[Federal Register Volume 62, Number 38 (Wednesday, February 26, 1997)]
[Notices]
[Pages 8790-8807]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-4573]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from February 1, 1997, through February 13, 1997.
The last biweekly notice was published on February 12, 1997 (62 FR
6567).
[[Page 8791]]
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By March 28, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
[[Page 8792]]
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts.
Date of amendment request: January 30, 1997.
Description of amendment request: The proposed amendment would
change the Updated Final Safety Analysis Report (FSAR) to include the
credit for containment overpressure in the Pilgrim Nuclear Power
Station net positive suction head (NPSH) analysis for the emergency
core cooling pumps.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will crediting post-LOCA [loss-of-coolant accident] wetwell
airspace pressure in ECCS [emergency core cooling system] analyses
involve a significant increase in the probability or consequences of
an accident previously evaluated?
Chapter 14 of the FSAR contains evaluations of the design basis
accidents, which include the refueling accident, the main steam line
break outside primary containment, the recirculation line break
inside primary containment, and the control rod drop accident. No
increase in the probability of the evaluated accidents will result
from crediting the post-LOCA wetwell airspace pressure because post-
LOCA wetwell airspace pressure does not represent an accident
initiator but is rather a byproduct of the conditions which will
exist in the containment after the pipe break inside containment.
The worst radiological consequences for the Pilgrim plant are
associated with the design basis LOCA which is the double guillotine
failure of the recirculation system piping. The radiological
analysis of this event, contained in FSAR Chapter 14, uses a TID-
14844 source term and assumes a 1.5% per day leakage from the
containment, which is greater than the maximum leakage allowed by
the Technical Specifications. The results of this analysis are
presented in Table 14.5-2 of the FSAR and indicate substantial
margin when compared to 10 CFR Part 100 limits.
The radiological consequences of the design basis accident are
not increased by taking credit for the post-LOCA wetwell airspace
pressure. Assuming containment integrity exists, the mechanism for
increasing the consequences of the accident would be an increased
leakage rate caused by an increase of the average differential
pressure between primary and secondary containment during the
accident response. However, the NPSH analyses performed for Pilgrim,
which credits the post-LOCA wetwell airspace, does not require that
the differential pressure between primary and secondary containment
be maintained above the minimum that exists due to the equilibrium
conditions based on the suppression pool temperature. Specifically,
the wetwell airspace pressure credited in the ECCS pump NPSH
analyses is provided by an increase in wetwell vapor pressure and
air/nitrogen partial pressure in equilibrium with increasing pool
temperature with an accounting for containment initial conditions
and leakage.
By crediting the post-LOCA wetwell airspace pressure in the
calculation of NPSH, no requirement is created to purposely maintain
a higher containment pressure than would otherwise occur; no
requirement is incurred to delay operating containment heat removal
equipment at the highest rate possible; no requirement is incurred
to deliberately continue any condition of high containment pressure
to maintain adequate NPSH; and no requirement is incurred for the
purposeful addition of air/nitrogen into the containment to increase
the available pressure.
Based on these reasons, the probability of accidents previously
evaluated is not increased, and the consequences of the design basis
accident are not increased.
(2) Will crediting post-LOCA wetwell airspace pressure create
the possibility for new or different kinds of accidents?
As stated above, Chapter 14 of the Pilgrim FSAR contains
evaluations of design basis accidents that include the refueling
accident, the main steam line break outside primary containment, the
recirculation line break inside primary containment, and the control
rod drop accident. New or different types of accidents are not
created by crediting the post-LOCA wetwell airspace pressure because
post-LOCA wetwell airspace pressure does not represent an accident
initiator but is rather a byproduct of the conditions which will
exist in the containment after the pipe break inside containment.
Therefore, crediting post-LOCA wetwell airspace pressure does
not create the possibility for new or different kinds of accidents
from those previously analyzed.
