated. Accordingly, leakage to the environment as a
result of induced tube loads from postulated accidents other than
steam line break is insignificant and therefore need not be
considered. The existing steam line break criteria is maintained in
that OTSG structural integrity is assured and postulated doses
remain within 10 CFR 100 limits. The new radiological consequences
of the revised steam line break dose calculation are below 10 CFR
100 limits for the exclusion area boundary (EAB) and low population
zone (LPZ). The 10 CFR 50, Appendix A, GDC [General Design
Criterion]-19 limits for the control room are not affected by this
change since the source term assumed for the TMI-1 control room
habitability analysis remains bounding.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. This change has no
impact on any plant structures, systems or components. OTSG tube
structural integrity is maintained. The only impact is the revised
radiological consequences of the steam line break analysis to
account for hypothetical accident induced primary-to-secondary
leakage.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. This change to the steam line break
[[Page 45460]]
dose consequences does not involve a significant reduction in a
margin of safety. The new radiological consequences of the revised
steam line break dose calculation are below 10 CFR 100 limits for
the EAB and LPZ, and do not affect the TMI-1 control room
habitability analysis results. This change has no impact on any
structures, systems or components.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Ronald B. Eaton, Acting
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: July 31, 1997
Description of amendment request: The proposed amendment would
change Action Statement 36 of Technical Specification (TS) Table 3.3.3-
1, ``Emergency Core Cooling System Actuation Instrumentation,'' so as
to specify actions to be taken if one or more channels per trip
function should be inoperable in the high-pressure core spray (HPCS)
drywell pressure and reactor water level instrumentation. Presently,
Action 36 only addresses actions for the plant condition of having one
channel per trip function inoperable. Specifically, Action 36 would be
changed to require that, with the number of operable channels less than
required by the minimum operable channels per trip function
requirement, then (1) with one channel inoperable, the inoperable
channel is to be placed in the tripped condition within 24 hours or the
HPCS system is to be declared inoperable, and (2) with more than one
channel inoperable, the HPCS system is to be declared inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The changes to Table 3.3.3-1, Action 36, will allow Action 36 to
be in effect for the plant condition where more than one channel is
inoperable per trip function in the HPCS drywell pressure and
reactor water level instrumentation and will clarify the actions
required if more than one channel is inoperable. Specifically, this
action statement will allow the HPCS to be declared inoperable
rather than to initiate plant shutdown per TS 3.0.3. None of the
precursors of previously evaluated accidents are affected and
therefore, the probability of an accident previously evaluated is
not increased.
The HPCS system will continue to perform its safety function to
automatically initiate and inject water into the vessel. The out of
service time for the initiating instruments remains bounded by the
out of service time for HPCS. Therefore, these changes will not
involve a significant increase in the consequences of an accident
previously evaluated.
2. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any previously evaluated.
The changes to Table 3.3.3-1, Action 36, will allow Action 36 to
be in effect for plant conditions where more than one channel is
inoperable per trip function in the HPCS drywell pressure and
reactor water level instrumentation and will clarify the actions
required if more than one channel is inoperable. No physical
modification of the plant is involved and no changes to the methods
in which plant systems are operated are required. The changes do not
introduce any new failure modes or conditions that may create a new
or different accident. Therefore, the changes do not by themselves
create the possibility of a new or different kind of accident [from
any accident] previously evaluated.
3. The operation of Nine Mile Point Unit 2, in accordance with
the proposed amendment, will not involve a significant reduction in
a margin of safety.
The change to Table 3.3.3-1, Action 36, will allow Action 36 to
be in effect for plant conditions where more than one channel is
inoperable per trip function in the HPCS drywell pressure and
reactor water level instrumentation and will clarify the actions
required if more than one channel is inoperable. The changes do not
adversely affect any physical barrier to the release of radiation to
plant personnel or to the public. The proposed change provides
consistency between the ECCS [emergency core cooling system]
instrumentation and system TS. The TS also continues to require the
operability of other injection systems coincidental with HPCS
inoperability. The change has the benefit of avoiding unnecessary
challenges to plant systems during an unnecessary plant shutdown.
Therefore, the changes do not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: Reference and Documents Department,
Penfield Library, State University of New York, Oswego, New York 13126
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Alexander W. Dromerick, Acting Director
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 amendment request: April 14, 1997
Description of amendment request: The proposed amendment would
allow the Safety Review Committee (SRC) to perform a review, rather
than an audit, of plant staff performance. The proposed amendment also
involves a title change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response:
This amendment application does not involve a significant
increase in the probability or consequences of an accident
previously analyzed. The proposed changes allow the SRC to perform a
review, rather than an audit, of plant staff performance. This
change does not diminish the SRCs effectiveness. A review of the
1995 QA [quality assurance] audit of plant staff performance shows
that no findings were issued. This indicates that the other review
mechanisms currently in place are sufficient to ensure that plant
staff performance is monitored.
The position title change is an administrative change as all
previously performed functions are being maintained and the
responsibilities and reporting chain for this position remain the
same. Therefore, the proposed changes do not affect the probability
or consequences of any previously analyzed accident.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
This amendment application does not create the possibility of a
new or different
[[Page 45461]]
kind of accident from any accident previously evaluated. The
proposed changes affect an SRC audit requirement and a position
title. These changes do not affect plant equipment or the way the
plant operates. Therefore, they cannot create a new or different
kind of accident.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
This amendment application does not involve a significant
reduction in a margin of safety. The requested Technical
Specification revisions require the SRC to review rather than audit
facility staff performance and will not diminish the effectiveness
of the SRC. A review of the 1995 audit confirms that performance of
the annual audit is redundant as no findings or recommendations
concerning plant staff performance were made. The QA/ORG [Operations
Review Group] quarterly trend reports and SRC review of plant staff
performance are adequate to ensure that plant staff performance is
properly monitored.
