[Federal Register Volume 59, Number 177 (Wednesday, September 14, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10914]
[[Page Unknown]]
[Federal Register: September 14, 1994]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating LicensesInvolving
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 22, 1994 through September 1, 1994.
The last biweekly notice was published on August 31, 1994 (59 FR
45015).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By October 14, 1994, 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 20555 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, 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 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
room for the particular facility involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: August 2, 1994
Description of amendments request: The proposed amendment would
revise Technical Specifications (TSs) 3.9.1 and 3.1.2.7 and the Bases
to Specification 3.1.2.7. Specifically, TS 3.9.1, ``Refueling
Operations, Boron Concentration,'' would be revised to require action
to restore boron concentration to within its limits in place of the
current requirement to initiate and continue boration at a rate greater
than or equal to 40 gpm of 2300 ppm boric acid solution or its
equivalent until the boron concentration is within its limit. TS
3.1.2.7, ``Borated Water Sources - Shutdown,'' gives the operability
requirement for borated water sources including the Refueling Water
Tank (RWT), in Modes 5 and 6. The minimum boron concentration is given
as 2300 ppm. While this minimum value is correct for Mode 5, a larger
boron concentration may be necessary in Mode 6. The RWT is the
preferred borated water source for restoring the required boron
concentration as required by TS 3.9.1. Therefore, the RWT boron
concentration in Mode 6 should be at least be that required by TS
3.9.1. The proposed change to TS 3.1.2.7 would clarify the boron
concentration requirements. In Mode 5, 2300 ppm will continue to be
required. In Mode 6, the boron concentration limit for the RWT will be
the boron concentration limits given in TS 3.9.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
During refueling operations, the reactivity condition of the
core is maintained consistent with the initial conditions assumed
for the boron dilution event in the accident analysis (Updated Final
Safety Analysis Report Section 14.3) and is sufficient to ensure the
core remains subcritical during core alterations. Technical
Specification 3.9.1 requires that the boron concentration be
maintained to ensure a keff [is less than or equal to] 0.95.
Should the boron concentration drop below the Technical
Specifications limit, the Action requires boration at a specified
flow rate and boron concentration until the boron concentration is
restored to within its limit. Refueling boron concentrations higher
than the concentration specified by the Action in [Technical]
Specification 3.9.1 are allowed by the Technical Specifications and
clarification of the Action for that circumstance is needed. The
proposed change eliminates the specified flow rate and boron
concentration in the Action and substitutes a directive to
immediately initiate action to restore the boron concentration to
within its limits. The accident analysis does not assume a specific
boration rate, but only assumes that the operator acts to terminate
the dilution.
Therefore, the consequences of the event are unchanged. In
addition, the proposed change revises the boron concentration limit
on the Refueling Water Tank in Mode 6 to make the boron
concentration limit on the tank the same as the boron concentration
limit on the reactor coolant system. This will ensure that the RWT
will contain water of a sufficient boron concentration to respond to
a boron dilution event.
The proposed change does not change the boron concentration or
shutdown margin required by [Technical] Specification 3.9.1 and
continues to meet the initial conditions of the boron dilution
event. Therefore, the probability of a boron dilution event is not
increased. Furthermore, the revised action ensures that the
appropriate actions for a boron dilution event will be taken and
that a borated water source of sufficient concentration is available
to respond to that event. Therefore, the consequences of a boron
dilution event are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The proposed change does not represent a significant change in
the configuration or operation of the plant. The proposed actions
will results in the same operator actions as the current Technical
Specifications. The minimum boron concentration of the Refueling
Water Tank in Mode 6 may be increased above the current value, but
the concentrations will be within the analyzed maximum concentration
for that tank,
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The margin of safety provided by [Technical] Specification 3.9.1
is to ensure that the core remains subcritical during a boron
dilution event and during core alterations. The proposed change does
not alter the required shutdown margin or significantly change the
actions to be taken if that shutdown margin is lost. The proposed
change ensures that all assumed borated water sources will have
sufficient boron concentration to respond to boron dilution event.
Therefore, the proposed change does not involve a significant
reducation 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
amendments request involves no significant hazards consideration.
Local Public Document Room: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Michael J. Case
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: August 2, 1994
Description of amendments request: The proposed change would
revise Technical Specifications (TSs) regarding surveillances
associated with the Emergency Diesel Generators (EDGs). Specifically,
TS 4.8.1.1.2.d.3.c would be revised to add high crankcase pressure to
the EDG trips which are verified to be automatically bypassed on a
Safety Injection Actuation Signal (SIAS). In addition, a footnote would
be added stating that verification of the high crankcase pressure trip
bypass will not be required on a particular EDG until the modification
has been completed for that EDG.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The Calvert Cliffs Emergency Diesel Generators (EDGs) are used
to provide electrical power for the operation of Engineered Safety
Features (ESF) and safe shutdown equipment for events involving a
loss of offsite power. The EDGs are also called upon to
automatically start if an accident condition (SIAS) is present. In
the event of an automatic start from a SIAS, the EDGs do not assume
any load until the preferred, offsite power source is actually lost.
On an undervoltage condition on a vital bus, the corresponding EDGs
automatically start and load.
Emergency diesel generator trips are provided to initiate engine
shutdown during abnormal diesel-run conditions, thereby protecting
the EDGs from any resulting damage. Under emergency conditions, EDG
reliability is a key accident-mitigating factor; therefore, upon
receipt of a SIAS, the EDG control logic blocks two of the normal
shutdown signals so that the only signals remaining are those
required to prevent rapid destruction of the diesel engine. High
crankcase pressure is typically not an indication of impending rapid
diesel engine failure; therefore, this trip will be added to those
shutdown signals bypassed on a SIAS. The proposed Technical
Specification change adds the high crankcase pressure trip as one of
the EDG trips verified to be bypassed by a SIAS. A high crankcase
pressure condition on one EDG will not impact either of the two
unaffected EDGs, or any other equipment required to mitigate
accident consequences, and satisfies the single failure criteria.
The manufacturer concurs with the proposed change to bypass this
trip on a SIAS. In blocking this trip on a SIAS, the ultimate effect
is an increase in the reliability of the effected EDG, and
therefore, no increase in the consequences of a previously evaluated
accident.
Additionally, the EDGs are not initiators to any previously
evaluated accident. Therefore, blocking the high crankcase pressure
trip on a SIAS will not increase the probability of an accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase to the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The function of the EDGs is to provide power to ESF and safe
shutdown equipment for events involving a loss of offsite power. The
proposed change does not represent a significant change in the
configuration or operation of the plant; therefore, the EDGs
continue to function in an accident mitigation role. The EDGs are
not accident precursors, either in the current configuration, or
following the modification to block the high crankcase pressure
trip.
Therefore, the proposed changes do not create the possibility of
a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The margin of safety credited with the EDG function associated
with this change is the reliability of the EDGs following an event
involving a loss of offsite power. By blocking high crankcase
pressure trips on a SIAS, this change increases the likelihood that
an EDG will be able to supply power when it is needed most, during a
SIAS, because the probability of an unnecessary EDG shutdown is
decreased. In effect, the margin of safety associated with this
function, EDG reliability, is increased.
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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Michael J. Case
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: August 4, 1994
Description of amendments request: The proposed amendment would
eliminate Technical Specifications 3/4.3.3.3, 6.9.2.b, and 6.9.2.d and
Bases 3/4.3.3.3 which gives requirements for seismic monitoring
instrumentation. Specifically, the requirements for operation and
testing of the seismic monitoring instrumentation would be relocated to
the Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis
Report (UFSAR) and plant procedures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change has been evaluated against the standards in
10 CFR 50.92 and has been determined to not involve a significant
hazards consideration, in that operation of the facility in
accordance with the proposed amendments:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The seismic monitoring system is used to measure the seismic
response of selected Class 1 structures, provide time-history
records of seismic events, and would indicate if predetermined
seismic acceleration values had been exceeded. The seismic
monitoring system itself has no safety function. The system measures
values which are used after the fact to assess the intensity of an
earthquake.
The proposed change will relocate requirements regarding the
operability and testing of the seismic monitors from the Technical
Specifications to the UFSAR and plant procedures. This will allow
changes to the requirements to be made without Commission approval
as long as the changes meet the criteria of 10 CFR 50.59. Associated
Technical Specification Special Report requirements and Bases will
be deleted. Changes to the seismic monitoring system requirements
which do not meet the criteria of 10 CFR 50.59 must be approved by
the Commission by license amendment.
The seismic monitoring system is not an initiator and does not
act to minimize the consequences of any accident previously
evaluated. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated?
The proposed relocation of seismic monitor requirements from the
Technical Specifications to the UFSAR and plant procedures does not
represent a change in the configuration or operation of the plant.
The seismic monitoring system will continue to be controlled under
10 CFR 50.59. Associated Technical Specification Special Report
requirements and Bases will be deleted. The proposed change will not
add any new hardware and will not introduce any new accident
initiators. Therefore, the proposed change does not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Does operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
The seismic monitoring system is used to measure the response of
selected Class 1 structures to seismic events. The plant is designed
to withstand the loads imposed by the maximum hypothetical accident
and the design seismic disturbance without loss of functions
required for reactor shutdown and emergency core cooling. As a
consequence, the seismic monitoring system makes no contribution to
the margin of safety, and neither do the associated special reports.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Michael J. Case
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: July 18, 1994
Description of amendment request: The purpose of the proposed
amendment is to separate the Technical Specification (TS) into two
separate volumes, one volume explicitly for Unit 1 and one volume
explicitly for Unit 2. At present, each unit has a single volume of TS
which contains the specifications covering both units. In anticipation
of the steam generator (SG) replacement project scheduled to begin in
the fall of 1994, the licensee is requesting that the TS reflect unit
specific data. Since the SG project outlines a schedule for single
units, the present documentation reflecting both units in one volume
will make it difficult to facilitate TS changes to a single unit. The
proposed TS will modify the current situation as follows:1) The pages
will now contain the same information as found before with the
exception of references to different units. The Unit 1 volume will only
contain parameter and setpoint values applicable to Unit 1; the Unit 2
volume will only contain information applicable to Unit 2.2) The limits
established by the TS (the definitions, the limiting conditions for
operation, the surveillance requirements, the Bases, etc.) will be
unchanged by this amendment, with the exception of (3) below. The
effect of the amendment will be that the Unit 1 TS will be found only
in the volume dedicated solely to Unit 1 and likewise for Unit 2. 3) TS
Sections 3.0.5 and 4.0.6 will be deleted and minor editorial changes,
such as the correction of misspellings and the deletion of obsolete
footnotes, will be made. TS 3.0.5 and 4.0.6 define the applicability of
the current joint TS volume to each unit individually. Since each
unit's TS will be located in a separate volume, no statements are
necessary to indicate differences in parameters between units and TS
3.0.5 and 4.0.6 may be deleted.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed amendments would not involve a significant increase
in the probability or consequences of a previously evaluated
accident. The separation of the existing technical specification
manual into unit-specific volumes is a strictly administrative
process which will not affect the probability or consequence of any
accident.
They will not create the possibility of a new or different kind
of accident from any accident previously evaluated. The changes do
not have any impact upon the design or operation of plant equipment;
therefore, they cannot serve to initiate a new type of accident.
The proposed amendments would not involve a reduction in a
margin of safety. The changes would not impact the design or
operation of any plant systems or components.
Based upon the preceding analysis, Duke Power Company concludes
that the proposed amendments do not involve a significant hazards
consideration as defined by 10 CFR 50.92.
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: York County Library, 138 East Black
Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: July 18, 1994
Description of amendment request: The purpose of the proposed
amendment is to separate the Technical Specifications (TS) into two
separate volumes, one volume explicitly for Unit 1 and one volume
explicitly for Unit 2. At present, each unit has a single volume of TS
which contains the specifications covering both units. In anticipation
of the steam generator (SG) replacement project scheduled to begin in
the fall of 1994, the licensee is requesting that the TS reflect unit
specific data. Since the SG project schedules SG replacement for each
unit at different times, the present common TS would make it difficult
to facilitate TS changes to a single unit. The proposed amendment will
modify the current TS as follows:1) The pages will now contain the same
information as found before with the exception of references to
different units. The Unit 1 volume will only contain parameter and
setpoint values applicable to Unit 1; the Unit 2 volume will only
contain information applicable to Unit 2.2) The limits established by
the TS (the definitions, the limiting conditions for operation, the
surveillance requirements, the Bases, etc.) will be unchanged by this
amendment, with the exception of (3) below. The effect of the amendment
will be that the Unit 1 TS will be found only in the volume dedicated
solely to Unit 1 and likewise for Unit 2.3) TS Sections 3.0.5 and 4.0.6
will be deleted and minor editorial changes, such as the correction of
misspellings and the deletion of obsolete footnotes, will be made. TS
3.0.5 and 4.0.6 define the applicability of the current joint TS volume
to each unit individually. Since each unit's TS will be located in a
separate volume, no statements are necessary to indicate differences in
parameters between units and TS 3.0.5 and 4.0.6 may be deleted.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed amendments would not involve a significant increase
in the probability or consequences of a previously evaluated
accident. The separation of the existing technical specification
manual into unit-specific volumes is a strictly administrative
process which will not affect the probability or consequence of any
accident.
They will not create the possibility of a new or different kind
of accident from any accident previously evaluated. The changes do
not have any impact upon the design or operation of plant equipment;
therefore, they cannot serve to initiate a new type of accident.
The proposed amendments would not involve a reduction in a
margin of safety. The changes would not impact the design or
operation of any plant systems or components.
Based upon the preceding analysis, Duke Power Company concludes
that the proposed amendments do not involve a significant hazards
consideration as defined by 10 CFR 50.92.
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: Atkins Library, University of North
Carolina, Charlotte (UNCC Station), North Carolina 28223
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi
Date of amendment request: June 17, 1994, as supplemented by letter
dated August 17, 1994.
Description of amendment request: The amendment requests the
removal of license conditions for Transamerica Delaval (TDI) Emergency
Diesel Generators (EDGs) associated with NUREG-1216.
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 the
consequences of an accident previously evaluated:
The proposed amendment would not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Elimination of the required teardowns and inspections has
no effect on the probability of an accident occurring, because the
diesel generators are not accident initiating equipment. Also,
deleting the teardowns and inspections would decrease the
consequences of an accident because the availability of the engines
would increase as a result of the less frequent teardowns.
Additionally, the high average reliability of the TDI engines would
not be negatively affected due to this change. NRC research has
shown there is a period of decreased reliability immediately
following intrusive teardowns, (break in period), followed by a long
period of high reliability.
2. Create the possibility of a new or different kind of accident
from any previously evaluated:
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 cause any physical change
to the plant or the design or operation of the diesel units.
3. Involve a significant decrease in the margin of safety.
The proposed amendment would not involve a significant reduction
in a margin of safety. The proposed amendment will increase the
reliability and availability of the EDGs and therefore will not
result in a decrease in a margin of safety at Grand Gulf.
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: Judge George W. Armstrong Library, Post
Office Box 1406, S. Commerce at Washington, Natchez, Mississippi 39120
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 9, 1994
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) by relocating the functions
under review and audit to the Waterford 3 quality assurance program
manual. The proposed change also incorporates the TS line-item-
improvement of Generic Letter 93-07, ``Modification Of The Technical
Specification Administrative Control Requirements For Emergency And
Security Plans,'' dated December 28, 1993. The changes are proposed to
reduce regulatory burden by relocating TS requirements that are
duplicated by other regulatory requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change will have no affect on design bases
accidents nor will the change directly affect any material condition
of the plant that could directly contribute to causing or mitigating
the effects of an accident. Relocating Review and Audit functions
from the TS is consistent with the NRC Final Policy Statement on
Technical Specifications Improvements and will have no negative
impact on plant operation or safety. Therefore, the proposed change
will not involve a significant increase in the probability or
consequences of any accident previously evaluated.