(3) Will crediting post-LOCA wetwell airspace pressure in ECCS
NPSH analyses involve a significant reduction in a margin of safety?
The integrity of the primary containment and the operation of
the ECCS systems in combination limit the off-site doses to values
less than those suggested in 10 CFR 100 in the event of a break in
the primary system piping. In order for the ECCS pumps to meet their
performance requirements, the NPSH available to the pumps throughout
the accident response must meet their specific NPSH requirements.
Excess NPSH margin will not improve the performance of the ECCS
pumps because NPSH available must only meet NPSH requirements for
the pump to operate on its pump curve and meet design expectations.
Crediting post-LOCA wetwell airspace pressure in ECCS NPSH
analyses increases the NPSH available to the pumps connected to the
suppression pool but limits the increase in NPSH available
consistent with the bounding leakage assumptions for the containment
system. The amount of post-accident pressure that is utilized in
ECCS NPSH analyses is calculated in a manner such that the pressure
credited represents a conservative lower bound of the pressure
available. Therefore, it is expected that the NPSH margin will
exceed that credited in the NPSH analyses.
Credit for wetwell airspace pressure in NPSH analyses is not
required under all circumstances. If the suction strainers for the
ECCS pumps remain relatively free of post-LOCA debris, adequate NPSH
will be available without credit for the wetwell airspace pressure
provided by the post-LOCA heatup of the air/nitrogen gas in the
containment. If debris accumulates on the pump suction strainers,
the NPSH available to the ECCS pumps will be decreased due to the
head loss caused by the debris. Credit for the post-LOCA wetwell
airspace pressure in the analyses indicates that there is adequate
NPSH margin such that NPSH available will remain above NPSH
required, and ECCS pump performance will meet applicable
requirements. Based on the above discussion, credit for wetwell
airspace pressure in ECCS NPSH analyses 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.
[[Page 8793]]
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: 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: Patrick D. Milano, Acting.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County,
North Carolina.
Date of amendments request: November 1, 1996.
Description of amendments request: The amendments would revise the
Technical Specifications (TS) to allow full implementation of the
Boiling Water Reactor Owners Group (BWROG) Enhanced Option 1-A Reactor
Stability Long Term Solution. In Safety Evaluation Reports (SERs)
transmitted to Kevin P. Donovan, Chairman, BWROG, by letters from
Robert C. Jones, Office of Nuclear Reactor Regulation, NRC, dated June
21, 1996, and September 20, 1996, the NRC staff concluded that Enhanced
Option 1-A generic technical specifications described in Topical Report
NEDO-32339, Supplement 4, were acceptable for referencing in license
applications.
The characteristics of a reactor system most important in
determining stability performance are power, core flow and power
distribution. The proposed changes would delete the current limits on
power and flow conditions in the technical specifications associated
with the implementation of the guidance in General Electric Service
Information Letter (SIL) #380, Revision 1 and the power/flow figure
(Figure 3.4.1.1-1), add two new specifications on the fraction of core
boiling boundary (FCBB) and the Period Based Detection System (PBDS)
and relocate certain requirements pertaining to the Average Power Range
Monitors (APRM) to the Core Operating Limits Report (COLR).
The current Technical Specifications for Units 1 and 2 permit
single loop operation (SLO) only for a 12-hour period and there are no
provisions for potential alterations of safety limits or operating
limits because of SLO conditions. Approval of the amendment
applications discussed above would permit SLO operation subject to the
compensatory actions and requirements that address this mode of
operation in the revised Technical Specifications. However, Brunswick
Unit 2's License currently has a condition, 2.C.(5) that states that
the reactor shall not be made critical unless both recirculation loops
are in service. This License Condition also requires the plant to be
placed in the hot shutdown condition within 24 hours if one
recirculation loop becomes out-of-service. The License Condition also
allows one or both recirculation loops to be out-of-service for the
purposes of testing (not to exceed 24 hours). Whereas the License
Condition would permit SLO for up to 24 hours, the current TS limit SLO
to 12 hours. The License Condition was added to permit natural
circulation testing as required by the startup test program but to
preclude long-term SLO or operation in the natural circulation mode.