The position title change is an administrative change as all
previously performed functions are being maintained and the
responsibilities and reporting chain for this position remain the
same. Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: White Plains Public Library, 100
Martine Avenue, White Plains, New York 10601
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019
NRC Project Director: Alexander W. Dromerick, Acting
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 amendment request: May 29, 1997
Description of amendment request: The amendment would revise the
definition of Containment Integrity in Section 1.10, and revise Section
3.6 and Table 3.6-1 for consistency. Several valves would be added to
Table 3.6-1 to be consistent with the revised definition in Section
1.10. The amendment would also add a footnote stating that valves SP-
SOV-506 and SP-SOV-507 in Table 4.4-1, ``Containment Isolation Valves''
are sealed from weld channel and containment penetration pressurization
system (WCCPPS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The revision of the definition of containment integrity in
Section 1.10, Section 3.6.A.1, the Basis, and the addition of
existing containment isolation valves into the Table of Containment
Isolation Valves in the Technical Specifications does not change the
design, operation or testing of the plant. Section 1.10 is being
revised to clearly cover all non-automatic containment isolation
valves, and the valves are being added to be consistent with the
revised definition. The valves being added are currently identified
as containment isolation valves and tested as specified in the Final
Safety Analysis Report. Additionally, valves CB-3, 4, 7 & 8 are
controlled in accordance with Section 1.10.5 (revised numbering) for
the airlock doors. Because the design and operation are not being
changed, the addition of the valves has no effect on the probability
or consequences of an accident.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Changing the definition in Section 1.10 and the list of
containment isolation valves for consistency does not change the
design, operation or testing of the plant. Section 1.10 is being
revised to clearly cover all non-automatic containment isolation
valves, and the valves are being added to be consistent with the
revised definition. The valves being added are currently identified
as containment isolation valves and tested as specified in the Final
Safety Analysis Report. Therefore, without changing design,
operation or testing of the plant this does not create a new or
different type of accident.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed changes in the definition for containment integrity
and the listings of Containment Isolation Valves in the Technical
Specifications does not involve a significant reduction in the
margin of safety because the change reflects current design,
operation and testing of the plant, and will not alter plant
operation.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: White Plains Public Library, 100
Martine Avenue, White Plains, New York 10601
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019
NRC Project Director: Alexander W. Dromerick, Acting
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 amendment request: June 25, 1997
Description of amendment request: The proposed amendment would
allow for up to +17/-12 steps of control rod misalignment for core
power greater than 85% rated thermal power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has determined that all pertinent licensing basis
acceptance criteria have been met, and the margin of safety as
defined in the TS [technical specification] Bases is not reduced in
any of the IP3 licensing basis accident analysis (even for
misalignments to [plus or minus] 24 steps for core power [less than
or equal to] 85% of RTP). Increasing the magnitude of allowed
control rod indicated misalignment is not a contributor to the
mechanistic cause of an accident evaluated in the FSAR [final safety
analysis report]. Neither the rod control system nor the rod
position indicator function is being altered. Therefore, the
probability of an accident previously evaluated has not
significantly increased. Because design limitations continue to be
met, and the integrity of the reactor coolant system pressure
boundary is not challenged, the assumptions employed in the
calculation of the offsite radiological doses remain valid.
Therefore, the consequences of an accident previously evaluated will
not be significantly increased.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has determined that all pertinent licensing basis
acceptance criteria have been met, and the margin of safety as
defined in the TS is not reduced in any of the IP3 licensing basis
accident analysis. Increasing the magnitude of allowed control rod
indicated misalignment is not a contributor to the mechanistic cause
of any accident. Neither the rod control system nor the rod position
indicator function is being altered. Therefore, an accident which is
new or different than any previously evaluated will not be created.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
No. Based on the Westinghouse evaluation in WCAP-14668, the
Authority has determined that all pertinent licensing basis
[[Page 45462]]
acceptance criteria have been met, and the margin of safety as
defined in the TS Bases is not reduced in any of the IP3 [Indian
Point Unit 3] licensing basis accident analysis based on the changes
to safety analyses input parameter values as discussed in WCAP-
14668. Since the evaluations in Section 3.0 of WCAP-14668
demonstrate that all applicable acceptance criteria continue to be
met, the proposed change will not involve a significant reduction in
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: White Plains Public Library, 100
Martine Avenue, White Plains, New York 10601
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019
NRC Project Director: Alexander W. Dromerick, Acting
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: June 19, 1997, as supplemented by
letters dated July 30 and 31, 1997
Description of amendment request: The proposed amendment would
provide changes to Technical Specification (TS) 4.1.3.1.2, ``Control
Rod Operability,'' TS 3.1.3.6, ``Control Rod Drive Coupling,'' TS
3.1.3.7, ``Control Rod Position Indication'', TS 3.1.4.1, ``Rod Worth
Minimizer,'' TS 3/4.1.4.2, ``Rod Sequence Control System,'' TS 3/
4.10.2, ``Special Test Exceptions - Rod Sequence Control System,'' the
Bases for TS 2.2.1.2, ``Average Power Range Monitor,'' the Bases for TS
3/4.1.4, ``Control Rod Program Controls,'' and the Bases for TS 3/
4.10.2, ``Rod Sequence Control System.'' The changes are proposed in
order to eliminate the Rod Sequence Control System (RSCS) Limiting
Condition for Operation and Surveillance Requirements from the TSs and
reduce the Rod Worth Minimizer (RWM) low power setpoint from 20% to
10%. Changes are also proposed as necessary to delete reference to the
RSCS from the TSs and to incorporate additional requirements necessary
to support the elimination of the RSCS.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
A. RSCS Deletion
The RSCS system restricts the pattern of control rods prior to a
postulated control rod drop accident (RDA) so as to minimize the
reactivity worth of the dropped rod. The RSCS provides no mitigation
following the postulated RDA. The ability to restrict the pattern of
control rods also allows the RSCS to be able to reduce the
probability of a Continuous Rod Withdrawal During Reactor Startup,
as described in the Hope Creek UFSAR [Updated Final Safety Analysis
Report] Section 15.4.1.2 and Appendix 15B. However, to determine the
consequence of such a rod withdrawal event, the RSCS is not
credited, and the rod is assumed to be fully withdrawn from the core
at its maximum rate. The RDA is therefore the only analyzed accident
impacted by the proposed deletion of the RSCS system. Since the RSCS
system plays no role in preventing a[n] RDA, it therefore does not
affect the probability of occurrence of this postulated accident.
As stated in an NRC Safety Evaluation Report dated December 27,
1987, the RSCS system is the result of requirements promulgated by
the NRC staff in the early 1970's in response to unknowns and
perceived problems relating to the RDA. The GE [General Electric]
calculational methodology being used at that time produced results
showing that, even without pattern errors, calculated enthalpies for
the RDA approached limiting values. In addition, the Rod Worth
Minimizer (RWM) Technical Specifications were not effective in
ensuring RWM availability and use, and the system was poorly
maintained and frequently bypassed thus providing no significant
protection. Second operator substitution for the RWM was used
routinely and was providing minimal protection. Finally, no reliable
study existed to address the probability of exceeding enthalpy
limits as a result of an RDA.
Information associated with the above concerns has been
significantly expanded or modified. Studies using improved
methodologies have proven significantly lower peak fuel enthalpy
values compared with methodologies in use when the RSCS was
originally developed. In addition, a reliable probability study has
been completed showing that the probability of an RDA exceeding NRC
limits is very low. As a result, NRC review of the RSCS requirements
has concluded that the RSCS system is not needed and operation
without it is acceptable provided: 1) TSs are modified to minimize
the use of the second operator option, 2) procedures and quality
control associated with the second operator option are reviewed to
ensure that this option provides an effective and truly independent
monitoring process; and 3) rod patterns used are at least equivalent
to Banked Pattern Withdrawal System (BPWS) patterns. Each of these
items has been addressed for the Hope Creek Generating Station.
As a result of the resolution of the original concerns
associated with the RDA, the RWM system and limited use of the
second operator option, when properly instituted, are now deemed to
provide adequate protection to maintain the consequences of the RDA
at an acceptable level. The remaining concerns regarding operation
without the RSCS system and proper use of the second operator
substitution option have been addressed for the Hope Creek
Generating Station. We therefore conclude that the redundant RSCS
system is no longer necessary and its deletion from the Technical
Specifications will not significantly increase the probability or
consequences of an RDA.