The proposed change will not alter the operation of the plant or
the manner in which the plant is operated. The change will not
involve a design change or introduce any new failure modes.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change is administrative in nature. The Waterford 3
safety margins are defined and maintained by the Technical
Specifications in Sections 2-5 which are unaffected. Therefore, the
proposed change will not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room: University of New Orleans Library,
Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of amendment request: August 16, 1994
Description of amendment request: The proposed changes revise VEGP
Technical Specification 3/4.7.1.1 and its bases regarding the setpoint
tolerance for the Main Steam Safety Valves (MSSVs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The setpoint tolerance change for the MSSVs from plus or minus 1% to
+2%, -3% is intended to accommodate setpoint drift that may occur
with these valves during plant operation. However, this change will
not adversely affect the pressure boundary integrity or safety
function of the valves. The increase in MSSV setpoint tolerance was
also reviewed with respect to the accident analyses presented in the
VEGP Final Safety Analysis Report (FSAR). The evaluation
demonstrated that the acceptance criteria of the accident analyses
continued to be met. Additionally, the radiological consequences
associated with the accident analysis are unaffected by the proposed
changes. Accordingly, since the performance and capability of the
MSSVs will be maintained as a result of the proposed changes with no
increase in radiological consequences, there will be no 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. The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment, and no
new failure modes have been defined for any plant system or
component. The design basis requirement for the MSSVs will continue
to be met and the structural integrity of the valves will not be
challenged. Also, the setpoint tolerance change will not adversely
affect the capability of the MSSVs to perform their pressure relief
function to ensure the secondary side steam design pressure is not
exceeded. Additionally, the as-left lift setpoints following testing
of the MSSVs will continue to be within plus or minus 1% of their
lift settings, further ensuring their safety function capability.
Therefore, since the function of the MSSVs is unaffected by the
proposed changes, the possibility of a new or different kind of
accident from any accident previously evaluated is not created.
3. The proposed changes do not involve a significant reduction
in a margin of safety. All applicable acceptance criteria associated
with increasing the MSSV setpoint tolerance will continue to be met.
This includes the structural integrity of the valves and the effect
of the setpoint change on the accident analyses presented in the
VEGP FSAR. Therefore, since the MSSVs remain in compliance with the
appropriate codes and standards and all applicable acceptance
criteria continue to be met, there will not be a significant
reduction in a margin of safety.
Based on the preceding analysis, Georgia Power Company has
determined that the proposed changes to the VEGP Technical
Specifications will not significantly increase the probability or
consequences of an accident previously evaluated, create the
possibility of a new or different kind of accident than any
previously evaluated, or involve a significant reduction in a margin
of safety. Therefore, the proposed changes meet the requirements of
10 CFR 50.92(c) and do 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
Local Public Document Room: Burke County Public Library, 412 Fourth
Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Herbert N. Berkow
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: August 19, 1994
Description of amendment request: The amendment updates and
clarifies the surveillance requirements for control rod exercising and
standby liquid control pump operability testing including the bases to
be consistent with Generic Letter 93-05.
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. Although the surveillance requirements are lessened by these
proposed changes, the changes are consistent with those found
acceptable by the NRC in GL 93-05. The proposed changes have been
determined to be compatible with our plant operating experience.
Based on these considerations, it is concluded that the changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed changes do not involve physical changes to the
plant or changes in plant operating configuration. The changes only
involve frequency of testing required to be performed. The changes
are consistent with those found acceptable by the NRC in GL 93-05.
Thus, it is concluded that the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Although the surveillance requirements are lessened by these
proposed changes, the changes are consistent with those found
acceptable by the NRC in GL 93-05. The proposed changes have been
determined to be compatible with our plant operating experience.
Based on these considerations, it is concluded that 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 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: Ocean County Library, Reference
Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: August 15, 1994
Description of amendment request: The proposed amendment would
increase the allowable main steam isolation valve (MSIV) leakage and
delete the Technical Specifications requirements applicable to the MSIV
leakage control system.
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:
Description of Amendment Request:
Proposed Change 1
This proposed change increases the allowable leak rate specified
in Technical Specification (TS) 4.7.A.2.c.3 from 11.5 standard cubic
feet per hour (scfh) for any one main steam isolation valve (MSIV)
when tested at 24 psig to 100 scfh for any one MSIV with a total
maximum pathway leakage rate of 200 scfh through all four main steam
lines when tested at 24 psig. If an MSIV exceeds 100 scfh, it will
be restored to less than or equal to 11.5 scfh.
Basis for proposed no significant hazards consideration
determination:
1. The change does not involve a significant increase in the
probability or consequences of an accident previously evaluated. The
proposed amendment does not involve a change to structures,
components, or systems which would affect the probability of an
accident previously evaluated in the DAEC Updated Final Safety
Analysis Report (UFSAR). It results in acceptable radiological
consequences for the design basis loss of coolant accident (LOCA)
which was previously evaluated in the UFSAR.
Plant specific radiological analyses have been performed to
assess the effects of the proposed increase in the allowable MSIV
leak rate in terms of control room, technical support center (TSC),
and offsite doses following a postulated design basis LOCA. These
analyses utilize the hold-up volumes of the main steam piping and
condenser as an alternate method for treating MSIV leakage. The
radiological analyses use standard conservative assumptions for the
release of source terms consistent with Regulatory Guide 1.3,
``Assumptions Used for Evaluating the Potential Radiological
Consequences of a Loss of Coolant Accident for Boiling Water
Reactors,'' Revision 2, dated June 1974.
Dose contributions from the proposed MSIV leakage rate limit of
100 scfh per MSIV (with a maximum pathway leakage rate not to exceed
200 scfh through all four main steam lines) were calculated. The
analysis demonstrated that the dose contributions from the proposed
MSIV leakage rate resulted in an acceptable increase to the LOCA
doses previously evaluated against the regulatory limits for the
offsite, control room, and TSC doses as contained in 10 CFR 100 and
10 CFR 50, Appendix A (General Design Criterion 19). The revised
LOCA doses are the LOCA doses previously evaluated in the UFSAR plus
the MSIV leakage doses calculated assuming use of the alternate
treatment method. Table 1 of Attachment 2 shows the previously
calculated doses and the newly calculated doses.
It is important to note that the resulting doses are dominated
by the organic iodine fractions which occur because of the
conservative source term assumptions used in this analysis. For a
total leakage rate of 200 scfh through all four main steam lines,
more than 90 percent of the offsite, control room, and TSC iodine
doses are due to the organic iodine from the Regulatory Guide 1.3
source term and organic iodine converted from the elemental iodine
deposited in main steam piping systems. If the actual iodine
composition from the fuel release (cesium iodine) is used in the
calculations, essentially all of this organic iodine dose would be
eliminated.
The TSC doses due to MSIV leakage are especially conservative.
It is not expected that there will be any radioactive releases to
the TSC due to MSIV leakage during the initial stages of a LOCA
since it would take considerable time for the MSIV leakage to travel
through the main steam lines and main steam line drain system to the
condenser, into the turbine building, and finally to the atmosphere
and TSC. It was conservatively estimated that the 30-day integrated
dose to personnel in the TSC would increase by only 0.02 rem. The
dose calculations were performed using control room occupancy
factors specified in NUREG-0800, Standard Review Plan (SRP) Section
6.4.
Therefore, we conclude that the proposed change will not
significantly increase the probability or consequences of any
previously analyzed accidents.
2. The proposed change will not create the possibility of a new
or different kind of accident from any previously evaluated. The
BWROG evaluated MSIV leakage performance and concluded that MSIV
leakage rates up to 100 scfh will not inhibit the capability and
isolation performance of the valves to isolate the primary
containment. There is no new modification to the MSIVs which could
impact their operability. The LOCA has been analyzed using the main
steam piping and condenser as a treatment method to process MSIV
leakage at the proposed maximum rate of 200 scfh through all four
main steam lines. Therefore, the proposed change will not create any
new or different kind of accident from any accident previously
analyzed in the UFSAR.
3. Operation of the DAEC in accordance with the proposed change
will not involve a significant reduction in the margin of safety.
The allowable leak rate limit specified for the MSIVs is used to
quantify a maximum amount of bypass leakage assumed in the LOCA
radiological analysis. Results of the analysis are evaluated against
the dose requirements contained in 10 CFR 100 for the offsite doses
and 10 CFR 50, Appendix A (General Design Criterion 19) for the
control room and TSC doses.
The margins of safety are not significantly affected because the
dose levels remain well below the limits of 10 CFR 100 and General
Design Criterion 19. Therefore, the proposed change does not involve
a significant reduction in the margin of safety at the DAEC.
Description of Amendment Request:
Proposed Change 2
This proposed change to delete TS 3.7.E and 4.7.E and Bases
section 3.7.E and 4.7.E involves eliminating the MSIV leakage
control system (LCS) requirements from the TS.
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. As currently described in the UFSAR, the LCS is manually
initiated after a design basis LOCA occurs. Since the LCS is
operated only after an accident has occurred, this proposed
amendment has no effect on the probability of an accident. The
proposed change results in acceptable radiological consequences of
the design basis LOCA previously evaluated in the UFSAR.
The DAEC has an inherent MSIV leakage treatment capability. IES
Utilities Inc. proposes to use the main steam line drains and
condenser as an alternative to the LCS. Figure 1.1 of Attachment 2
shows the primary and alternate drain paths. The proposed primary
drain path at DAEC employs an MSL drain downstream of the MSIVs.
There are two motor-operated valves (MOVs) in series in this line
between the MSL and the main condenser. Both valves must be open to
establish the required drain path. Both MOVs will be provided with
essential power to assure that they can be opened following the DBA
LOCA to establish a large enough drain path to support the
radiological analysis.
An alternate drain path will be available to convey MSIV leakage
to the isolated condenser if either MOV fails to open. The alternate
drain path consists of the bypass lines around the MOVs in the
primary drain path. This alternate path contains a ``fail open''
valve and a restricting orifice. Consequently, if either primary MOV
failed to open as required, the second drain path would be available
to convey MSIV leakage to the main condenser. Radiological dose
calculations have been performed for this alternate path as well as
for the primary path. The results were acceptable. IES Utilities
Inc. will update DAEC procedures as necessary to address the
applicable alternate leakage treatment methods.
IES Utilities Inc. contracted with EQE Engineering Consultants
(EQE) to confirm the seismic capability of the DAEC's main steam
piping and condenser to serve as an alternate leakage treatment
system. Seismic verification walkdowns were performed to assure that
the MSLs, the steam drain lines, the condenser, and interconnecting
piping and equipment that were not seismically analyzed fall within
the bounds of the design characteristics of the seismic experience
database as discussed in Section 6.7 of the BWROG report.
The DAEC main steam lines, main steam drain lines, condenser,
and applicable interconnecting piping and equipment, are well
represented by the earthquake experience data demonstrating good
seismic performance, are confirmed to exhibit excellent resistance
to damage from a design basis earthquake and have been shown to have
substantial margin for seismic capability. The outliers that were
identified are discussed in Attachment 7. They have been either
evaluated to demonstrate their acceptability as they currently
exist, or plant modifications will be implemented to resolve the
concerns. By taking the measures discussed in Attachment 7 to ensure
resolution for all of the identified outliers, IES Utilities Inc. is
assured that the damage reported for the database components should
not occur to the DAEC main steam piping and condenser or to the
associated support systems.
Therefore, the proposed method for MSIV leakage treatment is
seismically adequate to withstand the DAEC design basis earthquake
and maintain pressure retaining integrity and serve as an acceptable
alternative to the currently installed LCS. The capability of the
alternate MSIV leakage treatment system to withstand the effects of
the safe shutdown earthquake and continue to perform its intended
function (treatment of MSIV leakage) satisfies the intent of the
seismic requirement of Appendix A to 10 CFR 100.
Plant specific radiological analyses have been performed to
assess the effects of MSIV leakage in terms of control room and
offsite doses following a postulated design basis LOCA. While not
previously considered a requirement for the design of the LCS, dose
calculations were also performed for the TSC. These analyses utilize
the hold-up volumes of the main steam piping and condenser as an
alternate treatment method for the MSIV leakage. The analysis
demonstrates that the proposed change results in an acceptable
increase in the radiological consequences of a LOCA previously
evaluated in the UFSAR. The LOCA previously evaluated in the UFSAR
is still the bounding accident; the proposed change will not involve
a significant increase in the consequences of an accident previously
analyzed.
The LCS lines will be disconnected, capped and welded, ensuring
that the integrity of the primary containment is maintained. IES
Utilities Inc. will incorporate the alternate leakage treatment
system into the inservice inspection (ISI) and inservice testing
(IST) programs, consistent with program requirements.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated. The
purpose of the LCS is to reduce the untreated MSIV leakage when
isolation of the primary coolant system and containment are
required. Radiological dose contributions due to MSIV leakage are
bounded by a LOCA. The LOCA has been analyzed using the main steam
piping and condenser as a treatment method to process MSIV leakage
at the proposed maximum rate of 100 scfh per MSIV and 200 scfh total
maximum pathway leakage, and determined to be within the regulatory
requirements. The LCS lines connected to the main steam lines will
be permanently closed to assure the primary containment integrity,
isolation, and leak testing capability are not compromised.
3. The proposed change to delete TS 3.7.E and 4.7.E and Bases
section 3.7.E and 4.7.E does not involve a significant reduction in
the margin of safety. The intended function of the LCS for treatment
of MSIV leakage will be performed by using the more effective
alternate path via the main steam drain lines and condenser. This
treatment method is effective for treatment of MSIV leakage over an
expanded leakage range. Except for the requirement to assure that
certain valves are opened to establish a proper flow path from the
MSIVs to the condenser and that certain valves are closed to
establish the seismic boundary, the proposed method is passive and
does not require any logic controls or interlocks. On the other
hand, the LCS consists of complicated logic controls and sensitive
equipment which must be maintained at significant cost and radiation
exposure. The radiological effects on the margin of safety are
discussed above for Change 1. The safety significance of the LCS in
terms of public risk was addressed in NUREG/CR-4330 which contains
the evaluation for eliminating the LCS and disabling the systems
currently installed at BWRs. The conclusion was that the increased
public risk is less than 1 percent. Therefore, the proposed change
does not involve a significant reduction in the margin of safety at
the DAEC.
The various attachments referred to in the above analysis may be
found in the licensees request for amendment dated August 15, 1994.
This document is available in the NRC's Public Document Room located at
the Gelman Building, 2120 L. Street, NW., Washington, DC 20555 and at
the local public document room address below.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, 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: Cedar Rapids Public Library, 500 First
Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036
NRC Project Director: John N. Hannon
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: June 2, 1994, as supplemented August 25,
1994
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) to remove expired one-time
extensions of surveillances, remove an obsolete definition of charging
pump operability, and incorporate 11 line item improvements in
accordance with the guidance provided in Generic Letter (GL) 93-05.
Other editorial changes would be made to renumber some pages and delete
the blank pages from 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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated (10
CFR 50.92(c)(1)). The expired one-time extensions were in effect to
September 30, 1993. Since these extensions have expired and the
appropriate surveillances were performed, the proposed changes do not
effect the configuration, operation, or performance of any system, or
component.
The proposals to delete Definition 1.45, ``THE CHARGING PUMP
OPERABILITY,'' and modify the Index to reflect this change are
administrative changes. Definition 1.45 was applicable only for cycle 4
operation. Northeast Nuclear Energy Company (NNECO) has completed the
necessary modifications and no longer rely on a temporary heating
source. Therefore, the elimination of Definition 1.45 does not involve
a significant increase in the probability or consequences of an
accident previously analyzed.
The proposed changes to incorporate the recommendations of GL 93-05
do not affect the configuration, operation or performance of the
subject systems. Increasing the surveillance test intervals as proposed
will reduce the number of surveillance tests and minimize the potential
for inadvertent actuation of an engineered safety feature. The increase
in the surveillance test intervals will enhance the operational
effectiveness of plant personnel, by reducing the amount of time that
the plant staff has available to perform other tasks, such as
additional preventive maintenance. Additionally, increasing the
surveillance test interval will reduce unnecessary wear to equipment.