The startup test program was completed many years ago for Brunswick
Unit 2 and natural circulation operation is no longer allowed. The
License Condition is no longer relevant and if not deleted would negate
the objectives of the proposed license amendments discussed above. The
licensee has submitted proposed license amendments on the same date of
the subject application (i.e., November 1, 1996) to convert the
Brunswick Units 1 and 2 Technical Specifications to the Improved
Standard Technical Specifications (ISTS) consistent with NUREG-1433,
Revision 1, ``Standard Technical Specifications for General Electric
Plants, BWR 4.'' Attachment 6 of the later application was a proposed
revision of the Brunswick Unit 2 License to delete License Condition
2.C.(5). While the Notice of Consideration of Issuance of the ISTS
amendments (62 FR 3719) discussed deletion of License Condition
2.C.(5), the deletion is discussed in this Notice as well, since if the
subject amendment applications are approved, the License Condition
would thwart the considerable effort represented by the subject
amendments to finally resolve the thermal-hydraulic stability issues
for Brunswick Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendments allow the implementation of the Enhanced
Option 1-A (E1A) long term solution to the neutronic/thermal
hydraulic instability issue. Current Technical Specification
restrictions on power and flow conditions, number of operating
recirculation loops and operator actions implemented to reduce the
probability of neutronic/thermal hydraulic instability are
eliminated and new stability control requirements consistent with
NEDO-32339, Supplement 4, are imposed. These requirements include
restrictions on power and flow conditions and actions associated
with the modified APRM flow biased scram and control rod block
functions. These actions include adherence to the boiling boundary
limit stability control prior to entry and during operation in the
region of the power and flow operating domain which is potentially
susceptible to neutronic/thermal hydraulic instability in the
absence of the stability control. In addition, the proposed
amendments require operator actions based upon a new Period Based
Detection System (PBDS). The PBDS is designed to provide alarm
indication that conditions consistent with a significant degradation
in the stability performance of the reactor has occurred and the
potential for imminent onset of neutronic/thermal hydraulic
instability may exist.
The proposed amendments will permit operation in regions of the
power and flow operating domain postulated to be susceptible to
neutron/thermal hydraulic instability (i.e., Restricted and
Monitored Regions). Operation in these regions does not increase the
probability of occurrence of initiators and precursors of previously
analyzed accidents when neutronic/thermal hydraulic instability is
not possible. The proposed amendments also permit the implementation
of the features of the E1A solution which prevent neutronic/thermal
hydraulic instability including pre-emptive reactor scram upon entry
into the region of the power and flow operating domain most
susceptible to neutronic/thermal hydraulic instability (i.e.,
Exclusion Region). Furthermore, the E1A solution requires
implementation of stability control prior to entry into a region of
the power and flow operating domain which is potentially
susceptible, in the absence of stability control, to neutronic/
thermal hydraulic instability (i.e., Restricted Region). The E1A
solution prevents neutronic/thermal hydraulic instability during
operation in regions of the power and flow operating domain
previously excluded from operation and therefore does not
significantly increase the probability of a previously analyzed
accident.
Operation in the regions of the power and flow operating domain
excluded by current Technical Specification 3/4.4.1.1 and Figure
3.4.1.1-1 can occur as a result of anticipated operational
occurrences. The severity of these transients may increase in the
absence of operator actions due to the potential occurrence of
neutronic/thermal hydraulic instability as a result of operation in
these regions. The proposed amendments will permit the
implementation of the E1A long term solution to the stability issue.