B. RWM Setpoint Reduction
The RWM system restricts the pattern of control rods prior to a
postulated control rod drop accident (RDA) so as to minimize the
reactivity worth of the dropped rod. The RWM provides no mitigation
following the postulated RDA. The ability to restrict the pattern of
control rods also allows the RWM to be able to reduce the
probability of a Continuous Rod Withdrawal During Reactor Startup,
as described in the Hope Creek UFSAR Section 15.4.1.2 and Appendix
15B. However, to determine the consequence of such a rod withdrawal
event, the RWM is not credited, and the rod is assumed to be fully
withdrawn from the core at its maximum rate. The RDA is therefore
the only analyzed accident impacted by the proposed reduction in the
RWM setpoint. Since the RWM system plays no role in preventing a[n]
RDA, it therefore does not affect the probability of occurrence of
this postulated accident.
Existing calculations have demonstrated that no significant RDA
can occur above 10% power. Calculations by both General Electric and
the Brookhaven National Laboratory indicate that, even with
significant error patterns, peak fuel enthalpy is reduced well below
required limits at 10% power. The 20% limit was originally required
as an extreme bound because of the then existing uncertainties in
the analyses. Based on the current analyses, the 10% level is now
acceptable and deemed to provide adequate protection to maintain the
consequences of an RDA at an acceptable level. Changing the RWM
setpoint from 20% to 10% will therefore not significantly increase
the consequences of any previously analyzed accident.
2. Do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
A. RSCS Deletion
Operation of the RSCS cannot cause or prevent an accident; this
system functions to minimize the consequences of an RDA. The Bank
Position Withdrawal Sequence (BPWS) will still be used to ensure
that rod pull pattern[s] are constrained to those assumed in the
RDA. The RSCS has no impact on the operation of any other system,
and therefore its deletion will not contribute to a malfunction in
any other equipment nor create the possibility of a new or different
accident from any accident previously evaluated.
B. RWM Setpoint Reduction
Operation of the RWM cannot cause or prevent an accident; this
system functions to minimize the consequences of an RDA. The RWM has
no impact on the operation of any
[[Page 45463]]
other system, and therefore changing its setpoint from 20% to 10%
will not contribute to a malfunction in any other equipment nor
create the possibility of a new or different accident from any
accident previously evaluated.
3. Do not involve a significant reduction in a margin of safety.
A. RSCS Deletion
When the original decisions were made regarding the need for the
RSCS system, numerous perceived problems in the RDA analysis
existed. As noted in the discussion of the consequences of
previously analyzed accidents in Item 1 above: 1) the perceived RDA
problems have been resolved; 2) reviews of the RDA have concluded
that the RSCS is not needed to mitigate the consequences of an RDA;
and 3) operation without the RSCS is acceptable. The RWM and limited
use of second operator substitution, when properly instituted, are
now deemed adequate to ensure that peak fuel enthalpies remain below
NRC limits. Therefore, the deletion of the redundant RSCS system
will not significantly decrease any margin of safety.
B. RWM Setpoint Reduction
The Bases for the HCGS TSs state that when thermal power is
greater than 20%, there is no possible rod worth that, if dropped at
the design rate of the velocity limiter, could result in a peak
enthalpy of 280 calories per gram. Existing calculations demonstrate
that the RDA is not a significant concern above 10% power, and
therefore, a mitigation system is not needed for higher power level
operation. Calculations by both General Electric and the Brookhaven
National Laboratory indicate that, even with significant error
patterns, peak fuel enthalpy is reduced well below required limits
(280 calories per gram) at 10% power. The 20% limit was originally
required as an extreme bound because of the then existing
uncertainties in the analyses. Based on the current analyses, the
10% level is now acceptable and deemed to provide adequate assurance
that the peak fuel enthalpy will remain below the NRC limits during
a postulated RDA. Changing the RWM setpoint from 20% to 10% will
therefore not significantly reduce any 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.
Local Public Document location: Pennsville Public Library, 190 S.
Broadway, Pennsville, New Jersey 08070
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit - N21, P. O. Box 236, Hancocks Bridge, New Jersey 08038
NRC Project Director: John F. Stolz
Southern Nuclear Operating Company, Inc. Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: June 30, 1997
Description of amendments request: The proposed amendments would
change the Farley Technical Specifications to: revise and clarify the
requirements for the Control Room Emergency Filtration System (CREFS),
the Penetration Room Filtration System (PRFS) and the related Storage
Pool Ventilation System (SPVS); revise the required number of radiation
monitoring instrumentation channels; and delete the Containment Purge
Exhaust Filter (CPEF) specification.
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:
Pursuant to 10 CFR 50.92, SNC [Southern Nuclear Operating
Company, Inc.] has evaluated the proposed amendments and has
determined that operation of the facility in accordance with the
proposed amendments would not involve a significant hazards
consideration. The basis for this determination is as follows:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to convert from ANSI N510-1980 to ASME
N510-1989 for specific FNP [Joseph M. Farley Nuclear Plant]
filtration surveillance testing requirements and related changes do
not affect the probability of any accident occurring. The
consequences of any accident will not be affected since the proposed
changes will continue to ensure that appropriate and required
surveillance testing for FNP filtration systems will be performed
consistent with the revised accident analyses. The results of the
fuel handling accident remain well within the guidelines of I0 CFR
Part 100 and the doses due to a LOCA [loss-of-coolant accident],
including ECCS [emergency core cooling system] recirculation loop
leakage, remain within the guidelines of I0 CFR Part 100 and General
Design Criterion 19 of Appendix A to I0 CFR Part 50. Relocating
specific testing requirements to the FNP FSAR [Final Safety Analysis
Report] has no effect on the probability or consequences of any
accident previously evaluated since required testing will continue
to be performed.
Therefore, the proposed TS [Technical Specification] changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Testing differences between ANSI N510-1980 and ASME N510-1989
have been evaluated by SNC and none of the proposed changes have the
potential to create an accident at FNP. ASME N510-1989 has been
endorsed and approved by the NRC for licensee use in NUREG 1431
[Standard Technical Specifications Westinghouse Plants]. Testing the
additional channels of radiation monitoring and verification of
penetration room boundary integrity do not require the affected
systems to be placed in configurations different from design. Thus,
no new system design or testing configuration is required for the
changes being proposed that could create the possibility of any new
or different kind of accident from any accident previously
evaluated. Relocating specific testing requirements to the FSAR has
no effect on the possibility of creating a new or different kind of
accident from any accident previously evaluated since it is an
administrative change in nature.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Conversion from the testing requirements of ANSI N510-1980
sections 10, 12, and 13 to ASME N510-1989 sections 10, 11, and 15
has been previously approved by the NRC at other nuclear facilities.
ASME N510-1989 has been approved and endorsed by the NRC in NUREG
1431. The safety factor associated with the conservative charcoal
adsorber laboratory test methods and dose calculations ensures that
doses will continue to meet the guidelines of 10 CFR Part 100 and
GDC [General Design Criterion] 19 of Appendix A to 10 CFR Part 50.