NNECO's proposals to delete pages that were intentionally left blank,
to renumber remaining pages and renumber Sections, and modify the Index
to reflect these changes are purely administrative and editorial
changes. Proposals to correct typographical errors on TS pages are also
administrative changes. These changes would not affect the
configuration, operation, or performance of any system, structure, or
component.
The proposed changes do not affect the manner by which the facility
is operated and do not change any facility design feature or equipment.
The proposed changes involve administrative or programmatic
requirements or merely involve editorial changes, corrections, or
clarifications. Since there is no change to the facility or operating
procedures, there is no affect upon the probability or consequences of
any accident previously analyzed.
B. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because they do not affect the manner by which the
facility is operated and do not change any facility design feature or
equipment which affects the operational characteristics of the
facility. The proposed changes involve administrative or programmatic
requirements or merely involve editorial changes, corrections, or
clarifications.
C. The changes do not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)) because the proposed changes do not
affect the manner by which the facility is operated or involve
equipment or features which affect the operational characteristics of
the facility.
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: Learning Resource Center, Three Rivers
Community-Technical College, Thames Valley Campus, 574 New London
Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: John F. Stolz
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: July 22, 1994
Description of amendment request: The proposed amendment would
revise the Technical Specifications to incorporate a different setpoint
and transient methodology for determining the maximum allowable power
range neutron flux setpoint. The changes would allow Millstone Unit 3
to operate with a reduced number of main steam-line safety valves at a
reduced power level, as determined by the high neutron flux setpoint.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
...The proposed changes do not involve an SHC [significant
hazards consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Technical Specification Tables 3.7-1 and 3.7-2 are being revised
to reflect a reduction in the maximum allowable power range neutron
flux high setpoint with inoperable steam generator safety valves.
The new setpoints reflect a change in the methodology for
calculating the setpoints.
Westinghouse has determined that under certain conditions with
typical safety analysis assumptions, the current setpoints in Tables
3.7-1 and 3.7-2 may not provide adequate steam generator
overpressure protection for a Loss of Load/Turbine Trip transient at
reduced power levels. At reduced power levels, a reactor trip may
not be actuated early in the transient. An overtemperature delta T
trip may not be generated since the core thermal margins are
increased at lower power levels. The PORVs [power-operated relief
valves] and pressurizer spray may control RCS [Reactor Coolant
System] pressure such that a high pressurizer pressure trip isn't
generated. The reactor would eventually trip on low steam generator
water level, but this may not occur before steam pressure exceeds
110% of the design value if one or more MSSVs [main steam-line
safety valves] are inoperable.
To address this issue, Westinghouse has developed a new method
for determination of the required power range neutron flux high
setpoint. The new setpoint is based upon the heat removal capability
of the operable MSSVs, rather than the previous method based only on
flow capacity. The new equation is shown in the proposed changes to
the Technical Specification basis. This new method has been
developed by Westinghouse generically and a Millstone Unit No. 3
specific calculation has been performed. The new setpoints are being
incorporated in this proposed Technical Specification change.
The new method includes several conservative assumptions. The
equation is developed assuming that the maximum number of inoperable
MSSVs applies to each loop. For example, for four loop operation,
the maximum allowable power range neutron flux high setpoint of 65%
is based upon four inoperable MSSVs, one per steam generator. Thus,
in the event that only one MSSV is inoperable, the application of
the new setpoint is very conservative. In addition, the setpoint is
based upon the assumption that the largest capacity MSSV is
inoperable. For the case where one of the lower capacity MSSVs is
inoperable, the setpoint will be conservative.
The method of calculating the setpoint provides assurance that
the heat removal capability of the operable MSSVs is sufficient for
reactor power up to the power range neutron flux high setpoint
taking into account instrument and channel uncertainties.
Consequently, steam generator pressure will remain below 110% of
design in the event of the limiting overpressurization transient,
the Loss of Load/Turbine Trip.
Reducing the power range neutron flux high setpoint and
consequently the allowable reduced power level has no impact on the
consequences of any other accident. In addition, since the proposed
changes only involve a reduction in the allowable power range
neutron flux high setpoint, and operation at a lower power level,
they cannot affect the probability of any design basis accident.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Since the proposed changes just reduce the existing limit on the
power range neutron flux high setpoint with inoperable MSSVs, the
change cannot create the possibility for a new or different kind of
accident.
3. Involve a significant reduction in the margin of safety.
The reduced setpoint provides additional assurance that the
steam generator pressure will remain below 110% of design for the
limiting overpressurization transient, the Loss of Load/Turbine
Trip. Thus, the proposed changes do not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room: Learning Resource Center, Three Rivers
Community-Technical College, Thames Valley Campus, 574 New London
Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford,
Connecticut, 06141-0270.
NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket No. 50-387 Susquehanna
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania
Date of amendment request: July 27, 1994
Description of amendment request: By letter dated June 15, 1992,
Pennsylvania Power and Light Company (PP&L) submitted ``Licensing
Topical Report NE-092-001, Revision 0, Power Uprate With Increased Core
Flow,'' for Susquehanna Steam Electric Station, Units 1 and 2. The
report was submitted to support future amendments to the Units 1 and 2
licenses to permit a 4.5-percent increase in reactor thermal power and
an 8-percent increase in core flow for each unit. The initial submittal
was revised and supplemented by letters of July 24, September 17, and
December 18, 1992, and January 8, January 25, April 2, August 5, August
12, and September 29, 1993. The Commission's safety evaluation on these
submittals was issued November 30, 1993 (Letter, Thomas E. Murley, NRC,
to Robert G. Byram, PP&L). The Commission concluded that the revised
(Revision 2) licensing topical report adequately supports PP&L's
proposed power uprate. The Commission also concluded that SES, Units 1
and 2, can operate safety with the proposed 8-percent increase in core
flow, the proposed 4.5-percent increase in reactor thermal power, the
corresponding 5-percent increase in main turbine inlet steam flow, and
the corresponding increases in flows, temperatures, pressures, and
capacities required in supporting systems and components at these
uprated conditions.This amendment will change several Technical
Specifications sections (listed below in the no significant hazards
consideration) for Susquehanna Steam Electric Station, Unit 1, to
increase the licensed power level from the current 3293 MWt to a new
limit of 3441 Mwt.
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 following three questions are addressed for each of the
proposed Technical Specification Changes:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
3. Does the proposed change involve a significant reduction in a
margin of safety?
Section 1.0, Definitions, Definition 1.33, Rated Thermal Power
This change redefines Rated Thermal Power as 3441 megawatts
thermal.
1. No. Neither the probability (frequency of occurrence) nor
consequences of any accident previously evaluated is significantly
affected by the increased power level because the design and
regulatory criteria established for plant equipment remain imposed
for the uprated power level. The PP&L assessment to increase the
rated thermal power level at Susquehanna SES Unit 1, followed the
guidelines of NEDC-31879P (Generic Guidelines for General
Electric Boiling Water Reactor Power Uprate,'' G.E. Nuclear Energy,
June 1991). NEDC-31879P provides generic licensing criteria,
methodology, and a defined scope of analytical and equipment review
to be performed to demonstrate the ability to operate safely at the
uprated power level which have been approved by the NRC. NE-092-001
(Licensing Topical Report for Power Uprate With Increased
Core Flow,'' Pennsylvania Power & Light Company, December 1992)
provides the description of the power uprate licensing analysis
methodology and the results of the evaluations performed to support
the proposed uprated power operation consistent with the methodology
presented in NEDC-31879P. NE-092-001 provides a description of the
power uprate licensing analysis methodology which will be used to
determine cycle specific thermal limits for Unit 1, Cycle 9 and
future cycles and concludes that an uprated power level of 3441
megawatts thermal can be achieved without significant effect on
equipment or safety analyses.
2. No. The methodology and results described above do not
indicate that a possibility for a new or different kind of accident
from any previously evaluated has been created by uprated operation.
3. No. Based on the response to Question 1 above, the
methodology and results do not indicate a significant reduction in a
margin of safety.
Section 2.1, Safety Limits
The reference to ``rated core flow'' in Technical Specification
2.1.1 and 2.1.2 has been replaced with a reference to actual core
flow. The references to ``rated core flow'' have been deleted to
avoid confusion since allowable core flow is being increased by 8%.
10 Mlbm/hr is being used in these specifications to be consistent
with other similar Technical Specification changes (Technical
Specifications 3.2.2, 4.4.1.1.1.2, 4.4.1.1.2.5, 3.4.1.3 and Figure
3.4.1.1.1-1).
1. No. The probability and consequences of accidents previously
evaluated are not affected by this change. The basis for Technical
Specification 2.1.1 is that boiling transition will not occur in
bundles if core power is less than 25% of rated thermal power,
regardless of pressure or core flow. Consequently, the specification
of less than 10% rated core flow is not crucial to the basis and,
thus, the use of 10 Mlbm/hr. is acceptable and has no effect on the
probability or consequences of a previously evaluated accident.
For Technical Specification 2.1.2, the XN-3 critical power
correlation is valid for pressure greater than or equal to 580 psig
and bundle flow greater than or equal to 0.25 Mlbm/hr-ft2. As
stated in the basis for Technical Specification 2.1.1, if vessel
downcomer water level is above TAF [top of active fuel], and core
power greater than 25%, bundle flows for potentially limiting
bundles will be greater than 0.25 Mlbm/hr-ft2 due to natural
circulation. In addition, Technical Specification 3.4.1.1.1 requires
at least one (1) recirculation loop in operation to run in Condition
2, which would produce a core flow in excess of 30 Mlbm/hr.
Therefore, core flows below about 30 Mlbm/hr-ft2 are prohibited
when the reactor is at power. Thus, the change from ``10%'' to ``10
million lbm/hr'' is acceptable and has no effect on the probability
or consequences of a previously evaluated accident.
2. No. The basis for Technical Specification 2.1.1 is that
boiling transition will not occur in bundles if core power is less
than 25% of rated thermal power, regardless of pressure or core
flow. The proposed change is not crucial to this basis. The XN-3
critical power correlation is valid for pressures greater than or
equal to 580 psig and bundle flow greater than or equal to 0.25
Mlbm/hr-ft2. The specification is based upon vessel downcomer
water level being above TAF and core power greater than 25% which
yields a bundle flow for potentially limiting bundles greater than
0.25 Mlbm/hr-ft2 due to natural circulation. Based on Technical
Specification 3.4.1.1.1, core flows below about 30 Mlbm/hr-ft2
are prohibited when the reactor is at power. Therefore, the change
to a limit of 10 Mlbm/hr is acceptable and does not create the
possibility for a new or different kind of accident from any
accident previously evaluated.
3. No. As explained above, the margin of safety has not been
reduced.
Table 2.2.1-1 (Items 2.a, 2.b, and 2.c) and Specifications
3.2.2, 3.4.1.1.2.a.2, 3.4.1.1.2.a.3, 3.4.1.1.2.a.5.b and 3.3.6-2
(Item 2.a.1, 2.c, and 2.d), APRM Flow Biased Setpoints and Allowable
Values
Although the equation for determining these setpoints does not
change as a result of the power uprate, because the setpoints in
these technical specifications are referenced to rated thermal
power, the current limits do change in that the top portion of the
operating map (power vs. reactor flow) is raised by 4.5%.
1. No. The safety analyses contained in NE-092-001 evaluated
operation at both uprated power with 4.5% higher rod lines and
increased core flow. In addition, General Electric Co. has analyzed
and received generic approval for their BWR/4 product line operation
in the Maximum Extended Operating Domain (MEOD). Operation at the
4.5% higher rod lines is bounded by the MEOD analysis. Additional
justification for this small increase in the power flow operating
range is contained in Section C.2.3 of NEDC-31984P.
Cycle specific reload analyses will evaluate operation at the
increased power vs. flow conditions (100% uprated power vs. 87% core
flow to 100% uprate power vs. 108% core flow). These analyses will
ensure that the limits established in the Core Operating Limits
Report are applicable to rated power operation from 87% to 108% core
flow.
Based on the above analyses, increasing the current limits do
not represent a significant increase in the probability or
consequences of an accident previously evaluated.
2. No. The analyses described above in response to Question 1 do
not indicate that a possibility for a new or different kind of
accident from any previously evaluated has been created by the
proposed change.
3. No. Based on the response to Question 1 above, the proposed
change does not result in a reduction in the margin of safety.
Table 2.2.1-1, Item 3, Reactor Steam Dome Pressure - High Scram
The reactor steam dome pressure-high scram trip setpoint and
allowable values are being changed to less than or equal to 1087
psig and less than or equal to 1093 psig respectively.
1. No. This scram function is designed to terminate a pressure
increase transient not terminated by direct scram or high flux
scram. The nominal trip setpoint is maintained above the reactor
vessel maximum operating pressure and the specified analytical limit
is used in the transient analyses. The analytical limit of 1105 psig
is used in the uprated transient analyses. The results of the
overpressure protection analysis indicate that the peak pressure
will remain below the 1375 psig ASME limit which meets plant
licensing requirements. In accordance with the methodology described
in NE-092-001, transient analyses will be performed using the
analytic limit and the results will be incorporated into the Core
Operating Limits Report. Therefore, this proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. No. The purpose of this scram function is to terminate a
pressure increase transient not terminated by direct scram or high
flux scram. The nominal trip setpoint is maintained above the
reactor vessel maximum operating pressure and the specified
analytical limit is used in the transient analysis. 1105 psig is
being used as the analytical limit in the uprated transient
analysis. The results of the overpressure protection analysis
indicate peak pressure will remain below the ASME limit of 1375 psig
which satisfies plant licensing requirements. Based upon that
result, it is concluded that the proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. No. The results of the overpressure protection analysis
indicate peak pressure will remain below the 1375 psig licensing
limit, therefore, it is concluded that the proposed change does not
result in a significant reduction in a margin of safety.
Specification 4.1.5.c, Standby Liquid Control System
This specification has been revised to require SLC [Standby
Liquid Control] pumps to develop a discharge pressure of greater
than or equal to 1224 psig.
1. No. The ability of the SLC system to achieve and maintain
safe shutdown is a function of the amount of fuel in the core and is
not directly affected by core thermal power. The SLC pump test
discharge pressure acceptance criteria are based on the lowest
relief valve setpoint. The lowest setpoint is being increased by 30
psi (to 1106) due to power uprate. Operating with increased core
flow will result in additional friction losses through the core and
a slightly larger core differential pressure (approximately 4 psi).
Therefore, increasing the SLC pump test discharge pressure
acceptance criteria ensures the capability of SLC injection. The
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. No. The ability of the SLC system to achieve and maintain
safe shutdown is a function of the amount of fuel in the core and is
not directly affected by core thermal power. Therefore, the proposed
change does not result in a new or different kind of accident from
any previously evaluated.
3. No. The ability of the SLC system to achieve and maintain
safe shutdown is a function of the amount of fuel in the core and is
not directly affected by core thermal power. As stated in the
response to question 1 above, the SLC pump discharge pressure
acceptance criteria are based upon the lowest relief valve setpoint.
The lowest setpoint is being increased by 30 psi. As the SLC pumps
are positive displacement pumps, the uprate will not adversely
affect the performance of the pumps to achieve proper injection.
Based on above, the proposed change does not result in a significant
reduction in a margin of safety.
Specifications 3.2.2, 4.4.1.1.1.2, 4.4.1.1.2.5, 3.4.1.3 and
Figure 3.4.1.1.1-1, Rated Core Flow References
Technical Specification 3.2.2 contains the definition of ``W''
for the flow biased APRM scram equation. The word ``rated'' is being
deleted from the definition of ``W'' since rated core flow is being
increased. The definition of ``W'' is not altered. The change is
being made for editorial purposes.
Technical Specifications 4.4.1.1.1.1.2, 4.1.1.1.2.5, 3.4.1.3,
and Figure 3.4.1.1.1-1 specify performance requirements and limits
for the Reactor Recirculation System. These specifications are
referenced to the current rated core flow. The references to ``rated
core flow'' are being replaced with actual equivalent core flows.