Required features of the E1A solution include adherence to a boiling
boundary limit stability control prior to selection by the operator
of APRM flow biased scram and control rod block function setpoints
which
[[Page 8794]]
allow operation in a region of the power and flow operating domain
potentially susceptible, in the absence of the stability control, to
neutronic/thermal hydraulic instability. Upon entry, as a result of
an anticipated operational occurrence, into the region most
susceptible to neutronic/thermal hydraulic instability during
operation with the boiling boundary limit stability control met, the
pre-emptive reactor scram prevents neutronic/thermal hydraulic
instability. Therefore, the consequences of an accident do not
significantly increase while operating with the stability control
met. After exiting the region requiring the stability control to be
met, the setpoints are automatically returned to the values
applicable when anticipated operational occurrences can be initiated
from conditions with the stability control not met. This automatic
actuation of the more conservative setpoints ensures that the pre-
emptive reactor scram will prevent operation as a result of an
anticipated operational occurrence in the region most susceptible to
neutronic/thermal hydraulic instability should the operator not
select the more conservative setpoints appropriate for operation
following exit from the region requiring stability control. These
required features of the E1A solution prevent operation in the
region of the power and flow operating domain most susceptible to
postulated neutronic/thermal hydraulic instability by pre-emptive
reactor scram regardless of how the region was entered. Therefore,
the proposed amendments prevent the occurrence of neutronic/thermal
hydraulic instability as a consequence of an anticipated operational
occurrence and do not significantly increase the consequences of any
previously analyzed accident.
2. The proposed amendments do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendments eliminate restrictions on power and flow
conditions and impose alternative restrictions which permit the
implementation of the E1A long term stability solution. The current
restrictions on the power and flow conditions do not prevent the
entry into regions of the power and flow operating domain most
susceptible to neutronic/thermal hydraulic instability and therefore
the possibility of neutronic/thermal hydraulic instability exists in
the absence of operator action. The required features of the E1A
solution implement a pre-emptive scram upon entry into the region
most susceptible, without operator action, to neutronic/thermal
hydraulic instability. The accessible operating domain allowed by
the proposed amendments is a subset of the power and flow operating
domain currently allowed. Current initiators and precursors of
accidents and anticipated operational occurrences can not occur with
new or different initial conditions. Therefore, the proposed
amendments do not create the possibility of a new or different kind
of accident from that previously evaluated.
Concurrent with the implementation of the proposed amendments, a
modified Flow Control Trip Reference (FCTR) card and a new Period
Based Detection System (PBDS) will be installed as required by the
E1A solution. The function of the FCTR card is to aid the operator
in the identification of entry into regions of the power and flow
operating domain potentially susceptible to neutronic/thermal
hydraulic instability and to initiate a pre-emptive scram upon entry
into the regions most susceptible to neutronic/thermal hydraulic
instability. This is accomplished by altering the values of
setpoints of the APRM flow biased scram and the control rod block
functions generated by the modified FCTR card, which are existing
functions of the current FCTR card. The modified FCTR card design
includes components which may be susceptible to electromagnetic
interference or other environmental effects. The plant specific
environmental conditions (temperature, humidity, pressure, seismic,
and electromagnetic compatibility) have been confirmed to be
enveloped by the PBDS environmental qualification values and will be
confirmed to be enveloped by the E1A FCTR card environmental
qualification values prior to installation. Therefore, the potential
for spurious scrams or common mode failures induced by environmental
effects (e.g., electromagnetic interference) is considered
negligible. The installation of the modified FCTR card will
therefore not create the possibility of a new or different kind of
accident from any accident previously evaluated. The function of the
PBDS is to provide the operator with an indication that conditions
consistent with a significant degradation in the stability
performance of the reactor has occurred and the potential for
imminent onset of neutronic/thermal hydraulic instability may exist.