The enhanced testing of radiation monitoring instrumentation and the
penetration room boundary integrity provide additional assurance
that the acceptance criteria of the safety analyses and the
resultant margins of safety are not reduced. Relocating specific
testing requirements to the FSAR has no effect on the margin of
plant safety since required testing will continue to be performed.
Clarifying the 10 hour run with heaters on is consistent with the
Improved TS language and accomplishes the purpose for the
surveillance. Therefore, SNC concludes based on the above, that the
proposed changes do not result in a significant reduction of margin
with respect to plant safety as defined in the Final Safety Analysis
Report or the bases of the FNP technical specifications.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: Houston-Love Memorial Library, 212
W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
[[Page 45464]]
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: June 30, 1997
Description of amendments request: The proposed amendments would
change the Farley Technical Specifications to incorporate the
requirements necessary to change the basis for prevention of
criticality in the fuel storage pool. This change eliminates the need
for Boraflex as a neutron absorbing material in the fuel pool
criticality analysis for both Unit 1 and Unit 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 amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
There is no significant increase in the probability of a fuel
assembly drop accident in the spent fuel pool when considering the
presence of soluble boron in the spent fuel pool water for
criticality control. The handling of the fuel assemblies in the
spent fuel pool has always been performed in borated water.
The consequences of a fuel assembly drop accident in the spent
fuel pool are not affected when considering the presence of soluble
boron.
Although the probability of misloading an assembly in the spent
fuel racks may increase due to new assembly placement constraints,
there is no significant increase in the probability of an accidental
misloading of spent fuel assemblies into the spent fuel pool racks
that will cause a criticality accident when considering the presence
of soluble boron in the pool water for criticality control.
Sufficient soluble boron will be maintained in the spent fuel pool
to maintain keff below 0.95 following a postulated single
misload. Fuel assembly placement will continue to be controlled
pursuant to approved fuel handling procedures and will be in
accordance with the Technical Specification spent fuel rack storage
configuration limitations. The addition of the spent fuel pool
storage configuration surveillance in proposed new Technical
Specifications 3.7.14 for Unit 1 and 3.7.15 for Unit 2 will provide
increased assurance that a spent fuel pool inventory verification
will be completed in a timely manner (7 days) after the relocation
or addition of fuel assemblies in the spent fuel storage pool.
There is no significant increase in the consequences of the
accidental misloading of spent fuel assemblies into the spent fuel
pool racks because criticality analyses demonstrate that the pool
will remain subcritical following an accidental misloading if the
pool contains an adequate boron concentration. The proposed new
Technical Specifications limitations will ensure that an adequate
spent fuel pool boron concentration will be maintained.
In the event of failure of a spent fuel pool cooling pump, or
loss of cooling to a spent fuel pool heat exchanger, the second
spent fuel pool cooling train provides 100 percent backup
capability, thus ensuring continued cooling of the spent fuel pool.
However, even if a loss of spent fuel pool cooling were to occur,
there is sufficient soluble boron to prevent Keff from
exceeding 0.95.
There is no significant increase in the probability of the loss
of normal cooling to the spent fuel pool water when considering the
presence of soluble boron in the pool water for subcriticality
control since a high concentration of soluble boron has always been
maintained in the spent fuel pool water.
A loss of normal cooling to the spent fuel pool water causes an
increase in the temperature of the water passing through the stored
fuel assemblies. This causes a decrease in water density which would
result in a decrease in reactivity when Boraflex neutron absorber
panels are present in the racks.
However, since Boraflex is not considered to be present, and the
spent fuel pool water has a high concentration of boron, a density
decrease causes a positive reactivity addition. However, the
additional negative reactivity provided by the proposed 2000 ppm
boron concentration limit, above that provided by the concentration
required to maintain Keff less than or equal to 0.95 (400
ppm), will compensate for the increased reactivity which could
result from a loss of spent fuel pool cooling event. Because
adequate soluble boron will be maintained in the spent fuel pool
water, there is no significant increase in the consequences of a
loss of normal cooling to the spent fuel pool.
Therefore, based on the conclusions of the above analysis, the
proposed changes will not involve a significant 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
analyzed.
Spent fuel handling accidents are not new or different types of
accidents, they have been analyzed in Section 15.4.5 of the Final
Safety Analysis Report.
Criticality accidents in the spent fuel pool are not new or
different types of accidents, they have been analyzed in the Final
Safety Analysis Report and in Criticality Analysis reports
associated with specific licensing amendments for fuel enrichments
up to 5.0 weight percent U-235.
Proposed new Technical Specifications 3.7.13 for Unit 1 and
3.7.14 for Unit 2 on the spent fuel pool boron concentration do not
represent new concepts. The boron concentration in the spent fuel
pool has always been maintained near at the limit of the RWST
[refueling water storage tank] boron concentration for refueling
purposes. These new proposed Technical Specifications establish new
boron concentration requirements for the spent fuel pool water
consistent with the results of the revised criticality analysis [ ].
Since soluble boron has always been maintained in the spent fuel
pool water, the implementation of this new requirement will have
little effect on normal pool operations and maintenance. The
implementation of the proposed new limitations on the spent fuel
pool boron concentration will only result in increased sampling to
verify boron concentration. This increased sampling will not create
the possibility of a new or different kind of accident.
Because soluble boron has always been present in the spent fuel
pool, a dilution of the spent fuel pool soluble boron has always
been a possibility. However, it was shown in the spent fuel pool
dilution evaluation [ ] that a dilution of the Farley spent fuel
pool which could reduce the spent fuel storage rack Keff
to less than 0.95 is not a credible event. Therefore, the
implementation of new limitations on the spent fuel pool boron
concentration will not result in the possibility of a new kind of
accident.
Proposed new Technical Specifications 3.7.14 for Unit 1 and
3.7.15 for Unit 2, and 5.6.1.1.e., 5.6.1.1.f, and 5.6.1.1.g. (for
Unit 1) specify the requirements for the spent fuel rack storage
configurations, and do not represent new concepts. These proposed
new spent fuel pool storage configuration limitations are consistent
with the assumptions made in the spent fuel rack criticality
analysis, and will not have any significant effect on normal spent
fuel pool operations and maintenance and will not create any
possibility of a new or different kind of accident. Verifications
will continue to be performed to ensure that the spent fuel pool
loading configuration meets specified requirements.
As discussed above, the proposed changes will not create the
possibility of a new or different kind of accident. There is no
significant change in plant configuration, equipment design or
equipment. The accident analysis in the Final Safety Analysis Report
remains bounding.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The proposed Technical Specification changes and the resulting
spent fuel storage operating limits will provide adequate safety
margin to ensure that the stored fuel assembly array will always
remain subcritical. Those limits are based on a plant specific
criticality analysis [ ] performed in accordance the Westinghouse
spent fuel rack criticality analysis methodology described in [WCAP-
14416-NP-A, ``Westinghouse Spent Fuel Rack Criticality Analysis
Methodology,'' Revision 1, November 1996].