The specifications are equivalent and unchanged. This change is
being made for editorial purposes to avoid confusion since rated
core flow is being increased. These changes are also consistent with
the changes made in Section 2.1.
1. No. The proposed changes are editorial and do not effect the
probability or consequences of an accident previously evaluated.
2. No. The proposed changes are editorial and do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. No. The proposed changes are editorial and do not involve a
significant reduction in a margin of safety.
Specification Table 3.3.1-1, Note (j) and Action 6, Reactor
Protection System Instrumentation, and Table 3.3.4.2-1, Note b, End-
of-Cycle Recirculation Pump Trip System Instrumentation
The turbine first stage pressure scram bypass at 30% power in
Technical Specification Table 3.3.1-1, Note (j) and Table 3.3.4.2-1,
Note (b) is revised to indicate that the uprated equivalent
allowable value of first stage turbine pressure is 136 psig. This
value ensures that the analytical limit of 147.7 psig, which
represented 30% rated thermal power, is not exceeded.
As currently written Note (j), Note (b) and Table 3.3.1-1,
ACTION 6 are unclear and could be misinterpreted. They apply only
when RPS scram functions and End-of-Cycle Recirculation Pump Trip on
turbine main stop valves closure or control valve fast closure are
not automatically bypassed. ACTION 6 provides no guidance in the
event the bypass fails to lift when thermal power is above 30%. In
the worst case, the action statement could be interpreted literally
to allow full power operation with the RPS function still bypassed.
Such operation would violate the licensing basis analysis for the
MCPR operating limit (for the Generator Load Rejection Without
Bypass transient), which takes credit for operation of the
anticipatory scram on control valve fast closure at greater than 30%
of rated thermal power.
1. No. The revisions to Table 3.3.1-1, ACTION 6, Table 3.3.1-1,
Note (j), and Table 3.3.4-1 Note (b) clarify the current
requirements; they do not change their intent.
FSAR Chapter 15 transient analyses and reload licensing analyses
take credit for operation of the anticipatory scram function on
turbine stop valve closure and control valve fast closure for power
levels greater than 30% of rated thermal power. The proposed
revision to Table 3.3.1-1, ACTION 6 provides better assurance of the
availability of the anticipatory scram function, since the current
specifications could be interpreted literally to allow full power
operation with the RPS function bypassed.
The proposed revision to Table 3.3.1-1, Note (j) and Table
3.3.4.2-1, Note (b) does not change the operation of the RPS and
EOC-RPT bypasses on turbine stop valve closure and control valve
fast closure below 30% power. The turbine first stage pressure
switches will still be calibrated in the same manner, and, by
procedure, the reactor operator will not exceed 30% power if the
trip bypass annunciator does not clear.
The setpoints for the RPS and EOC-RPT bypass functions were
selected to allow sufficient operating margin to avoid scrams during
low power turbine generator trips. As discussed in NEDC-31894P,
Section F4.2(c) and in Section 5.1.2.8 of NEDC 31948P, this small
absolute setpoint increase maintains the safety basis for the
setpoint.
Based on the above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. No. The changes proposed are clarifications and do not change
specification intent. The proposed change to Table 3.3.1-1, Action 6
provides better assurance of the availability of the anticipatory
scram function as the specification could currently be interpreted
to allow full power operation with the RPS function bypassed. The
proposed changes to Table 3.3.1-1, Note (j) and Table 3.3.4-1, Note
(b) do not change the operation of the RPS and EOC-RPT bypasses on
turbine stop valve closure and control valve fast closure below 30%
power. Therefore, the possibility for a new or different kind of
accident is not created.
3. No. The proposed changes are clarification and do not change
intent. Operation of the RPS and EOC-RPT bypasses on turbine stop
valve closure and control valve fast closure below 30% power is not
changed. Therefore, there is no reduction in the margin of safety.
Specification Table 3.3.2-2, Item 3.d, Main Steam Line Flow
Differential Pressure Setpoint
The main steam line flow high differential pressure setpoint and
allowable value are revised to read trip setpoint and allowable
values of 113 psid and 121 psid respectively. Footnote ``**'' was
added to Table 3.3.2-2 to indicate that these values will be
confirmed during the power uprate start-up testing. If revisions to
the setpoint and allowable value are required, they will be
forwarded to the Commission for approval within 90 days of
completion of the test program.
1. No. The main steam line flow high differential pressure
setpoint changes reflect the redefinition of rated main steam line
flow that occurs with power uprate. The allowable value is
maintained at the same percentage of rated steam flow as the
differential pressure changes due to the increased uprate steam
flow. The analytical limit of 140% of uprated steam flow is
maintained for the uprated analyses. The relationship between the
allowable value and the analytical limit was retained to ensure that
a trip avoidance margin is maintained for the normal plant testing
of MSIV's and turbine stop valves. The increase in the absolute
value of the trip setpoint still provides a high assurance of
isolation protection for a main steam line break accident which
satisfies the original intent of the design. Therefore, the proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. No. The increase in the absolute value of the trip setpoint
still provides a high assurance of isolation protection for the main
steam line break accident which satisfies the original intent of the
design and, therefore does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. No. The increase in the absolute value of the trip setpoint
still provides a high assurance of isolation protection for a main
steam line break accident which satisfies the original intent of the
design and, therefore, does not involve a significant reduction in a
margin of safety.
Specification Table 3.3.2-2, Item 4.f, Isolation Actuation
Instrumentation Setpoints
The RWCU system flow-high isolation trip setpoint and allowable
value are being changed. System flow is being increased by 10% to
maintain reactor coolant water chemistry at a level equal to pre
uprate levels. The isolation setpoint change is being made to
adequately maintain operating margin between normal process values
and the isolation setpoints.
1. No. The basis for the RWCU flow-high isolation is to ensure a
RWCU System isolation in case of a pipe break. The high flow
setpoint is set high enough to avoid spurious trips from normal
operating transients but low enough to ensure an isolation during a
pipe break. The proposed Technical Specification limits will result
in a negligible reduction in the margin between the RWCU isolation
setpoint and the 4350 gpm flow postulated during a RWCU line break
and will avoid spurious isolations. Therefore, the proposed change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. No. As stated above, the proposed change will result in only
a negligible reduction in the margin between the RWCU isolation
setpoint while avoiding spurious isolation. Therefore, this change
maintains the original design intent and does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. No. See 1. above.
Specification Table 3.3.2-2, Items 5.a and 6.1, Isolation
Actuation Instrumentation Setpoints
The HPCI and RCIC Steam Line Flow-High Technical Specifications
are being changed to account for changes in steam conditions and
flows that result from operation at the uprated conditions. The
setpoint and allowable value for HPCI Steam Line Flow-High isolation
are less than or equal to 387 inches H2O setpoint and allowable
value for the RCIC Steam Line Delta Pressure-High isolation are less
than or equal to 188 inches H2O and less than or equal to 193
inches H2O respectively.
1. No. The bases for these setpoints are contained in the
General Electric Design Specification Data Sheets for the HPCI and
RCIC systems. The Design Specification Data Sheets specify that the
setpoint and allowable value be set so that the isolation occurs at
greater than 272% normal steam flow and less than 300% steam flow.
General Electric has historically seen start-up transients as high
as 272% of normal steam flow. Setting the isolation above this value
prevents spurious isolations and ensures availability of the system
and its safety function. Setting the isolation at less than or equal
to 300% of normal flow insures that the isolation will occur if a
steam line should rupture.
The existing setpoints were calculated using information
obtained during the recent surveillance tests. The revised setpoints
and allowable values were calculated using the current system
performance and adjusted for uprate conditions in accordance with
additional guidance provided in General Electric Information Letter
(SIL) No. 475, Revision 2, NEDC-31336, ``General Electric Setpoint
Methodology,'' and GE Letter SPU-9378, ``HPCI and RCIC Steam Line
Break Detection Setpoints''.
Based on the above approach, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. No. The setpoint and allowable value are set so that
isolation occurs at greater than 272% normal steam flow and less
than 300% steam flow. Setting the isolation at less than or equal to
300% of normal flow ensures that the isolation will occur if a steam
line rupture should occur. Therefore, no new events are postulated
as a result of this change.
3. No. The proposed change does not involve a significant
reduction in a margin of safety as the setpoint and allowable value
are set to isolate at greater than 272% normal steam flow and less
than 300% steam flow which are the setpoints contained in the
General Electric Design Specification Data Sheets for the HPCI and
RCIC systems.
Specification Table 4.3.2.1-1, footnote ``**''
The footnote is being changed to delete reference to reactor
pressure.
1. No. The original purpose of Footnote ``**'' to Technical
Specification Table 4.3.2.1-1 was to describe the functioning of the
permissive circuitry that allowed the MSIV low condenser pressure
isolation to be bypassed. The original circuitry required the Mode
Switch not be in Run, the Turbine Stop Valves closed, and reactor
pressure to be above setpoint. In the start-up phase of the
Susquehanna Units, General Electric deleted the reactor pressure
setpoint input to the bypass circuitry. Therefore, this change is
being made to make the footnote conform to the installed
configuration. The revised footnote is the same as found in the BWR/
4 Standard Technical Specifications (NUREG 1433). This change is
editorial in nature and, therefore, does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. No. Based on the response to Question 1 above, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. No. Based on the response to Question 1 above, the proposed
change does not involve a significant reduction in a margin of
safety.
Specification Table 3.3.6-2, Item 1.a and Specification
3.4.1.1.2.a.5.a, Rod Block Monitor Flow Biased Rod Blocks
The Rod Block Monitor (RBM) flow biased rod blocks are being
changed as follows:
a. Technical Specification Table 3.3.6-2, Item 1.a is revised to
read trip setpoint and allowable values of less than or equal to
0.63 W + 41% and less than or equal to 0.63 W + 43%, respectively.
b. Technical Specification 3.4.1.1.2.a.5.a is being revised to
read trip setpoint and allowable values of less than or equal to
0.63 W + 35% and less than or equal to 0.63 W + 37%, respectively.
1. No. These Technical Specification changes do not represent a
change from current limits. The change reflects the rescaling made
necessary by the re-definition of rated thermal power.
The RBM flow biased rod blocks are used in the Rod Withdrawal
Error (RWE) analysis. In order to maintain Critical Power Ratio
(CPR) margins similar to previous Susquehanna cycles, the flow
biased rod blocks were changed in terms of megawatts thermal but the
change was not appreciable. The rescaling of the RBM flow biased rod
block to reflect the re-definition of Rated Thermal Power maintains
the same level of protection as previously provided. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. No. These changes do not represent a change from current
limits but are rather a rescaling made necessary by the re-
definition of rated thermal power.
3. No. These changes do not represent a change from current
limits but are rather a rescaling made necessary by the re-
definition of rated thermal power. The rescaling of the RBM flow
biased rod block maintains the same level of protection as
previously provided.
Specification Table 3.3.6-2, Item 2.a, Control Rod Block
Instrumentation Setpoints
The APRM rod block upscale value has been changed to add a high
flow clamp setpoint at 108% with a high flow clamped allowable value
at 111%.
1. No. The addition of the high flow clamp to the flow biased
APRM rod block function maintains the normal margins between the rod
block and the scram power levels in the increased core flow regions.
When the reactor core flow is greater than 100 million lbm/hr, the
APRM clamp provides an alarm to help the operator avoid scrams while
operating in the ICF region. This action does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. No. The changes maintain the normal margins between the rod
block and the scram power levels in ICF regions. The clamp provides
an alarm to avoid scrams in the ICF region.
3. No. The changes maintain the normal margins between the rod
block and the scram power levels.
Specification Table 3.3.6-2, Item 6.a, Reactor Coolant System
Recirculation Flow Upscale Rod Block Setpoint and Allowable Value
Change
The reactor coolant system recirculation flow upscale rod block
setpoint and allowable value are being increased to 114/125
divisions of full scale and 117/125 divisions of full scale
respectively.
1. No. The Reactor Coolant System recirculation flow upscale rod
block setpoint and allowable value are being increased to allow
operation in the ICF region. The 114/125 divisions setpoint and 117/
125 divisions allowable value, specified by General Electric, are
based on BWR operating history.
The purpose of the Reactor Coolant System recirculation flow
upscale rod block is to prevent rod movement when an abnormally high
increase in reactor recirculation flow exists. An increase in
reactor recirculation flow causes an increase in neutron flux that
results in an increase in reactor power. However, this increase in
neutron flux is monitored by the Neutron Monitoring System that can
provide a rod block. No design basis accident or transient analysis
takes credit for rod block signals initiated by the Reactor Coolant
Recirculation System. Therefore, this change does not increase the
probability or consequences of an accident previously evaluated.
2. No. Rod block signal initiation by the Reactor Coolant
Recirculation System is not taken credit for in the mitigation of a
design basis accident or in any transient analysis.3. No. Rod block
signal initiation by the Reactor Coolant Recirculation System is not
taken credit for in any transient analysis or in the mitigation of a
design basis accident.
Specification 4.4.1.1.1.2 and 4.4.1.1.2.5 Reactor Coolant System
The reactor recirculation pump motor generator set scoop tube
electrical and mechanical overspeed stop setpoints are being
increased to a core flow of 109.5 million lbm/hr. and 110.5 million
lbm/hr., respectively.
1. No. The reactor recirculation pump motor generator set scoop
tube stops are being increased to allow operation at core flows in
the ICF region of up to 108 million lbm/hr.
The electrical stop is maintained above the maximum operating
core flow and below the mechanical stop. The 109.5 million lbm/hr.
electrical stop setpoint, specified by General Electric, is based on
BWR operating history. The electrical stop is a system design
feature and is not used in any safety analyses.
The 110.5 million lbm/hr. mechanical stop setpoint is used in
transient analysis to limit core flow during a recirculation pump
controller failure. The 110.5 million lbm/hr. mechanical stop
setpoint, specified by General Electric, is also based on BWR
operating history. The cycle specific analyses, performed for power
uprate, used the 110.5 million lbm/hr. mechanical stop setpoint.
Based on the above, this change does not involve a significant
increase of the probability or consequences of an accident
previously evaluated.
2. No. Increasing the reactor recirculation motor generator set
scoop tube electrical and mechanical overspeed stop setpoints is
being done to allow operation at core flows in the ICF region up to
108 Mlbm/hr. The electrical stop setpoint is a design feature and is
not used in any safety analysis. The mechanical stop setpoint is
used in transient analysis to limit core flow during a recirculation
pump controller failure. Changing of this setpoint was considered in
appropriate transient analyses, and will not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. No. See 1. above. This change does not significantly reduce
the margin of safety.
Specification Figure 3.4.1.1.1-1, Thermal Power Restrictions
This figure has been redrawn to reflect the new definition of
Rated Thermal Power to retain the same stability operating
restrictions in terms of megawatts thermal as were previously
described by this graph.
1. No. The core thermal hydraulic stability curve and associated
bases are maintained at the current rod lines and power levels.
Those values are redefined to reflect the redefinition of rated
thermal power. Since the current operating restrictions are
maintained, power uprate has no detrimental effect on the level of
protection provided by these Technical Specifications. This position
is consistent with NEDC-31894P, Section 5.3.3 and with NEDC-31984P,
Section 3.2.
2. No. The core thermal hydraulic stability curve and associated
bases are maintained at the current rod lines and power levels.
Those values are changed to reflect the redefinition of rated
thermal power. Since the current operating restrictions are
maintained, power uprate has no detrimental effect on the level of
protection provided and does not create the possibility for a new or
different kind of accident.
3. No. The core thermal hydraulic stability curve and associated
bases are maintained at the current rod lines and power levels.
Those values are redefined to reflect the redefinition of rated
thermal power. Since the current operating restrictions are
maintained, there is no detrimental effect on the level of
protection provided, and therefore no significant decrease in any
margin of safety.