This is accomplished by the installation of a new PBDS card in the
Neutron Monitoring System. The PBDS card takes inputs from
individual local power range monitors and provides displays
indicating alarm and status conditions to the operator in the
control room. These displays can not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The PBDS card design includes components which may be
susceptible to electromagnetic interference or other environmental
effects. The plant specific environmental conditions (temperature,
humidity, pressure, seismic, and electromagnetic compatibility) have
been confirmed to be enveloped by the PBDS environmental
qualification values and will be confirmed to be enveloped by the
E1A FCTR card environmental qualification values prior to
installation. Therefore, the installation of the PBDS card will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed amendments do not involve a significant
reduction in a margin of safety. The proposed amendments permit the
implementation of the E1A long term solution to the stability issue.
Under certain conditions, existing BWR designs are susceptible to
neutronic/thermal hydraulic instability. General Design Criterion
(GDC) 12 OF 10 CFR 50, Appendix A, requires thermal hydraulic
instability to be prevented by design or be readily and reliably
detected and suppressed. When the design of the reactor system does
not prevent the occurrence of neutronic/thermal hydraulic
instability, instability is an anticipated operational occurrence.
GDC 10 of 10 CFR 50, Appendix A, requires that specified acceptable
fuel design limits not be exceeded during anticipated operational
occurrences.
Analyses performed by the BWROG indicate that neutronic/thermal
hydraulic instability induced power oscillations could result in
conditions exceeding the Minimum Critical Power Ratio (MCPR) Safety
Limit (SL) prior to detection and suppression by the current design
of the Neutron Monitoring System and Reactor Protection System. To
ensure compliance with GDC 12, the BWROG developed Interim
Corrective Actions (ICAs) to enhance the capability of the operator
to readily and reliably detect and suppress neutronic/thermal
hydraulic instability. The BWROG ICAs also provided additional
guidance for monitoring local power range monitors beyond the
requirements of current Technical Specification 3/4.4.1.1 to ensure
adequate margin to the onset of neutronic/thermal hydraulic
instability. Reliance on operator actions to comply with GDC 12 was
accepted on an interim basis by the NRC pending final implementation
of a long term solution to the stability issue.
The modified design of the Reactor Protection System (APRM flow
biased scram) implemented with the E1A solution prevents neutron/
thermal hydraulic instability. The E1A solution also requires
implementation of the stability control prior to entry into a region
of the power and flow operating domain which is potentially
susceptible, in the absence of the stability control, to neutronic/
thermal hydraulic instability. As a result, the margin to the onset
of neutronic/thermal hydraulic instability provided by the existing
Technical Specification requirements and BWROG ICAs recommendations
is not significantly reduced by the implementation of the E1A
solution. The E1A solution assures compliance with GDC 12 by the
prevention of neutronic/thermal hydraulic instability and therefore
precludes neutronic/thermal hydraulic instability from becoming a
credible consequence of an anticipated operational occurrence. The
consequences of anticipated operational occurrences and the margin
to the MCPR SL will not change upon the implementation of the E1A
solution. Therefore, the proposed amendments 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 location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
[[Page 8795]]
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart (Acting).
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois.
Date of amendment request: January 20, 1997.