The criticality analysis utilized credit for soluble boron to
ensure Keff will be less than or equal to 0.95 under
normal circumstances, and storage configurations have been defined
using a 95/95 Keff calculation to ensure that the spent
fuel rack Keff will be less than 1.0 with no soluble
boron.
[[Page 45465]]
Soluble boron credit is used to provide safety margin by
maintaining Keff less than or equal to 0.95, including
uncertainties, tolerances, and accident conditions in the presence
of spent fuel pool soluble boron.
The loss of substantial amounts of soluble boron from the spent
fuel pool which could lead to exceeding a Keff of 0.95
has been evaluated [ ] and shown to be not credible.
The evaluations which...show that the dilution of the spent fuel
pool boron concentration from 2000 ppm to 400 ppm is not credible,
combined with the 95/95 calculation, which shows that the spent fuel
rack Keff remain less than 1.0 when flooded with
unborated water, provide a level of safety comparable to the
conservative criticality analysis methodology required by [USNRC
Standard Review Plan for the Review of Safety Analysis Reports for
Nuclear Power Plants, LWR Edition, NUREG-0800, June 1987, USNRC
Spent Fuel Storage Facility Design Bases (for comment) Proposed
Revision 2, 1981, Regulatory Guide 1.13, and ANS, Design
Requirements for Light Water Reactor Spent Fuel Storage Facilities
at Nuclear Power Stations, ANSI/ANS-57.2-1983].
Therefore, the proposed changes in this license amendment will
not result in a significant reduction in the plant's margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document location: Houston-Love Memorial Library, 212
W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: July 11, 1997
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to implement 10 CFR Part 50,
Appendix J, Option B, by referring to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Test Program,'' with four
exceptions as detailed in the licensee's application. Specifically,
changes are requested for TSs 3.7/4.7, STATION CONTAINMENT SYSTEMS,
their associated BASES, and changes to TS Table 4.7.2. Included in the
above changes is a revision to the conservative wording of Surveillance
Requirement (SR) 4.7.A.3 that is being replaced by wording from the
Standard Technical Specifications, and the relocation of this SR to the
Limiting Condition for Operation. The change to TS Table 4.7.2 updates
the information in the Table to the current operational practices, as
approved by an NRC letter dated May 3, 1982. In addition, a description
of Vermont Yankee's Primary Containment Leakage Rate Testing Program
(PCLRTP) will be added to the Administrative Controls Section (6.0) of
the TSs. The testing intervals for the containment system and for the
components that penetrate the primary containment, under Option B of
Appendix J will be performance-based.
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:
Option B
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that contribute to
initiation of any accidents previously evaluated. Thus, the proposed
change cannot increase the probability of any accident previously
evaluated.
The proposed change potentially affects the leak-tight integrity
of the containment structure designed to mitigate the consequences
of a loss-of-coolant accident (LOCA). The function of the
containment is to maintain functional integrity during and following
the peak transient pressures and temperatures which result from any
LOCA. The containment is designed to limit fission product leakage
following the design basis LOCA. Because the proposed change does
not alter the plant design or test method, only the frequency of
measuring Type A, B and C leakage, the proposed change does not
directly result in an increase in containment leakage. However,
decreasing the test frequency can increase the probability that an
increase in containment leakage could go undetected for an extended
period of time. Based upon the results of the periodic containment
Type A or Integrated Leak Rate Tests (ILRTs) and Type B and C or
Local Leak Rate Tests (LLRTs) surveillance tests, this is not
expected during the remaining life of the plant. The risk resulting
from the proposed changes is as follows:
Type A Testing
NUREG/CR-4330 (NRC86) found that the effect of containment
leakage on overall accident risk is small since risk is dominated by
accident sequences that result in failure or bypass of the
containment. It is also determined that on an expected individual
dose basis, the effect of containment leakage is small.
Industry wide, ILRTs have only found a small fraction of the
leaks that exceed current acceptance criteria. Only three percent of
all leaks are detected by ILRTs, and therefore, by extending Type A
testing intervals, only three percent of all leaks have a potential
for remaining undetected for longer periods of time. In addition,
when leakage has been detected by ILRTs, the leakage rate has been
only about two times the allowable leakage rate.
NUREG-1493, ``Performance-Based Containment Leakage Test
Program'', found that these observations, together with the
insensitivity of reactor accident risk to the containment leakage
rate, show that reducing the Type A leakage test frequency would
have a minimal impact on public risk.
Type B and C Testing
NUREG-1493 found that while Type B and C tests can identify the
vast majority (greater than 95 percent) of all potential leakage
paths, performance-based alternatives are feasible without
significant risk impacts. The risk model used in NUREG-1493 suggests
that the number of components tested would be reduced by about 60
percent with less than a three-fold increase in the incremental risk
due to containment leakage. Since, under existing requirements,
leakage contributes less than 0.1 percent of overall accident risk,
the overall impact is very small. NUREG-1493 found that while the
extended testing intervals for Type B and C tests led to minor
increases in potential offsite dose consequences the actual decrease
of on-site (worker) doses would be reduced in proportion to the
number of Type B or C tests not performed.
EPRI Research Project Report TR-104285, ``Risk Impact Assessment
of Revised Containment Leak Rate Testing Intervals,'' also concluded
that a relaxation of the test intervals for Type B and C
penetrations results in a negligible increase in total plant risk.
Based on the above VYNPC [Vermont Yankee Nuclear Power
Corporation] has concluded that the proposed change will not result
in a significant 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 involve a change to the plant
design or operation. As a result, the proposed change does not
affect any of the parameters or conditions that could contribute to
initiation of any accidents. This change involves the reduction in
Type A, B, and C test frequency. The methods of performing the tests
are not changed. No new accident modes are created by extending the
testing intervals. No safety-related equipment or safety functions
are altered as a result of this change. Extending the test frequency
has no influence over nor does it contribute to, the possibility of
a new or different kind of accident or malfunction from those
previously analyzed.
Based upon the above, VYNPC has concluded that the proposed
change will not create the possibility of a new or different kind of
accident from those previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
As stated in the Technical Support Document (TSD) for the NRC's
Option B to
[[Page 45466]]
Appendix J rule change, NUREG-1493 concludes a reduction in the
frequency of Type A testing from the current three per ten years to
one per ten years leads to an imperceptible increase in risk. It
also concludes that a reduction in the frequency of Type B testing
of electrical penetrations should be possible with no adverse impact
on risk. A vast majority of leakage paths are identified by Type C
testing of containment isolation valves and, based on the model of
component failure with time, performance-based alternatives to the
current Type C testing intervals are feasible without significant
risk impacts.