Specifications 3.4.1.1.2.5, 3.4.1.1.2.6, Reactor Coolant System,
Recirculation Loops - Single Loop Operation
Specification 3.4.1.1.2.5 is being renumbered to 3.4.1.1.2.6. A
new specification 3.4.1.1.2.5 is being added to specify that a 0.70
LHGR multiplier has been applied to Specification 3.2.4 when in
single recirculation loop operation.
1. No. Operation with one recirculation loop out of service is
allowed, but it is not considered a normal mode of operation. Single
loop operation (SLO) is a special operational condition when only
one of the two recirculation loops is operable. In this operating
condition, the reactor power will be limited to less than 80% of
rated by the maximum achievable core flow, which is typically less
than 60% of rated core flow. A postulated LOCA occurring in the
active recirculation loop during SLO would cause a more rapid
coastdown of the recirculation flow than would occur in two loop
operation, where one active loop would remain intact. This rapid
coastdown causes an earlier boiling transition and deeper
penetration of boiling transition into the bundle, which tends to
increase the calculated PCT. However, the PCT effects of early
boiling transition are substantially offset by the mitigating effect
of the lower power level achievable at the start of such an event.
The SAFER/GESTR-LOCA analysis results for Susquehanna for SLO and
two loop operation are well below 2200 deg.F and are documented in
NEDC-32064P-1, Revision 1, ``Power Uprate with Increased Core Flow
Safety Analysis for Susquehanna 1 and 2'', GE Nuclear Energy, July
1993.
The ECCS performance for Susquehanna under SLO was evaluated
using SAFER/GESTR-LOCA. Calculations for the DBA were performed
using both nominal and Appendix K inputs. The SLO SAFER/GESTR-LOCA
analysis for the DBA assumes that there is essentially no period of
recirculation pump coastdown. Thus, dryout is assumed to occur
simultaneously at all axial locations of the hot bundle shortly
after initiation of the event. Dryout is assumed to occur in one
second for the nominal case and 0.1 second for the Appendix K case.
These assumptions are very conservative and provide bounding results
for the DBA under SLO.
The two-loop Appendix K break spectrum documented in NEDC-
32064P-1 is representative of SLO because the two-loop spectrum was
analyzed assuming a one second dryout time for all axial locations
of the hot bundle. As shown by the two-loop break spectrum, the DBA
is the limiting case for SLO. With breaks smaller than the DBA,
there is a longer period of nucleate and/or film boiling prior to
fuel uncovery to remove the fuel stored energy.
An LHGR multiplier of 0.70 will be imposed when the plant is in
SLO. As shown in Table 5-6 of NEDC-32064P-1, the SLO results are
less limiting (i.e., lower PCT's) than the results for the two loop
DBA LOCA.
Thus, the licensing PCT is based appropriately on two loop
operation rather than SLO.
2. No. The licensing PCT is based upon two loop operation rather
than SLO, thus the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. No. Based on the response to Question 1 above, the proposed
change does not involve a significant reduction in a margin of
safety.
Specification 4.4.1.1.2.3, Reactor Coolant System
Footnote **** to this Specification is being changed to
reference the power uprate startup test program.
1. No. This footnote provided a mechanism for changing the power
limits specified if the results of the initial startup test program
determined that it was necessary. The footnote is being modified to
allow operation at uprated power with the present power limits.
Should the power uprate startup test program determine a need to
change the power limits they will be submitted to the Commission
within 90 days as required by the revised footnote. This is
consistent with the original BWR startup test program philosophy and
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. No. See 1. above; this change is administrative in nature and
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. No. See 1. above; this change is administrative in nature and
does not involve a significant reduction in a margin of safety.
Specification 3.4.2, Reactor Coolant system, Safety Relief
Valves
The safety relief valve specification is being changed to reduce
the number of setpoint groups from 5 to 3. Two valves will be set at
1175 psig plus or minus 1%, 6 will be set at 1195 psig plus or minus
1%. Also, the number of Operable safety valves is being increased
from 10 to 12.
1. No. This change does not increase the probability of
occurrence of an accident previously evaluated as, with one
exception, the accidents described in FSAR Sections 5.2.2, 7.2.3,
15.1, 15.2 and 15.3 do not document any cases where the SRV's are
designated as the cause or initiator of an accident. The exception
is inadvertent safety relief valve opening which results in a
decrease in reactor coolant inventory and/or reactor coolant
temperature. The revised setpoints and proposed groupings will not
increase the probability of occurrence of this type of accident.
The change does not increase the probability of occurrence of a
malfunction of equipment important to safety as previously evaluated
in the FSAR. The margin between peak allowable pressure and the
maximum safety setpoint is unchanged. The reactor vessel and
components were evaluated for the setpoint change to assure
continued compliance with the structural requirements of the ASME
Code. Analysis was performed on the effects of the setpoint change
for the design conditions, the normal and upset conditions and the
emergency and faulted conditions. The increasing RPV dome pressure
does not affect the design condition and, therefore, stresses remain
unchanged.
The proposed change will also not adversely affect HPCI and RCIC
system performance.
There is no indication that changed setpoints contribute to an
increase in probability of SRV malfunction. Reduction in the simmer
margin will be compensated for by more stringent leak test
requirements during valve refurbishment.
2. No. This change does not involve any hardware changes or
changes in system function. Relief and safety setpoints are only
slightly increased and the maximum safety setpoint remains
unchanged, thus the margin between peak allowable pressure and the
setpoint remains unchanged.
3. No. The technical specifications were reviewed for margins of
safety applicable to the components and systems affected by the
change. Analysis has been performed that demonstrates that reactor
pressure will be limited to within ASME Section III allowable values
for the worst case upset transient. The margin of safety is inherent
in the ASME Section III allowable pressure values.
Specification 3.4.3.2.d, Reactor Coolant System, Operational
Leakage
This specification is being revised to indicate that the 1 gpm
leakage rate limit currently applicable applies at the uprated
maximum allowable pressure of 1035 psig, plus or minus 10 psig.
1. No. The steam dome pressure for leakage is being increased by
35 psig to 1035 psig (reactor design pressure). This pressure is
chosen on the basis of steam line pressure drop characteristics and
excess steam flow capability of the turbine observed during plant
operation up to the current rated power level. Increasing the
leakage rate pressure to 1035 psig is consistent with the expected
uprated operating pressure. Increasing the reactor steam dome
pressure has been analyzed and found to be within allowable limits.
Maintaining the leakage rate limit at 1 gpm does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. No. This change does not involve any hardware changes or
change in safety function. The reactor steam dome pressure has been
analyzed and found to be within allowable limits.
3. No. Maintaining leakage the rate limit at 1 gpm is
conservative and does not involve a reduction in the margin of
safety.
Specifications 3.4.6.2 and 4.4.6.2, Reactor Coolant System,
Reactor Steam Dome
The reactor steam dome pressure limits have been changed to 1050
psig.
1. No. Operating pressure for uprated power is increased by a
minimum amount necessary to assure that satisfactory reactor
pressure control is maintained. The operating pressure was chosen on
the basis of steam line pressure drop characteristics and excess
steam flow capability of the turbine observed during plant operation
up to the current rated power level. Satisfactory reactor pressure
control requires an adequate flow margin between the uprated
operating condition and the steam flow capability of the turbine
control valves at their maximum stroke. An operating dome pressure
of 1032 psig is expected and is being assumed in the transient
analyses. The 1050 psig limit was chosen to maintain an adequate
level of operating flexibility while maintaining an adequate
distance from the high pressure scram for trip avoidance. This limit
is the initial pressure value used in the overpressure protection
safety analysis for power uprate, for which all licensing criteria
have been met. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. No. Based on the response to Question 1. above, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. No. As described in 1. above, the 1050 psig limit was chosen
to maintain an adequate level of operating flexibility while
maintaining an adequate distance from the high pressure scram. This
limit is the initial pressure value used in the over pressure
protection safety analysis for power uprate, for which all licensing
criteria have been met. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
Specification 4.5.1.b.3, Emergency Core Cooling Systems
This specification has been revised to permit a test line
pressure for the flow surveillance of greater than or equal to 1140
psig at nominal reactor operating conditions.
1. No. Currently, the HPCI pump test acceptance criteria
discharge pressure is greater than or equal to 1266 psig. This is
based, in part, on the lowest SRV setpoint of 1146 psig plus a 1%
tolerance and line flow losses. For this test, the HPCI turbine is
supplied with steam at the nominal operating reactor pressure of 920
+140/-20 psig. Therefore, the test requires the HPCI pump/turbine to
produce an output that exceeds that which would be commensurate with
the input conditions. Stated differently, HPCI would be required to
develop a pump discharge pressure associated with a steam dome
pressure of 1187 psig (1175 plus or minus 1% psig), while being
supplied with a steam dome pressure as low as 900 psig.
The purpose of this specification is to demonstrate that the
system is capable of producing the required flow at the required
pressure. The concern with this approach is that while it
demonstrates the required capability by achieving the actual
Technical Specification value, it requires the pump turbine to
``over perform''. It also reduces the margin available to compensate
for normal wear and tear [that] occurs and is monitored under the
ASME Section XI Pump and Valve Test Program. Power uprate will be
further increasing the demand because of the increase in reactor
steam dome pressure.
The intent of Surveillance 4.5.1b.3 is to demonstrate that the
HPCI System will produce its design flow rate at an expected reactor
pressure during a LOCA. Confirmation of the capability to achieve
the required flow and pressure can be satisfactorily demonstrated
without requiring the pump/turbine to ``over perform''. This can be
done by producing the nominal operating design pressure from the
pump with steam supplied to the turbine at nominal reactor operating
pressure. From these conditions extrapolation via pump affinity laws
will show the pump discharge pressure that would be developed at
emergency reactor operation conditions (i.e. lowest SRV setpoint).
This value could then be compared to the calculated value required
for assuring adequate core cooling in both SSES specific and generic
evaluations. The HPCI System has been evaluated and shown to be
capable of achieving the required pressure and flow conditions for
power uprate.
Applying the method of pump affinity laws, the new Technical
Specification pump discharge pressure would become greater than or
equal to 1140 psig. This value is determined based on the maximum
allowable test steam dome pressure of 920 + 140 = 1060 psig, plus
head losses. Through the use of pump affinity laws it has been shown
by calculation that achieving a value of 1140 psig at nominal
reactor operating conditions will produce the required flow and
pressure during emergency conditions.
Therefore, the Technical Specification HPCI pump discharge
pressure at power uprate conditions is changed to greater than or
equal to 1140 psig.
2. No. The methodology and the supporting change described above
in the response to Question 1 above do not alter the function nor
the operation of the HPCI system. Therefore, they do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. No. The methodology and the supporting change described above
in response to Question 1 do not involve a significant reduction in
a margin of safety.
Specification 5.4.2, Design Features, Reactor Coolant System,
Volume
This specification is being changed to show that the nominal
Tave is being changed from 528 deg.F to 532 deg.F. This change
is being made to reflect the higher average saturation temperature
that results from a 30 psi increase in reactor design pressure.
1. No. The effects of power uprate have been evaluated to ensure
that the increase in system temperatures causes minor increases in
thermal loadings on pipe supports, equipment nozzles, and in-line
components. The results of analyses show that at uprated conditions
all ASME components will satisfy design specification requirements
and code limits when evaluated to the rules of Subsection NB-3600 of
the ASME Boiler and Pressure Vessel Code Section III. The effects of
thermal expansion as a result of power uprate were found to be
insignificant. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. No. Increases in system temperatures as a result of power
uprate have been evaluated to show that increase in thermal loadings
on pipe supports, equipment nozzles and in-line components are
minor. Analysis shows that at all uprated conditions all ASME
components will satisfy design specification requirements and code
limits when evaluated to the rules of subsection NB-3600 of Section
IV to the Boiler and Pressure Vessel Code. The effects of power
uprate with respect to thermal expansion were found to be
insignificant and, therefore, not found to create the possibility of
a new or different kind of accident.
3. No. As stated above, the effects of thermal expansion as a
result of power uprate were found to be insignificant. Consequently,
the nominal increase in Tave does not involve a significant
reduction in a margin of safety.
Specification Table 5.7.1-1, Component Cyclic or Transient
Limits
This specification is being changed to raise the upper limit for
a heat cycle from 546 deg.F to 551 deg.F. This change is being made
to reflect the higher average saturation temperature that results
from a 30 psi increase in reactor design pressure.
1. No. The purpose of this specification is to limit the number
of heatup and cooldown cycles. The effects of power uprate have been
evaluated to ensure that the reactor vessel components continue to
comply with the existing structural requirements of the ASME Boiler
and Pressure Vessel Code. The analyses were performed for the
design, normal, upset, emergency and faulted conditions. The
increase in the temperature limitation is not significant with
respect to the affect it has upon the RPV and associated components.
2. No. The effects of uprating power have been evaluated for the
design, normal, upset, emergency and faulted conditions to ensure
that the reactor vessel components continue to comply with the
existing structural requirements of the ASME Boiler and Pressure
Vessel Code. The increase in the temperature limitation has been
found not to be significant and, therefore, does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. No. This specification is intended to limit the number of
heatup/cooldown cycles. The increase in the temperature limitation
has not been found to be significant with respect to its effects
upon the RPV and its associated components and, therefore, does not
significantly reduce the margin of safety.
Specification 6.9.3.2, Core Operating Limits Report
Administrative Control Section 6.9.3.2 describes and lists
topical reports that are used to determine core operating limits.
Topical reports 15 through 19 are LOCA methodology reports and are
being deleted. These reports describe Siemens LOCA methodology. As
stated in Reference 1, the GE SAFER/GESTR LOCA methodology is being
used for this uprated cycle. In addition, other minor methodology
changes were made for power uprate transient analysis. GE topical
report NEDC-32071P, PP&L topical report NE-092-001 and the NRC
Safety Evaluation Report on the PP&L power uprate licensing topical
are proposed to be added as Topical Reports No. 15, 16, and 17,
respectively.
1. No. These changes are editorial in nature in that only the
references to documents are being changed. The methodology used to
determine core limits have been previously reviewed and approved by
the NRC.
2. No. See the response to Question 1 above.
3. No. See the response to Question 1 above.
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: Osterhout Free Library, Reference
Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: Mohan C. Thadani, Acting
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: July 27, 1994
Description of amendment request: This amendment will change the
definition of a CORE ALTERATION included in Technical Specification
Section 1.0 for each unit to allow movement and replacement of local
power range monitors and control rods in a defueled cell. The new
definition is consistent with the Improved Standard Technical
Specifications.
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. In the submittal, the licensee stated that:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change eliminates two previous evolutions, LPRM and
Control Rod movement from a defueled cell, from being considered
CORE ALTERATIONS. Thus the issue is whether the elimination of these
constraints could contribute to a significant increase in the
probability or consequences of a reactivity event.
Adding local power range monitors to the list of detectors which
can be moved without invoking CORE ALTERATION requirements allows
for the removal of these detectors for repair and replacement.
Movement of these components does not impact the reactivity of the
core. Therefore, allowing the movement of these detectors without
invoking CORE ALTERATION provisions, does not contribute to a
significant increase in the probability or consequences of a
reactivity event.
Removal of a Control Rod from a defueled cell results in a
negligible increase in core reactivity. Appropriate Technical
Specification controls and refueling interlocks are applied during
the fuel movements preceding the control rod removal to protect from
or mitigate a reactivity excursion event. In addition, the design of
a control rod precludes its replacement without all fuel assemblies
in the cell removed. Therefore, allowing the movement of control
rods from a defueled cell without invoking CORE ALTERATION
provisions, does not contribute to a significant increase in the
probability or consequences of a reactivity event.
The proposed Technical Specification change to adopt the revised
CORE ALTERATION definition (NUREG 1433, as amended) does not effect
the probability or consequences of an accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change eliminates two previous evolutions, LPRM and
Control Rod movement from a defueled cell, from being considered
CORE ALTERATIONS. Thus the issue is whether the elimination of these
constraints could create the possibility of a new or different kind
of accident from any accident previously evaluated.
For local power range monitors, Technical Specification 3/4.3.1
defines the minimum number of LPRMs required to be maintained
operable in OPCON 5 and during Shutdown Margin Demonstration. The
addition of LPRMs as an exclusion under the CORE ALTERATION
definition does not change the operability requirements for the
LPRMs under Technical Specification 3/4.3.1. Thus the ability of the
LPRMs to perform their monitoring function is not affected by the
proposed CORE ALTERATION definition change. In addition, movement of
these components does not impact the reactivity of the core.
Therefore, allowing the movement of these detectors without invoking
CORE ALTERATION provisions, does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
For Control Rods, in the unlikely event that the wrong control
rod was inadvertently withdrawn from a fueled cell during evolutions
which were not intended to be CORE ALTERATIONS, adequate protective
measures are provided by design and core monitoring instrumentation
required to be operable in OPCON 5. Withdrawal of a single control
rod from a cell containing fuel is bounded by Shutdown Margin
analysis and demonstration. However, assuming the inadvertent
control rod withdrawal resulted in a significant reactivity
addition, the Reactor Protection System (RPS) would respond by
inserting all control rods via the Scram function. The RPS monitors
for recriticality during OPCON 5 with SRMs (except during specific
controlled evolutions), IRMs, and APRMs. The Scram circuitry is
completely redundant from the insert and withdrawal circuitry for
the control rods. Therefore, allowing the movement of control rods
from a defueled cell without invoking CORE ALTERATION provisions,
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed Technical Specification change to adopt the revised
CORE ALTERATION definition (NUREG 1433, as amended) does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
To evaluate the potential effect on safety margin, the proposed
change was evaluated as to its effect on Shutdown Margin. Shutdown
Margin defines the amount of reactivity by which the reactor is
subcritical, and thus is a measure of the safety margin in avoiding
unanticipated criticality events.
The movement of LPRMs does not impact the reactivity of the
core, and thus does not reduce the Shutdown Margin. Removal of a
Control Rod from a defueled cell results in a negligible increase in
core reactivity. Therefore, the removal of a Control Rod from a
defueled cell will have a negligible effect on the core Shutdown
Margin. Per Technical Specification 3/4.9.10.2(c), adequate core
Shutdown Margin must exist during refueling when multiple control
rods and the surrounding fuel assemblies are removed from the core.
Appropriate Technical Specification controls and refueling
interlocks are applied during the fuel movements preceding the
control rod removal to protect from or mitigate a reactivity
excursion event. In addition, the core is analyzed to maintain
Shutdown Margin even with the withdrawal of the highest worth rod
from a fueled cell.
The proposed Technical Specification change to adopt the revised
CORE ALTERATION definition (NUREG 1433, as amended) 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: Osterhout Free Library, Reference
Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: Mohan Thadani, Acting
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: July 20, 1994
Description of amendment request: The amendments would raise the
Steam Leakage Detection system set-points that isolate the High
Pressure Coolant Injection System (HPCI) and Reactor Core Isolation
Cooling (RCIC) system equipment on high equipment room temperature and
high delta temperature. The amendments are supported by a Limerick
Generating Station modification to increase the environmental
qualifications limits of the HPCI and RCIC systems to allow the systems
to remain operable when equipment room cooling is unavailable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
Those accident which are potentially impacted by these changes
are any accident or events that require the isolation of the HPCI or
RCIC system steam supply lines. This would include gross failures
(pipe breaks) or significant leaks (pipe cracks) in steam lines.
Minor leaks that do not significantly affect the environment in the
equipment compartments are only considered with regard to being
potential precursors to the development of a larger crack or break.
The ability to detect small steam leaks is not dependent on the
isolation instrumentation and the proposed changes to the isolation
instrumentation will not impact the detection methods.
The proposed TS changes will not increase the probability of an
accident since the changes will only increase the trip set-points of
the instrumentation which detect increases in the temperature in the
HPCI and RCIC equipment rooms. The physical establishment and
setting of the proposed set-points of these accident detection and
mitigation instruments will have no direct physical impact on the
plant's normal operating conditions. This instrumentation is
normally in a ``monitoring mode,'' and is not actively supporting
normal plant operation. Therefore, the proposed set-points can have
no impact on the operating plant that would make an accident more
likely to occur.
Two perspectives were evaluated regarding the potential impact
on the consequences of accidents. One case is the impact on
accidents which do not require HPCI or RCIC steam line isolation,
but that may require the operation of the HPCI or RCIC Systems. The
other case is the impact resulting from HPCI and RCIC steam line
break accidents.
In the first case, the proposed changes to the set-points of
these accident mitigation instruments will have no direct physical
impact on the plant's accident response, except during the HPCI or
RCIC pipe break accidents. During all other pipe breaks or
accidents, the bounding peak HPCI and RCIC equipment compartment
temperatures will still be at least 35 deg.F below the proposed TS
lower allowable values (i.e., 218 deg.F and 198 deg.F,
respectively), and the isolation instrumentation will remain in a
``monitoring mode.'' The isolation instrumentation will only be
required to continue to passively monitor the HPCI and RCIC
compartment temperatures and will meet the design basis by not
inadvertently isolating the HPCI or RCIC systems.
In the second case, the HPCI and RCIC pipe break accidents
described in LGS, Updated Final Safety Analysis Report (UFSAR)
Section 3.6 ``Protection Against Dynamic Effects Associated with the
Postulated Rupture of Piping,'' determine the peak pressures and
temperatures for the affected compartments. These peak pressures for
the HPCI and RCIC breaks are the bounding pressures for breaks in
these lines and, since they occur quickly, they are unaffected by
the leak detection and isolation actuation systems. The peak
pressures predicted in the UFSAR for the largest HPCI and RCIC steam
line breaks, in the HPCI, RCIC and isolation valve compartments, are
the bounding values for breaks of all sizes in these compartments.
In addition, the peak temperatures are not affected by the proposed
changes to the isolation actuation set-points. Therefore, the
isolation of the HPCI and RCIC steam lines following a HPCI or RCIC
steam line guillotine break is not dependent on the temperature trip
functions, rather, the isolation is dependent on the high flow or
low pressure trip functions where a delay in the response of the
temperature isolation instrumentation will have no adverse impact on
the consequences of the accidents described in the SAR.
An evaluation was performed to determine the potential impacts
due to the proposed changes affecting the room temperatures used in
the environmental qualification program. The results of this
evaluation determined that the postulated peak temperatures for the
HPCI pump room and the HPCI and RCIC piping areas would be at the
saturation temperature for the HPCI or RCIC break blow-down in these
compartments, therefore, these compartment temperatures values will
not be exceeded. The RCIC pump room and isolation valve compartment
environmental qualification temperatures were not postulated to be
at the saturation temperature. However, this does not increase the
consequences of any of the accident described in the SAR because the
equipment which is normally required for RCIC system operation and
which is located in the RCIC pump compartment is not required to
operate following breakage of the RCIC steam supply line. The only
equipment in the RCIC pump compartment that is required to operated
following a RCIC steam line break is the RCIC leak detection
instrumentation which are qualified to operate at temperatures
greater than the saturation temperature. Finally, the isolation
valve compartment postulated peak temperatures result from a HPCI
steam line break in the Unit 1 and 2 isolation valve compartments.
This line break produces the highest isolation valve compartment
temperatures which bounds the results of a RCIC steam line break in
the isolation valve compartment and the HPCI and RCIC steam lien
breaks in the HPCI and RCIC pump rooms and piping areas. However,
since the leak detection and isolation actuation trip set-points for
the instruments in the isolation valve compartment are not being
changed, then the environmental conditions in the isolation valve
compartment will remain unchanged. This will assure that the
isolation valves will be able to provide isolation when required.
For HPCI or RCIC leaks, the environmental conditions were not
the only design basis considerations evaluated. The radiological
affects were also considered. By increasing the upper allowable high
ambient temperature or high delta temperature values for certain
line break sizes there will be a larger total mass blow-down from
the break due to the corresponding lengthening of the time to reach
the higher temperature limit. However, the total integrated mass of
blowdown prior to isolation of the HPCI or RCIC steam line break
will still be bounded by the LGS UFSAR accident analysis and
therefore, the radiological consequences of these breaks as
described in the SAR will remain unchanged. These conclusions are
supported by an evaluation that provided the design basis for the
main steam line break and then examines the radiological
consequences at the upper and lower end of the HPCI and RCIC break
spectrum. Since the largest HPCI and RCIC breaks are isolated based
on high flow and not based on compartment temperature increases,
then the proposed changes in the temperature set-points have no
impact on the radiological consequences of the design basis HPCI or
RCIC pipe break accidents as described in the SAR.
The impact of the proposed changes on the probability of a
malfunction of the system isolation instrumentation, valves, or the
HPCI or RCIC systems was evaluated. The isolation actuation
instruments are qualified for the expected environmental conditions
and the proposed set-points are within the normal operating range of
the instruments. Therefore, these isolation actuation instruments
are more likely to randomly fail than before. In addition, by
ensuring that there is no adverse impact on the ability of the HPCI
or RCIC systems to respond to events which are caused by
malfunctions of equipment, then the consequences of these events are
not increased. An adequate margin between the proposed lower
allowable trip values and the postulated equipment room
environmental conditions is being maintained such that an
inadvertent actuation of the HPCI or RCIC system isolation function
is also no more likely to occur. The increase in the temperature
isolation allowable trip values will allow increased blow-down from
a pipe break or crack which will result in higher pump compartment
temperatures and pressures than before for a given break size;
however, the overall impact is still bounded by the LGS UFSAR
Section 3.6 ruptured piping analyses. The isolation actuation
instruments are qualified for the expected environmental conditions,
and the proposed set-points are also within the normal operating
range of the isolation instruments. Therefore, the instruments are
no more likely to randomly fail and cause the loss of the HPCI or
RCIC system than before. In fact, by increasing the qualification
limits of the HPCI and RCIC systems, the systems will be able to
remain operable with an even large steam leak in the room when room
cooling is available. Therefore, the changes will have no impact on
the operating plant that would increase the possibility or
consequences of a malfunction of equipment important to safety.
Since the proposed changes will maintain the HPCI or RCIC steam
isolation system design basis, where the consequences are bounded by
an analysis contained in the LGS UFSAR, and will only change the
set-points of the existing instrumentation without impacting
equipment important to safety, the proposed Technical Specifications
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes will not create the possibility of a
different type of accident or malfunction of equipment since the
changes will only increase the trip set-points of the
instrumentation which detect increases in the temperature in the
HPCI and RCIC equipment rooms. The physical establishment and
resetting of the set-points of these accident detection and
mitigation instruments will have not direct physical impact on the
plant's normal operating conditions and will not create any new
accident initiators or failure modes. The severity of the potential
piping system pressure transients caused by the isolation of the
HPCI or RCIC steam lines at higher room temperatures remains
unchanged since the isolation occurs after the postulated break
blow-down has dropped to its steady state rate. Therefore, the
changes will not result in a pipe break or result in any malfunction
of equipment that has not previously been postulated to occur.
Therefore, the proposed set-points will not create the
possibility of a different type of accident or possibility of a
different type of malfunction of equipment important to safety than
previously evaluated in the SAR.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety for the isolation actuation instrumentation
as defined in the TS bases is not reduced. The proposed system
isolation TS trip set-points were selected to provide equivalent
margins that ensure the effectiveness of the isolation systems to
mitigate the consequences of accidents without compromising the
operability of the HPCI and RCIC systems. The proposed trip set-
points and proposed allowable value ranges maintain adequate margins
between these new values and the operating range of the HPCI and
RCIC systems in order to prevent the inadvertent actuation of the
isolation system and the loss of either the HPCI or RCIC systems.
The differences between the trip set-points and the allowable values
are being maintained as an allowance for instrument drift. The trip
set-points and the allowable ranges are within the specified range
of the instruments and therefore, the accuracy and drift will
provide the same margin of safety as previously assumed.
Therefore, the proposed TS change 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 Room: Pottstown Public Library, 500 High
Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Mohan C. Thadani, Acting
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: July 22, 1994
Description of amendment request: This amendment would remove the
surveillance frequency details which govern 10 CFR 50, Appendix J, Type
B and C testing from Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes involve the removal of repetitious
surveillance details from TS also found in 10 CFR 50, Appendix J,
and rewording of TS. The removal and rewording involves no technical
changes to the existing TS. The changes to the existing TS are
proposed in order to be consistent with NUREG-1433. During the
development of NUREG-1433, certain wording preferences or English
language conventions were adopted. The proposed changes to this TS
section are administrative in nature and do not impact initiators of
analyzed events. They also do not impact the assumed mitigation of
accidents or transient events. Therefore, the changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a physical alteration of the
plant or changes in methods governing normal plant operation. The
proposed changes will not impose any new or different requirements
or eliminate any existing requirements. Therefore, the changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The changes are administrative in nature and will not involve
any technical changes. The proposed changes will not reduce a margin
of safety because they have no impact on any safety analysis
assumptions. In addition, because the changes are administrative in
nature, no question of safety is involved. 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 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: Pottstown Public Library, 500 High
Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: Mohan C. Thadani, Acting
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: August 19, 1994
Description of amendment request: This change would reduce the
minimum setpoints and allowable values for the Steam Generator Level -
Low-Low and Low reactor protection system signals. The bases would also
be modified to expand the description of the relationship between
setpoints, allowable values and the plant safety analysis.
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.
The Steam Generator Water Level--Low-Low signal and the Low
Steam Generator Level coincident with Steam Flow/Feed Flow Mismatch
signal are designed to mitigate design basis transients involving
significant reductions of steam generator inventory (e.g., Loss of
Normal Feedwater, Turbine Trip, Loss of Offsite Power, Feedwater
Line Break). The setpoints and allowable values for these protection
signals are prescribed by Technical Specifications such that
performance of the signals is consistent with the plant safety
analyses, considering the effects of channel uncertainties. The
proposed reductions to the setpoints and allowable values for the
low-low and low steam generator level signals would not affect the
probability of any transient that the protection signals are
designed to mitigate. The changes would reduce the probability of
unnecessary reactor trips and Auxiliary Feedwater (AFW) system
actuations by providing greater operating margin for plant
evolutions involving steam generator level changes (e.g., plant
startup). Therefore, the proposed changes do not involve any
increase in probability of an accident previously evaluated.
The changes to the Steam Generator Water Level--Low-Low signal
would not result in any increase in consequences of a previously
analyzed accident because the proposed setpoint and allowable value
would continue to ensure the safety analysis assumptions remain
valid. As described in the accompanying changes to the Technical
Specifications Bases, the channel uncertainty calculations performed
to establish the relationships between the setpoints, allowable
values and safety analyses are consistent with NRC Regulatory Guide
1.105, Revision 2. Low Steam Generator Level coincident with Steam
Flow/Feed Flow Mismatch signal is not credited in the UFSAR Chapter
15 safety analyses. The proposed changes to the low steam generator
level setpoint and allowable value would continue to provide
reliable backup to the low-low level trip signal, consistent with
IEEE-279-1971. Therefore, the proposed changes would not involve an
increase in 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.
The proposed changes would continue to ensure the appropriate
reactor protection system functions (reactor trip and AFW
initiation) are initiated in the event that steam generator water
level decreases to the value used in the plant safety analyses. The
proposed changes would not involve any changes in protection system
logic or function, and do not involve any plant configurations that
could adversely affect the initiation or progression of any accident
sequence. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3) do not involve a significant reduction in a margin of safety.
The proposed setpoints and allowable values would continue to
ensure that the assumptions in the safety analyses remain valid,
with appropriate consideration of protection system channel
uncertainties. Therefore, the proposed changes do not involve a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room: Salem Free Public library, 112 West
Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: Mohan C. Thadani, Acting
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 amendment request: July 20, 1994
Description of amendment request: The proposed change would modify
the Virgil C. Summer Nuclear Station (VCSNS) Technical Specification
(TS) Tables 2.2-1, ``Reactor Trip System Instrumentation Setpoints,''
and 3.3-4, ``Engineered Safety Features Actuation System
Instrumentation Trip Setpoints,'' and several associated bases. The
proposed change would remove three columns from the Tables. The columns
contain specific rack and sensor allowable drift values.