4.7.A.3
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not result in any hardware or operating
procedure changes. Closed and de-activated automatic valves, closed
manual valves or blind flanges that serve as primary containment
isolation valves are not assumed to be initiators of any analyzed
event. The role of these devices is to isolate containment during
analyzed events, thereby limiting consequences. The change
establishes compensatory measures using closed and de-activated
automatic valves, closed manual valves or blind flanges as an
isolation barrier which is equivalent to those already included in
the current Technical Specifications. The proposed change does not
introduce any new failure modes, such that a single active failure
could allow a primary containment release through an un-isolated
path. Therefore, this change will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
This change does not result in any changes to equipment design
or capabilities or the operation of the plant. The change still
ensures the primary containment boundary is maintained. Thus,
this change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Closed and de-activated automatic valves, closed manual valves
or blind flanges which are used to satisfy the compensatory measures
of 4.7.A.3 are primary containment isolation devices will be leak
tested per the PCLRTP. In addition, the Technical Specification
establishes these devices as an isolation barrier that cannot be
adversely affected by a single active failure. As a result, any
reduction in a margin of safety will be insignificant and offset by
the benefit gained with equivalent compensatory measures to ensure
the primary containment boundary is maintained, which reduces
unnecessary plant shutdown transients.
Table 4.7.2 Editorial Change
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
This change updates the information presented in this Table to
reflect current practice. The methods of maintaining an inerted
containment and differential pressure between the drywell and
suppression pool have been previously docketed. The valves to now be
shown normally closed on the Table are large (6'' and 18'') purge
valves and the valves to be shown as normally open to provide makeup
nitrogen are both 1'' in size. The probability of an accident is not
significantly increased, since the subject valves are not considered
to be initiators of any accident previously evaluated. The
consequences of an accident are not significantly increased, since
each of the subject valves receives a close signal from PCIS
[primary containment isolation system]. In addition, PCIS closure of
the two one inch valves will terminate the associated release
pathway more rapidly than the existing valve lineup reflected on the
Table. Thus it is concluded that this change will not involve any
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from amy previously evaluated?
All four valves whose listed normal positions are proposed to be
changed are PCIS valves and receive the same closing signal. All are
tested in accordance with our Appendix J and IST [inservice testing]
programs. No changes in equipment design or operation are proposed,
only the listed normal positions of the subject valves. Thus, this
change will not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The valves to be listed as normally open are significantly
smaller and faster closing than the purge valves currently listed as
open. Thus the change in the listed normal position of these four
valves provides a more conservative initial condition than is
currently depicted in Table 4.7.2. No changes in equipment design or
operation are proposed. Thus, it is concluded that there is no
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.
Local Public Document location: Brooks Memorial Library, 224 Main
Street, Brattleboro, VT 05301
Attorney for licensee: R. K. Gad, III, Ropes and Gray, One
International Place, Boston, MA 02110-2624
NRC Project Director: Ronald B. Eaton, Acting
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: August 14, 1997 (TSCR 199)
Description of amendment request: These amendments would revise: TS
15.4.2.B. ``In-Service Inspection and Testing of Safety Class
Components Other than Steam Generator Tubes,'' to modify item 2 to
change the reference from TS 15.4.4 to the Containment Leakage Rate
Testing Program; TS 15.6.12.A.1, ``Containment Leakage Rate Testing
Program,'' to eliminate the one-time requirement for Unit 2 Type A
testing since the testing has been completed; and TS Bases 15.4.4 to
delete the specific bases for containment purge valve testing and to
delete a reference that is no longer used.
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.
The proposed administrative changes correct discrepancies in the
Technical Specifications introduced as a result of Amendment 169 to
Operating License DPR-24 for Point Beach Nuclear Plant Unit 1 and
Amendment 173 to Operating License DPR-27 for Point Beach Nuclear
Plant Unit 2. These changes correct references to containment
isolation valve testing in the Specifications and Bases. These
amendments were evaluated as acceptable in a safety evaluation dated
October 9, 1996. Therefore, these changes do not result in an
increase in the probability or consequences of any accident
previously evaluated.
The Point Beach Nuclear Plant Unit 2 containment was tested and
found acceptable within the maximum interval defined by a one-time
Technical Specifications requirement. Subsequent testing will be
performed in accordance with the approved testing program defined by
Technical Specifications 15.6.12. Therefore, the Technical
Specification requirements are met. These requirements are
established to ensure the containment performs and is maintained as
designed and assumed in the safety analyses. The removal of the one-
time specific periodicity requirements for the Unit 2, Type A
containment integrated leak rate test does not result in a
significant increase in the probability or consequence of any
accident previously evaluated.
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.
The proposed changes to the Technical Specifications do not
change the requirements for the Point Beach Nuclear Plant
containments to perform as designed and evaluated in the safety
analyses. Test requirements in the Technical Specifications continue
to meet the standards evaluated and approved by the NRC to ensure
the containments continue to perform as
[[Page 45467]]
designed and analyzed. Administrative discrepancies in the
Specifications and bases are also corrected. Therefore, no new or
different kind of accident from any accident previously evaluated is
created.
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 changes to the Technical Specifications ensure
consistency with Amendment 169 to Point Beach Nuclear Plant Unit 1
Operating License DPR-24 and Amendment 173 to Point Beach Nuclear
Plant Unit 2 Operating License DPR-27. Testing of the Unit 2
containment has been performed within the maximum time limit allowed
by the one-time test requirement of Technical Specification 15.6.12.
Testing requirements continue to meet NRC requirements and ensure
the containment continues to operate as designed and analyzed.
Administrative corrections to the Specifications and bases ensure
consistency with previously approved amendments. Therefore, a margin
of safety is not reduced.
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.
Local Public Document location: The Lester Public Library, 1001
Adams Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John N. Hannon
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: July 29, 1997
Description of amendment request: This license amendment request
revises the wording of Action Statement 5.a to Technical Specification
Table 3.3-1. ``Reactor Trip System Instrumentation.'' This action
statement prescribes a set of actions to be accomplished when a source
range neutron detector is inoperable with the plant shut down. The
proposed wording change will clarify the times and order in which these
actions are to be performed.
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.
In MODE 3, 4, or 5 with the rod control system capable of rod
withdrawal or rods not fully inserted, the source range neutron
detectors provide a reactor trip signal on high neutron flux to
provide core protection against an uncontrolled rod cluster control
assembly bank withdrawal from a subcritical or low power startup
condition. This trip function is actuated when either of two
independent source range channels indicates a neutron flux level
above a preselected manually adjustable setpoint. If the
rod control system is not capable of rod withdrawal with rods
fully inserted, the source range detectors are not required to trip
the reactor.
NUREG-1431, Revision 1, ``Standard Technical Specifications
Westinghouse Plants,'' allows one source range neutron detector to
be out of service for up to 48 hours. One additional hour is allowed
to open the reactor trip breakers and suspend operations involving
the addition of positive reactivity. This was the same action
sequence prescribed for the source range neutron detectors prior to
the implementation of Amendment No. 96 to the Wolf Creek Technical
Specifications, which inadvertently resulted in an ambiguous
rewording of the action. The proposed rewording of the action
statement clarifies the proper timing of the required actions, and
is consistent with NUREG-1431, Revision 1.
The proposed change does not introduce any new potential
accident initiating conditions and does not alter any plant
operating procedures or method of operation of any plant components
or systems. Allowing positive reactivity changes during the 48 hour
period in which one source range neutron detector is inoperable is
acceptable since the remaining detector will still provide the
reactor trip function and control room indication when the reactor
trip breakers are closed, and control room indication
when the reactor trip breakers are open. This is consistent with
the provisions in NUREG-1431, Revision 1. Thus, the proposed change
does not affect any system's ability to mitigate the consequences of
an accident and will not increase the probability of occurrence of
any previously evaluated accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not affect the method of operation of
any plant component or system, and does not create any new, or alter
any existing, accident initiators. The proposed change clarifies
that positive reactivity changes may be allowed during the 48 hour
period in which a source range neutron detector is inoperable, as
provided for in NUREG-1431, Revision 1. This action does not affect
the capability of the remaining source range neutron detector to
provide a reactor trip signal on high neutron flux during this
period when the reactor trip breakers are closed, nor does it affect
the ability of the remaining detector of providing control room
indication. This function of the source range neutron detectors is
discussed in Chapter 15 of the Wolf Creek Updated Safety Analysis
Report. This proposed change does not modify any existing plant
equipment, add any new plant equipment, or alter any component or
system operating parameters or procedures. Therefore, this proposed
change will
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 source range neutron detectors provide a reactor trip
function during shutdown conditions when the reactor trip breakers
are closed. When the reactor trip breakers are open they provide
control room alarm/indication, only. The proposed change clarifies
that positive reactivity changes may be allowed during the 48 hour
period in which a source range neutron detector is inoperable. This
is consistent with the provisions in NUREG-1431, Revision 1 and with
Wolf Creek Technical Specification Table 3.3-1, Action 5.a, prior to
the implementation of Amendment No. 96. In Amendment No. 96 the
wording of this action was changed such that this allowance was no
longer clear. With one source range neutron detector inoperable with
the reactor trip breakers closed, the reactor trip on high neutron
flux function is still provided by the remaining source range
neutron detector. With one source range neutron detector inoperable
with the reactor trip breakers open, control room indication of high
neutron flux is still provided. As stated above, this is consistent
with NUREG-1431, Revision 1, as well as with the action requirements
prior to the implementation of Amendment No. 96. This proposed
change, then, does not affect the margin of safety provided by the
source range neutron detectors.
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.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the
[[Page 45468]]
same as above. They were published as individual notices either because
time did not allow the Commission to wait for this biweekly notice or
because the action involved exigent circumstances. They are repeated
here because the biweekly notice lists all amendments issued or
proposed to be issued involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
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: July 25, 1997
Brief description of amendments: The proposed amendments would
modify Technical Specification (TS) 4.0.5.f in a manner that would
allow exceptions to the NRC staff's positions on intergranular stress
corrosion cracking in boiling water reactor austenitic stainless steel
piping, where specific written relief has been granted by the NRC. TS
4.0.5.f now requires that the Brunswick Steam Electric Plant, Units 1
and 2, Inservice Inspection program be performed in accordance with the
positions identified in NRC Generic Letter 88-01. Date of publication
of individual notice in Federal Register: August 12, 1997 (62 FR 43187)
Expiration date of individual notice: September 11, 1997
Local Public Document location: University of North Carolina at
Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida
Date of application for amendment: August 4, 1997
Brief description of amendment: The proposed amendment would revise
the Technical Specifications to extend the frequency for certain
surveillances related to the emergency diesel generators. Date of
publication of individual notice in the FEDERAL REGISTER:August 12,
1997 (62 FR 43189)
Expiration date of individual notice: September 11, 1997
Local Public Document location: Coastal Region Library, 8619 W.
Crystal Street, Crystal River, Florida 32629
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: August 6, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification Table 2.2-1 and 3/4.2.5 to allow the
reactor coolant system total flow to be determined using cold leg elbow
tap differential pressure measurements. Date of individual notice in
the Federal Register: August 14, 1997 (62 FR 43556)
Expiration date of individual notice: September 15, 1997
Local Public Document location: Wharton County Junior College, J.
M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Commonwealth Edison Company, Docket No. 50-455, Byron Station, Unit
No. 2, Ogle County, Illinois, Docket No. STN 50-457, Braidwood
Station, Unit No. 2, Will County, Illinois
Date of application for amendments: May 24, 1997, as supplemented
by letters dated May 31, June 20 and June 24, 1997
Brief description of amendments: The amendments revise Technical
Specification 4.5.2.b.1 to include the use of Ultrasonic Testing (UT)
to verify that the emergency core cooling system (ECCS) is completely
filled with water. For the ECCS subsystem with high point vent valves
in direct communication with the operation system, UT is acceptable in
lieu of physically opening the vents.
Date of issuance: August 13, 1997
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 91 and 84
Facility Operating License Nos. NPF-66 and NPF-77: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 10, 1997 (62 FR
31633) The May 31, June 20, June 24, and July 18, 1997, submittals
provided additional clarifying information that did not change the
proposed initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated August 13, 1997. No significant hazards
consideration comments received: No
Local Public Document location: For Byron, the Byron Public Library
District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: June 9, 1997
Brief description of amendments: The amendments authorize a change
to the realistic dose values for the process gas system rupture in
Section 15.0 of the
[[Page 45469]]
Byron/Braidwood (B/B) Updated Final Safety Analysis Report (UFSAR).
During preparation of a UFSAR change package, ComEd discovered that the
Final Safety Analysis Report (FSAR) had not been updated to correct an
error from the previous revision of the dose calculation. Since the
correct dose value is greater than that previously reported, the
consequences of the accident had increased, and an unreviewed safety
question resulted.
Date of issuance: August 13, 1997
Effective date: August 13, 1997
Amendment Nos.: 92, 92, 85, 85
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments authorize a change to the Byron/Braidwood UFSAR.
Date of initial notice in Federal Register: July 10, 1997 (62 FR
37079). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 13, 1997. No significant
hazards consideration comments received: No
Local Public Document location: For Byron, the Byron Public Library
District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481
Consumers Energy Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of application for amendment: April 30, 1997
Brief description of amendment: The amendment revises the Big Rock
Point Plant license and technical specifications to reflect the
licensee's name change from ``Consumers Power Company'' to ``Consumers
Energy Company.''
Date of issuance: August 14, 1997
Effective date: August 14, 1997
Amendment No.: 119
Facility Operating License No. DPR-6: Amendment revised the license
and the Technical Specifications.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30630) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 14, 1997. No significant
hazards consideration comments received: No.
Local Public Document location: North Central Michigan College,
1515 Howard Street, Petoskey, Michigan 49770
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: May 8, 1997, as supplemented
June 10, and July 25, 1997
Brief description of amendment: The amendment incorporates
additional NRC-approved topical reports into the Technical
Specifications (TS).
Date of issuance: August 12, 1997
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 202
Facility Operating License No. DPR-50: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 4, 1997 (62 FR
30633) The June 10 and July 25, 1997, letters provided clarifying
information that did not change the scope of the May 8, 1997,
application or the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated August 12, 1997. No
significant hazards consideration comments received: No
Local Public Document location: Law/Government Publications
Section, State Library of Pennsylvania (REGIONAL DEPOSITORY), Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
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: February 29, 1996
(AEP:NRC:1232), and supplemented November 15, 1996 (AEP:NRC:1232A), and
February 4, 1997 (AEP:NRC:1232B)
Brief description of amendments: The amendments revise the
Technical Specifications and associated bases to increase the minimum
borated water volume in the boric acid storage system and decrease the
required boron concentration.