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 VCSNS in accordance with the proposed license
amendment does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
This change does not alter or delete any setpoints or Allowable
Values, and as such, has no affect on any assumptions used for
accident analysis. No hardware or software changes are involved, so
no common mode or common cause failures can occur as a result of
this change. This change has no impact on the daily operation of
VCSNS. The performance of periodic calibrations and channel checks
will assure the setpoints remain within tolerance. Since this
amendment request affects only information that is no longer used in
the daily operation of the plant and has no impact on accident
analysis, the probability or consequences of an accident previously
evaluated are not increased.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
This change revises two TS tables which contain both setpoints
and Allowable Values as well as other information for safety trip
functions. However, the revision only deletes three columns of data
that were used in determining the operability of one channel of the
safety function. These values are also used in determining the
setpoints and are based on measured or published tolerances and
uncertainties. Although these columns are being deleted, no changes
to any hardware, software, or setpoints will occur. Since these
changes do not have any plant impact, no new failure mechanisms are
introduced. Only the information not used on a daily basis is being
removed from these tables; this will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
This change revises the format of TS Tables 2.2-1 and 3.3-4
which list the setpoint and Allowable Values for safety trip
functions. The data that is being removed from these tables was used
to establish clear reportability requirements for any portion of one
channel of any of the listed safety trip functions. Since the
reporting requirements have changed and an LER is not required if
one coincident channel is inoperable, this data is no longer used in
daily operations. The margin of safety was established when
setpoints and Allowable Values were determined, and no changes to
these values are involved. There is no reduction in a margin of
safety that could affect the plant, SCE&G employees, or the public.
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: Fairfield County Library, Garden and
Washington Streets, Winnsboro, South Carolina 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: David B. Matthews
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 amendment request: July 20, 1994
Description of amendment request: The proposed change would modify
the Virgil C. Summer Nuclear Station, Unit 1, (VCSNS) Technical
Specifications (TS) to allow alternative, equivalent testing of diesel
fuel used in the emergency diesel generators (EDG). These alternative
methods are necessary due to recent changes in Environmental Protection
Agency (EPA) regulations that are designed to limit the use of high
sulfur fuels.
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 probability or consequences of an accident previously
evaluated is not significantly increased.
The change in testing methods for the EDG fuel oil has no impact
on the probability or consequences of any design basis accident.
These tests have been determined to be equivalent to the previously
approved testing methods and are needed due to changes in the EPA's
regulations regarding sulfur in motor vehicle fuels. The dye used to
identify high sulfur fuels will have no adverse affect on the
performance of the EDG's. The proposed testing assures a continued
high level of quality of the diesel fuel received and stored on
site.
The change in revision level of a reference in TS section
6.9.1.11 has no impact on the probability of occurrence or
consequences of any design basis accident. All design and
performance criteria will continue to be met and no new single
failure mechanisms will be created. The change in revision level for
WCAP-10216-P-A does not involve any alterations to plant equipment
or procedures which could affect any operational modes or accident
precursors. This change only incorporates by reference, the
methodology for determining the penalty to be used in calculating
Core Operating Limits. This methodology allows the penalty to be
cycle specific and is primarily affected by the core configuration.
This penalty is used for normal operation and provides more
conservatism to the core operation for the cycle.
2. [The proposed license amendment does not] create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The change in testing methods for the EDG fuel oil will not
create the possibility of a new or different kind of accident from
any accident previously evaluated. These tests have been determined
by the EPA and other organizations to be equivalent to the
previously approved testing methods. The effect of the blue dye,
used to identify high sulfur fuels, on the performance of the EDGs
has been evaluated and determined to be insignificant. The testing
proposed assures a continued high level of quality for the diesel
fuel received and stored on site.
The change of revision level of a reference in TS section
6.9.1.11 has no impact on the probability of occurrence or
consequences of any design basis accident. All design and
performance criteria will continue to be met and no new single
failure mechanisms will be created. The change in revision level for
WCAP-10216-P-A does not involve any alterations to plant equipment
or procedures which could affect any operational modes or accident
precursors. This change only incorporates, by reference, the
methodology for determining the penalty to be used in calculating
Core Operating Limits. This methodology allows the penalty to be
cycle specific and is primarily affected by the core configuration.
This penalty is used for normal operation and provides more
conservatism to the core operation for the cycle.
3. [The proposed license amendment does not] involve a
significant reduction in a margin of safety.
The change in testing methods for the EDG fuel oil will not
involve a significant reduction in a margin of safety. The proposed
testing methods have been determined to be equivalent to the
previously approved testing methods. The test for sulfur assures
that the sulfur content is within the allowable range for weight-
percent. The test for color and clarity assures that the fuel is
relatively free of water and particulate contaminants. The proposed
tests provide at least an equivalent level of quality and
repeatability for the fuel oil analysis, thus assuring that the
margin of safety is not reduced.
The change in revision level of a reference in TS section
6.9.1.11 does not change the proposed reload design or safety
analysis limits for each cycle reload core. The associated change to
WCAP-10216-P-A due to the revision will be specifically evaluated
using approved reload design methods. The larger penalty actually
provides for an increase in margin during certain burnup ranges.
Since the safety analysis limits are unaffected, and the cycle
specific analysis will show that the analysis limits are met, the
change proposed will have no adverse impact on 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: Fairfield County Library, Garden and
Washington Streets, Winnsboro, South Carolina 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: David B. Matthews
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: August 19, 1994 (TS 93-09)
Description of amendment request: The proposed change would revise
the implementation schedule for Amendment Nos. 182 and 174 from that
stated in the amendments when they were approved by the Commission by
letter dated May 24, 1994. As issued, the amendments reflected the
licensee's plans to implement the changes during the Unit 2 Cycle 6
refueling outage. However, the licensee has determined that
implementation would be more appropriate following the refueling outage
when both units are operating in 1995. No changes to the technical
specification pages other than those approved when the amendments were
issued are needed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
determined that the no significant hazards consideration exists. This
analysis was provided in the original submittal for the amendment from
the licensee dated October 1, 1993, and was used in the preparation of
the amendments. The licensee has determined that this analysis remains
valid for the proposed revision to the implementation dates and that
the changes do not constitute a significant hazard. The staff
previously issued the proposed finding in the Federal Register (59 FR
4947) and there were no public comments on the finding. This analysis
is reproduced as follows:
TVA has evaluated the proposed technical specification (TS)
change and has determined that it does not represent a significant
hazards consideration based on criteria established in 10 CFR
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance
with the proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed revision supports the implementation of design
logic and setpoint changes to the loss-of-power relaying. This
relaying is designed to ensure adequate voltage is available to
safety-related loads in order to enhance their operability and
support accident mitigation functions and to provide for auxiliary
feedwater (AFW) pump starts. The design changes alter relay logic
and delete unnecessary relaying, but do not change the diesel
generator (D/G) start and load-shedding actuations that result from
loss-of-power conditions. Therefore, no new actuations or functions
have been created; and because the existing and proposed functions
provide for accident mitigation considerations that are not the
source of an accident, the probability of an accident is not
increased. The deletion of the 6.9-kilovolt shutdown board normal-
feeder undervoltage relays actually reduces the potential for
inadvertent shutdown board blackouts as a result of short-duration
voltage transients or instrument failures.
The setpoints and time delays for loss-of-power functions have
been modified based on the guidelines developed by the Electrical
Distribution System Clearinghouse as evaluated and determined
through detailed analysis by TVA. This design is documented in TVA
Calculations SQN-EEB-MS-TI06-0008, 27DAT, and DS-1-2 and is
available for NRC review at the SQN site. The assigned values are
conservative settings that will ensure adequate voltage is supplied
to safety-related loads for accident mitigation and safety functions
under normal, degraded, and loss-of-offsite-power voltage conditions
with appropriate time delays to prevent damage to electrical loads
and minimize premature or unnecessary actuations. The identification
of loss-of-voltage conditions is enhanced by the design changes to
ensure the timely sequencing of loads onto the D/G and the
initiation of AFW pump starts for accident mitigation. Because there
are no reductions in safety functions resulting from the design
logic, setpoint, and time-delay changes to the loss-of-power
instrumentation and offsite dose levels for postulated accidents
will not be increased, the consequences of an accident are not
increased.
The applicable mode addition, TS 3.0.4 exclusion deletion, and
response time measurement clarification incorporated in the proposed
change do not affect plant functions. These changes reflect the
requirements that SQN has been maintaining and serve to clarify the
requirements to provide consistency of application and easier
understanding. The AFW footnote addition and bases revision only
clarify operability conditions that are consistent with the plant
design for the AFW pump and loss-of-power instrumentation. Because
there are no changes to plant functions or operations, these
revisions have no impact on accident probabilities or consequences.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
As described above, the loss-of-power instrumentation ensures
adequate voltage to safety-related loads by initiating D/G starts
and load shedding and provides for AFW pump starting, but is not
considered to be the source of an accident. Although the design
logic, setpoint, and time-delay actuation criteria have changed, the
output functions to various plant systems that actuate for load
shedding and D/G starts remain the same. Therefore, actuation
criteria have been affected, but not safety functions, and the TVA
evaluation has confirmed that the new design enhances the ability to
maintain adequate voltage to support safety functions. Since safety
functions have not changed and the new loss-of-power instrumentation
design continues to support operability of safety-related equipment,
no new or different accident is created.
The applicable mode addition, TS 3.0.4 exclusion deletion, and
response time measurement clarification, as well as the AFW
operability clarifications, do not affect plant functions and will
not create a new accident.
3. Involve a significant reduction in a margin of safety.
The proposed loss-of-power TS changes support design logic,
setpoint, and time-delay requirements that have been verified by TVA
analysis to provide acceptable voltage levels for safety-related
components. In determining the acceptability of these voltage
levels, the minimum voltage for operation as well as detrimental
component heating resulting from sustained degraded-voltage
conditions were considered. This design ensures that safety-related
loads will be available and operable for normal and accident plant
conditions. The applicable mode addition, TS 3.0.4 exclusion
deletion, response time measurement clarification, and AFW
operability clarifications provide enhancements to TS requirements
and do not affect plant functions. Therefore, no safety functions
are reduced by these changes and there is no reduction in the margin
of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room: Chattanooga-Hamilton County Library,
1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Philadelphia Electric Company, Public Service Electric and Gas
Company, Delmarva Power and Light Company, and Atlantic City
Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of amendment request: June 23, 1993
Brief description of amendment request: The amendments would revise
the licenses and the technical specifications to change the maximum
core power limit from 3293 MWt to 3458 MWt.Date of publication of
individual notice in Federal Register: August 29, 1994 (59 FR 44432)
Expiration date of individual notice: September 28, 1994
Local Public Document Room: Government Publications Section, State
Library of Pennsylvania, (REGIONAL DEPOSITORY) Education Building,
Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
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 amendments request: August 9, 1994
Brief description of amendments request: These amendments revise
the Technical Specifications (TS) 5.3.4, ``Steam and Power Conversion
Systems,'' and 15.3.7, ``Auxiliary Electrical Systems,'' to increase
the allowed outage times for one motor driven auxiliary feedwater pump
and for the standby emergency power for the Unit 1, Train B4160 Volt
safeguards bus (A06) from 7 to 12 days. The proposed amendments would
also modify TS 15.3.3, ``Emergency Core Cooling System, Auxiliary
Cooling Systems, Air Recirculation Fan Coolers, and Contained Spray,''
to provide the clarification that the service water pump (P-32E)
operating with power supplied by the Alternative Shutdown System is
operable from offsite power. The changes are one-time extensions of
specific allowed outage times.Date of publication of individual notice
in the Federal Register: August 19, 1994 (59 FR 42870).
Local Public Document Room: Joseph P. Mann Library, 1516 Sixteenth
Street, Two Rivers, Wisconsin 54241.
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 20555, and at the local public document
rooms for the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: November 3, 1993
Brief description of amendments: The amendments revise the Calvert
Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Technical Specifications
(TSs) by removing the TSs that are applicable to the incore instrument
(ICI) system. The limitations on the use of the ICI system will be
relocated to the Updated Final Safety Analysis Report. The core power
distribution limits, which the ICI system is used to verify, remain in
the TSs which is consistent with 10 CFR 50.36.Date of issuance: August
24, 1994Effective date: As of the date of issuance to be implemented
within 30 days.
Amendment Nos.: 191 and 168
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64601) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 24, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Calvert County Library, Prince
Frederick, Maryland 20678.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: November 3, 1993
Brief description of amendments: The amendments modify the
surveillance requirements to reflect the removal of the auto-closure
interlock from the shutdown cooling system and revises the setpoint for
the open permissive interlock.
Date of issuance: August 24, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 192 and 169
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64600) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 24, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Calvert County Library, Prince
Frederick, Maryland 20678.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: May 27, 1994
Brief description of amendments: The amendments revise the
Technical Specification surveillance test intervals from monthly to
quarterly for several channel functional tests for the Reactor
Protection System and the Engineered Safety Feature Actuation System.
In addition, an administrative change was made to remove an out-of-date
footnote concerning the Emergency Diesel Generator logic circuit
modifications.
Date of issuance: August 24, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 193 and 170
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37062) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 24, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Calvert County Library, Prince
Frederick, Maryland 20678.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: November 5, 1993, as
supplemented March 11, 1994
Brief description of amendments: The amendments consist of two
related changes. The first change revises the containment penetration
Technical Specifications (TSs) to resemble the containment penetration
TSs in NUREG-1432, ``Standard Technical Specifications for Combustion
Engineering Pressurized Water Reactors.'' The second revises the TSs to
allow the containment personnel airlock to be open during fuel movement
and core alterations. The TS Bases have also been revised to reflect
the changes as the result of issuing these amendments.
Date of issuance: August 31, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 194 and 171
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64602) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated August 31, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Calvert County Library, Prince
Frederick, Maryland 20678.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of application for amendment: June 16, 1994
Brief description of amendment: The amendment removes from
Technical Specification 3/4.8.3, ``Onsite Power Distribution,'' a
footnote applicable for Cycle 18 only, and adds surveillance
requirement 4.8.3.1.2, to test the MCC-5 automatic bus transfer feature
once per refueling.
Date of Issuance: August 23, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 176
Facility Operating License No. DPR-61. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37067) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated August 23, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room: Russell Library, 123 Broad Street,
Middletown, Connecticut 06457.
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: April 28, 1994
Brief description of amendments: The amendments revised Technical
Specification 4.6.1.3.e to add an option that will allow the personnel
airlock pneumatic system leak test to be completed in 8 hours with a
pressure drop of 0.50 psi.
Date of issuance: August 29, 1994
Effective date: August 29, 1994
Amendment Nos.: Unit 1 - Amendment No. 64; Unit 2 - Amendment No.
53
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27057) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 29, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Wharton County Junior College, J. M.
Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: November 15, 1993
Brief description of amendments: The amendment revises the
Technical Specifications to extend the surveillance interval for the
chemical analysis, inventory, and flow area of the ice condenser from 9
to 18 months.
Date of issuance: August 23, 1994
Effective date: August 23, 1994
Amendment Nos.: 180 & 164
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67849) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 23, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room: 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, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: September 24, 1992 and
supplemented March 2, 1994.
Brief description of amendments: The amendment removes the list of
containment isolation valves and associated references to the list from
the Technical Specifications.
Date of issuance: August 29, 1994
Effective date: August 29, 1994
Amendment Nos.: 181 and 165
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 17, 1993 (58
FR 8773) The March 2, 1994, letter provided supplemental information
that was not outside the scope of this initial notice. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated August 29, 1994.No significant hazards consideration
comments received: No.
Local Public Document Room: Maud Preston Palenske Memorial Library,
500 Market Street, St. Joseph, Michigan 49085.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: July 1, 1994
Brief description of amendment: The amendment revises the secondary
containment drawdown time testing requirement of Technical
Specification (TS) 4.6.5.1.c.1 and the secondary containment inleakage
testing requirement of TS 4.6.5.1.c.2. The amendment supports a revised
design basis radiological analysis which supports an increase in
secondary containment drawdown time from 6 to 60 minutes by taking
credit for fission product scrubbing and retention in the suppression
pool which were not assumed in the original radiological analysis but
are currently assumed in the NRC's Standard Review Plan (NUREG-0800).
The revised analysis also takes credit for additional mixing of primary
containment and engineered safety feature system leakage with 50
percent of the secondary containment free air volume prior to the
release of radioactivity to the environment. The revised radiological
evaluation has determined that the radiological doses remain below 10
CFR Part 100 guideline values and General Design Criterion 19 criteria.
Date of issuance: August 30, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 56
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37074) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 30, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Reference and Documents Department,
Penfield Library, State University of New York, Oswego, New York 13126.
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Dates of application for amendment: November 30, 1993 and June 30,
1994.
Brief description of amendment: The proposed amendment would delete
the requirements for a chlorine detection system from the following
sections of Technical Specifications: 3.2.I, 3.17.A, 4.17.A, tables
4.2.1 and Technical Bases 3.2 and 3.17.A. Due to design changes at the
Monticello Nuclear Generating Plant, chlorine is no longer stored
onsite as a liquified gas and regulations requiring early warning of an
onsite chlorine release do not apply.
Date of issuance: August 25, 1994
Effective date: August 25, 1994
Amendment No.: 89
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10010) The June 30, 1994, letter provided documents cited in the
amendment application and did not affect the staff's initial no
significant hazards determination. The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated August 25,
1994.No significant hazards consideration comments received: No.
Local Public Document Room: Minneapolis Public Library, Technology
and Science Department, 300 Nicollet Mall, Minneapolis, Minnesota
55401.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: May 21, 1993, as supplemented by letters
dated September 10, 1993, and May 25, 1994
Brief description of amendment: The amendment changed the Technical
specifications to reflect the relocation of the old 10 CFR 20.106
requirements to the new 10 CFR 20.1302, and to implement administrative
changes.
Date of issuance: August 24, 1994
Effective date: August 24, 1994
Amendment No.: 164
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36442) The additional information contained in the supplemental letters
dated September 10, 1993, and May 25, 1994, was clarifying in nature
and thus, within the scope of the initial notice and did not affect the
staff's proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated August 24, 1994.No significant hazards
consideration comments received: No.
Local Public Document Room: W. Dale Clark Library, 215 South 15th
Street, Omaha, Nebraska 68102
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: February 12, 1993, as supplemented by
letters dated August 20, 1993, and June 6, 1994
Brief description of amendment: This amendment revised Technical
Secification 2.1.4, ``Reactor Coolant System Leakage Limits,'' to
implement the reactor coolant system leak-before-break methodology
detection criteria. Additionally, administrative changes were made.
Date of issuance: August 25, 1994
Effective date: August 25, 1994
Amendment No.: 165
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37076) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 25, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room: W. Dale Clark Library, 215 South 15th
Street, Omaha, Nebraska 68102.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of application for amendments: December 8, 1993 (Ref. LAR 93-
07)
Brief description of amendments: The amendments revise the combined
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit
Nos. 1 and 2 to revise TS 3/4.8.1, ``A.C. Sources'' to increase the
required quantity of emergency diesel generator (EDG) fuel oil stored
in the engine-mounted tank (day tank) from 200 gallons to 250 gallons.
The amendment also revises TS 3/4.7.11, ``Area Temperature
Monitoring,'' and 3/4.8.1 to remove references to a five EDG
configuration, based on the installation of a sixth EDG.
Date of issuance: August 23, 1994
Effective date: August 23, 1994
Amendment Nos.: 93 and 92
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 16, 1994 (59
FR 7694) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 23, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Philadelphia Electric Company, Docket No. 50-352, Limerick
Generating Station, Unit 1, Montgomery County, Pennsylvania
Date of application for amendment: June 6, 1994
Brief description of amendment: This amendment removes the controls
for a remote shtudown system control valve and deletes the isolation
signal for certain primary containment isolation valves from TS Tables
3.3.7.4-1 and 3.6.3-1 respectively, as a result of eliminating the
steam condensing mode of the Residual Heat Removal system.
Date of issuance: August 23, 1994
Effective date: August 23, 1994
Amendment Nos. 74
Facility Operating License No. NPF-39: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37076) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 23, 1994. No significant
hazards consideration comments received: No
Local Public Document Room: Pottstown Public Library, 500 High
Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: January 10, 1994, as
supplemented by letter dated July 20, 1994
Brief description of amendments: The amendments relocate the
seismic monitoring instrumentation Limiting Condition for Operation,
Surveillance Requirements, and associated tables and Bases contained in
TS Sections 3.3.7.2 and 4.3.7.2 to the Updated Final Safety Analysis
Report, Section 3.7.4.
Date of issuance: August 29, 1994
Effective date: August 29, 1994
Amendment Nos. 75 and 36
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12364) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 29, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Pottstown Public Library, 500 High
Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Public Service Electric and Gas
Company Delmarva Power and Light Company, and Atlantic City
Electric Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic
Power Station,Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: March 28, 1994, as supplemented
on June 27, 1994 and July 8, 1994
Brief description of amendments: These amendments relocate the fire
protection requirements from the Technical Specifications to the
Updated Final Safety Analysis Report in accordance with the guidance in
Generic Letter (GL) 86-10, ``Implementation of Fire Protection
Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements
from Technical Specifications.''
Date of issuance: August 24, 1994
Effective date: August 24, 1994
Amendments Nos.: 194 and 198
Facility Operating License Nos. DPR-44 and DPR-56: Amendments
revised the Technical Specifications and the licenses.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22012) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 24, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Government Publications Section, State
Library of Pennsylvania, (REGIONAL DEPOSITORY) Education Building,
Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
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: February 3, 1994
Brief description of amendment: The licensee commenced operating on
a 24-month fuel cycle, instead of the previous 18-month fuel cycle,
with fuel cycle 9. Fuel cycle 9 started in August 1992; however, the
facility has been shut down since February 1993 for a ``Performance
Improvement Outage'' and a restart date has not yet been established.
In order to accommodate operation on a 24-month cycle after the
facility restarts, the following Engineered Safety Features (ESF)
instrument calibration intervals have been extended:
(1) Reactor coolant temperature instrument channels (specified in
TS Table 4.1-1)
(2) Steam generator level instrument channels (specified in TS
Table 4.1-1)
(3) Containment pressure instrument channels (specified in TS Table
4.1 1)
(4) Steam line pressure instrument channels (specified in TS Table
4.1-1)
(5) Turbine first stage pressure instrument channels (specified in
TS Table 4.1-1)
(6) Turbine trip low auto stop oil pressure instrument channels
(specified in TS Table 4.1-1)
(7) 480V bus undervoltage and alarm relays (specified in TS Table
4.1-1)
These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to
Accommodate a 24-Month Fuel Cycle,'' as applicable.Additionally, the
following changes were also incorporated:
(8) A limiting conditions for operation requirement for a wide
range containment pressure variable was added to TS Table 3.5-5 to
ensure consistency with Regulatory Guide 1.97 commitments and the IP3
Emergency Operating Procedures (EOPs).
(9) A quarterly functional test surveillance requirement for the
low average temperature actuation circuits of the reactor coolant
temperature channels was added to Item 4 of TS Table 4.1-1.
(10) Item 14 of TS Table 4.1-1 was expanded to specify surveillance
requirements for the wide range containment pressure instrumentation
channels.
(11) Item 20 to TS Table 4.1-1 was revised to clarify that both the
reactor trip and the ESF actuation relay logic channels are
functionally tested.
Date of issuance: September 1, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 150
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14894) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 1, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: White Plains Public Library, 100
Martine Avenue, White Plains, New York 10610.
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: June 17, 1993, as supplemented
February 24, 1994, and June 13, 1994
Brief description of amendment: The amendment adds Section 3/
4.2.J., ``Remote Shutdown Capability,'' and associated Table 3.2-10,
``Remote Shutdown Capability Instrumentation and Controls,'' to the
Technical Specifications (TSs) to provide Limiting Conditions for
Operation and surveillance requirements for the remote/alternate
shutdown equipment. The amendment also adds an associated Bases section
to the TSs. These additions to the TSs were based on NUREG-1433,
``Standard Technical Specifications - General Electric Boiling Water
Reactors (BWR/4).'' Several administrative changes were also made to
accommodate the additions to the TSs.
Date of issuance: August 31, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 216
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41511) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 31, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Reference and Documents Department,
Penfield Library, State University of New York, Oswego, New York 13126.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: March 4, 1994, as supplemented
on June 14, 1994 and by phone on July 22, 1994
Brief description of amendments: These amendments modify Section
5.3.1 of the Technical Specifications (TS) to allow the use of
Westinghouse Vantage+ fuel with ZIRLO cladding. The previous TS
required the fuel cladding to be Zircaloy-4, which is used in the
Westinghouse Standard and Vantage 5H fuel designs.
Date of issuance: August 22, 1994
Effective date: August 22, 1994
Amendment Nos. 154 and 134
Facility Operating License Nos. DPR-70 and DPR-75. These amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14896) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 22, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Salem Free Public Library, 112 West
Broadway, Salem, New Jersey 08079
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: December 13, 1993, as
supplemented February 2, 1994, and March 11, 1994.
Brief description of amendment: The amendment changes the Technical
Specifications to allow for the storage of fuel with an enrichment not
to exceed a nominal 5.0 weight percent (w/o) U-235 in the VCSNS new
(fresh) and spent fuel storage racks. The changes would also allow
UO2 with a maximum nominal enrichment up to 5.0 w/o U-235 to be
used as fuel in the VCSNS core.
Date of issuance: August 23, 1994
Effective date: August 23, 1994
Amendment No.: 116
Facility Operating License No. NPF-12. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12365) The March 11, 1994, letter provided clarifying information that
did not change the initial determination of no significant hazards
consideration as published in the Federal Register. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated August 23, 1994. No significant hazards consideration comments
received: No
Local Public Document Room: Fairfield County Library, Garden and
Washington Streets, Winnsboro, South Carolina 29180.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: May 20, 1994
Brief description of amendment: The proposed amendment would remove
Core Spray High Sparger Instrumentation from the Vermont Yankee
Technical Specifications for Emergency Core Cooling System Actuation
Instrumentation. In addition, an unrelated administrative change is
also made.
Date of issuance: August 22, 1994
Effective date: August 22, 1994
Amendment No.: 140
Facility Operating License No. DPR-28. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34669) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 22, 1994.No significant
hazards consideration comments received: No
Local Public Document Room: Brooks Memorial Library, 224 Main
Street, Brattleboro, Vermont 05301.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: June 9, 1994
Brief description of amendments: These amendments revise the NA-1&2
Technical Specifications (TS) by removing the Reactor Trip System and
the Engineered Safety Features Actuation System response times from the
TS to station-controlled documents.
Date of issuance: August 24, 1994
Effective date: August 24, 1994
Amendment Nos.: 187 and 168
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37088) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 24, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: January 6, 1994
Brief description of amendment: This amendment relocates the
requirements related to seismic monitoring instrumentation from the
Technical Specifications (TS) to the Final Safety Analysis Report
(FSAR) and plant procedures. The existing requirements will be
maintained and controlled in accordance with the requirements of 10 CFR
50.59 and TS 6.8.1.
Date of issuance: August 22, 1994
Effective date: August 22, 1994, to be implemented within 30 days
of issuance
Amendment No.: 131
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14902) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 22, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room: Richland Public Library, 955 Northgate
Street, Richland, Washington 99352.
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: September 29, 1993.
Brief description of amendments: The amendments changed the
inservice test frequency of the safety injection pumps, residual heat
removal pumps, and containment spray pumps from monthly to quarterly.
Also, the amendments added the administration of the inservice testing
program to TS 15.4.2. The amendments added requirements to verify the
containment sump suction is not blocked and to verify on a monthly
basis, valve alignments of the emergency core cooling system and
containment cooling systems.
Date of issuance: August 25, 1994
Effective date: Date of issuance to be implemented within 45 days
Amendment Nos.: 150 and 154
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4949) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 25, 1994.No significant hazards
consideration comments received: No.
Local Public Document Room: Joseph P. Mann Library, 1516 Sixteenth
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: October 6, 1992
Brief description of amendments: The amendments changed all
references of rod position in the Technical Specifications to units of
steps rather than inches. The amendments also changed Figure 15.3.10-1
by referencing rod position in units of steps instead of percent
withdrawn. Further, the amendments revised the basis for Section
15.3.10 by clarifying the definition of ``fully withdrawn'' as it
concerns Rod Cluster Control Assemblies, and modified the basis for
Section 15.3.10 to be consistent with the above changes.
Date of issuance: August 26, 1994
Effective date: Immediately, to be implemented within 45 days.
Amendment Nos.: 151 and 155
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 25, 1993 (58 FR
16234) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 26, 1994No significant
hazards consideration comments received: No.
Local Public Document Room: Joseph P. Mann Library, 1516 Sixteenth
Street, Two Rivers, Wisconsin 54241.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: June 7, 1994
Brief description of amendment: The amendment revises Technical
Specification Table 2.2-1, ``Reactor Trip System Instrumentation
Setpoints,'' to change the over-temperature-delta-temperature (OTDT)
axial flux difference (AFD) limits to reflect the results of the Cycle
8 core maneuvering analysis.
Date of issuance: August 25, 1994
Effective date: August 25, 1994
Amendment No.: 79
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34672) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated Augusty 25, 1994.No significant
hazards consideration comments received: No.
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.
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,
and at the local public document room for the particular facility
involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By October 14, 1994, 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 20555 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. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, 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 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: June 9, 1994, as supplemented
August 10, 1994
Brief description of amendment: This amendment increases the
allowed out-of-service time from 7 days to 14 days for the automatic
depressurization system, the high pressure coolant injection system,
and the reactor core isolation cooling system. A change is also made to
Section 4.5.H, ``Maintenance of Filled Discharge Pipe'' to reflect
Amendment 149 issued September 28, 1993.
Date of issuance: August 22, 1994
Effective date: August 22, 1994
Amendment No.: 156
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: No The Commission's related
evaluation of the amendment, consultation with the State, and final
determination of no significant hazards consideration are contained in
a Safety Evaluation dated
Local Public Document Room: Plymouth Public Library, 11 North
Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Walter R. Butler
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley
Power Station, Unit 2, Shippingport, Pennsylvania
Date of application for amendment: August 17, 1994
Brief description of amendment: The amendment changes the Technical
Specifications (TS) by revising Surveillance Requirement (SR) 4.6.2.2.d
of Limiting Condition For Operation (LCO) 3.6.2.2, entitled
``Containment Recirculation Spray System,'' by adding a new footnote
number (1) pertaining to 2RSS*P21A pump performance requirements. In
addition, SR 4.6.2.2.e.2 is revised by deleting the footnote, denoted
by a single asterisk, which pertains to an extension to the 18-month
surveillance interval for first fuel cycle.
Date of issuance: August 22, 1994
Effective date: As of the date of issuance.
Amendment No: 62
Facility Operating License No. NPF-73. Amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: No. On August 17, 1994, the staff
issued enforcement discretion, which was immediately effective and
remained in effect until the staff's review of this amendment was
completed.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, consultation with the Commonwealth of
Pennsylvania and final no significant hazards considerations
determination are contained in a Safety Evaluation dated August 22,
1994.
Local Public Document Room: B. F. Jones Memorial Library, 663
Franklin Avenue, Aliquippa, Pennsylvania 15001.
Dated at Rockville, Maryland, this 7th day of September 1994.
For The Nuclear Regulatory Commission
Jack W. Roe,
Director, Division of Reactor Projects - III/IV, Office of Nuclear
Reactor Regulation
[Doc. 94-22593 Filed 9-13-94; 8:45 am]
BILLING CODE 7590-01-F