Date of issuance: August 7, 1997
Effective date: August 7, 1997, with full implementation when the
required plant modifications are completed, but not later than August
31, 1998.
Amendment Nos.: 216 and 200
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 24, 1996 (61 FR
18172) The November 15, 1996, and February 4, 1997, supplements only
provided the schedule for the plant modifications and procedure changes
associated with this amendment and did not change the staff's proposed
determination of no significant hazards consideration. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated August 7, 1997.No significant hazards consideration
comments received: No.
Local Public Document location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
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: December 20, 1996
Brief description of amendments: The amendments reduce the
frequency and scope of reactor coolant pump flywheel inspections.
Date of issuance: August 8, 1997
Effective date: August 8, 1997, with full implementation within 45
days.
Amendment Nos.: 217 and 201
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 18, 1997 (62 FR
33126) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 8, 1997. No significant
hazards consideration comments received: No.
Local Public Document location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of application for amendment: September 13, 1996, as
supplemented by letter dated September 25, 1996
Brief description of amendment: The amendment revised Technical
Specification 5.5.B to designate the President, Maine Yankee as the
responsible official for matters related to the Nuclear Safety Audit
and Review (NSAR) Committee. The amendment includes some minor
editorial changes to the same technical specification.
Date of issuance: August 8, 1997
Effective date: August 8, 1997, to be implemented within 30 days of
the date of issuance.
Amendment No.: 159
Facility Operating License No. DPR-36: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 6, 1996 (61 FR
[[Page 45470]]
57487) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 8, 1997. No significant
hazards consideration comments received: No.
Local Public Document location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: June 13, 1997
Brief description of amendment: The amendment modifies Technical
Specification (TS) Surveillance Requirement 4.4.1.3.3 to be consistent
with the requirements of TS 3.4.1.3. Specifically, the change brings TS
4.4.1.3.3 into agreement with TS 3.4.1.3 by requiring that the
specified reactor coolant and/or residual heat removal system loops be
verified in operation and circulating reactor coolant at least once per
12 hours during Mode 4.
Date of issuance: August 5, 1997
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 145
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 2, 1997 (62 FR
35850) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 5, 1997. No significant
hazards consideration comments received: No.
Local Public Document location: Learning Resources Center, Three
Rivers Community-Technical College, 574 New London Turnpike, Norwich,
Connecticut 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, Connecticut 06385
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: January 27, 1997, as
supplemented May 16, 1997
Brief description of amendment: The amendment changes the Technical
Specifications to permit control rod misalignment of up to plus or
minus 18 steps when the core thermal power is less than 85% of rated
power.
Date of issuance: August 11, 1997
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 176
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 19, 1997 (62 FR
33445) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 11, 1997. No significant
hazards consideration comments received: No
Local Public Document location: White Plains Public Library, 100
Martine Avenue, White Plains, New York 10610
South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: March 26, 1997
Brief description of amendment: The amendment revises TS 4.5.2.a
for the two charging/high head safety injection (HHSI) pump cross
connect valves (XVG-8133A and XVG-8133B) and charging pump mini-flow
header isolation valve (XVG-8106) in the emergency core cooling system
(ECCS). The proposed amendment adds these valves to the list of valves
in TS Surveillance Requirement 4.5.2.a on page 3/4 5-4, consequently
these valves will be verified once every 12 hours to indicate that they
are in the required position with power to the valve operators removed.
Date of issuance: August 8, 1997
Effective date: August 8, 1997
Amendment No.: 136
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 21, 1997 (62 FR
27801) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 8, 1997. No significant
hazards consideration comments received: No
Local Public Document location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: November 14, 1995, as
supplemented July 11, 1996 and July 24, 1997
Brief description of amendment: The amendment revises Technical
Specification 3/4.8.4.2 for motor-operated valves thermal overload
protection and bypass devices at Virgil C. Summer Nuclear Station.
Date of issuance: August 13, 1997
Effective date: August 13, 1997
Amendment No.: 137
Facility Operating License No. NPF-12: Amendment adds a new License
Condition and revises the Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65684) The July 11, 1996, and July 24, 1997 submittals contained
clarifying information only and did not change the proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated August 13, 1997. No significant hazards consideration comments
received: No
Local Public Document location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
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: September 26, 1996, as supplemented on
August 12, 1997 (TS 96-04)
Brief description of amendments: The amendments change the
Technical Specifications (TS) by relocating the fire protection program
details to the Updated Final Safety Analysis Report and Fire Protection
Plan in accordance with Generic Letters 86-10 and 88-12.
Date of issuance: August 12, 1996
Effective date: August 12, 1996
Amendment Nos.: 227 and 218
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TS.
Date of initial notice in Federal Register: July 2, 1997 (62 FR
35843) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 12, 1997. No significant
hazards consideration comments received: No
Local Public Document location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: August 22, 1996, as revised
July 14, 1997
Brief description of amendments: These amendments revise Section
3.A of Facility Operating Licenses DPR-24 and
[[Page 45471]]
DPR-27 from a licensed power level of 1518 megawatts thermal to 1518.5
megawatts thermal. A similar revision is made in the bases of Technical
Specification 15.3.1.B, ``Pressure/Temperature Limits.''
Date of issuance: August 6, 1997
Effective date: August 6, 1997
Amendment Nos.: 175 and 179
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the licenses.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52972) The July 14, 1997, supplement provided a corrected bases page
and did not affect the staff's no significant hazards considerations
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 6, 1997. No significant
hazards consideration comments received: No.
Local Public Document location: The Lester Public Library, 1001
Adams Street, Two Rivers, Wisconsin 54241
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: February 12, 1997, as
supplemented on March 11, 1997 (TSCR 196)
Brief description of amendments: These amendments revise Point
Beach Nuclear Plant's (PBNP) Technical Specifications (TSs) to relocate
turbine overspeed protection specifications, limiting conditions for
operation, surveillance requirements, and associated bases from TS
Section 15.3.4, ``Steam and Power Conversion System,'' and Section
15.4.1, ``Operational Safety Review,'' to the Final Safety Analysis
Report (FSAR) in accordance with Generic Letter 95-10.
Date of issuance: August 6, 1997
Effective date: These license amendments are effective as of the
date of issuance and shall be implemented by incorporating the turbine
overspeed protection specifications, limiting conditions for operation,
surveillance requirements, and associated bases into the FSAR by June
30, 1998.
Amendment Nos.: 176 and 180
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 23, 1997 (62 FR
19838) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 6, 1997. No significant
hazards consideration comments received: No.
Local Public Document location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Dated at Rockville, Maryland this 20th day of August 1997.
For the Nuclear Regulatory Commission
John A. Zwolinski,
Acting Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation.
[Doc. 97-22635 Filed 8-26-97; 8:45 am]
BILLING CODE 7590-01-F
This is a copy of a public record, reproduced as it was published. It is not legal advice, and it may not be the version a court would rely on. Check the official source before you cite it.