Description of amendment request: The proposed amendments would
relocate the surveillance requirements for selected instrumentation
from the Technical Specifications to licensee controlled documents
because the instrumentation provides indication or an alarm only. The
affected surveillance requirements are: 4.1.3.5.b, ``Control Rod Scram
Accumulators''; 4.5.1.d.2.c, `` Emergency Core Cooling Systems--
Operating''; 4.5.3.1.b, ``ECCS--Suppression Chamber''; and 4.6.2.1.c,
``Containment Systems--Suppression Chamber''. In addition, the proposed
amendments would replace TS SR 4.4.3.2.1, ``Reactor Coolant System
Leakage'' and SR 4.5.1.d.1, ``ECCS--Operating'' with surveillances more
appropriate to the associated LCOs and action statements. Also, the
proposed amendments add an action statement to TS 3.5.1, ``ECCS--
Operating'' regarding pressure of the ADS accumulator backup compressed
gas system bottle, and delete action statements 3.5.3.c, 3.5.3.d,
3.6.2.1.c and 3.6.2.1.d regarding suppression chamber water level
instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because:
The proposed change relocates instrumentation requirements,
which provide no post-accident function from the Technical
Specifications to the Bases, UFSAR, procedures, or other plant
controlled documents. These requirements are part of routine
operational monitoring and are not considered in the safety
analysis. The Bases, UFSAR, procedures, and other plant controlled
documents containing the relocated information will be maintained in
accordance with 10 CFR 50.59. In addition to 10 CFR 50.59
provisions, the Technical Specification Bases are subject to the
change control provisions in the Administrative Controls Chapter of
the Technical Specifications. The UFSAR is subject to the change
control provisions of 10 CFR 50.71(e), and plant procedures and
other plant controlled documents are subject to controls imposed by
plant administrative procedures, which endorse applicable
regulations and standards. Since any changes to the Bases, UFSAR,
procedures, or other plant controlled documents will be evaluated
per the requirements of 10 CFR 50.59, no significant increase in the
probability or consequences of an accident previously evaluated will
be allowed. Therefore, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The Reactor Coolant Operational Leakage limits monitoring
surveillance 4.4.3.2.1 has been modified to eliminate procedural
details of what instrumentation/leakage detection systems to use in
verifying limits. The proposed surveillance requires verification
that the reactor coolant system leakage is within limits at the same
frequency as the current surveillance requirement. The reactor
coolant leakage detection systems operability requirements are
controlled by Technical Specification 3/4.4.3.1. Since any changes
to procedures describing the method of monitoring leakage will be
evaluated per the requirements of 10 CFR 50.59, no significant
increase in the probability or consequences of an accident
previously evaluated will be allowed. Therefore, this change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The monitoring action and the surveillance requirements added
for the Automatic Depressurization System (ADS) pneumatic supply
help assure the continued operability of ADS for the mitigation of
accidents involving high reactor vessel pressure and the loss of the
high pressure core spray system. The surveillance frequency is
reasonable for the ADS supply header pressure due to the redundancy
of the instrument nitrogen system, [and] several alarms [that warn]
of system trouble. The ADS accumulator backup compressed gas system
bottle pressure monitoring surveillance frequency and the proposed
action on low bottle pressure is reasonable due to the [presence of
the] ADS accumulator check valves and the [availability of the]
normal ADS supply header. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because:
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. The
proposed change will not impose or eliminate any requirements, and
adequate control of the requirements will be maintained. Thus, these
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Involve a significant reduction in the margin of safety
because:
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumption. In addition, the
requirements to be transposed from the Technical Specifications to
procedures, or other plant controlled documents are the same as the
existing Technical Specifications. Since any future changes to these
requirements in the Bases, UFSAR, procedures, or other plant
controlled documents will be evaluated per the requirements of 10
CFR 50.59, no significant reduction in a margin of safety will be
allowed.
Based on 10 CFR 50.92, the existing requirement for NRC review
and approval of revisions to these requirements proposed for
relocation does not have a specific margin of safety upon which to
evaluate. However, since the proposed change is consistent with the
BWR Standard Technical Specifications, NUREG-1434, approved by the
NRC Staff, revising the Technical Specifications to reflect the
approved level of instrumentation requirements ensures no
significant reduction in the margin of safety.
The Reactor Coolant Operational Leakage limits monitoring
surveillance 4.4.3.2.1 has been modified to eliminate procedural
details of what instrumentation/leakage detection systems to use in
verifying limits. The proposed surveillance requires verification
that the reactor coolant system leakage is within limits at the same
frequency as the current surveillance requirement. The reactor
coolant leakage detection systems operability requirements are
controlled by Technical Specification 3/4.4.3.1. Because there are
no changes to either the reactor coolant leakage detection systems
and the reactor coolant leakage continues to be maintained within
the specified limits, at the required frequency, there is no
reduction in the margin of safety.
The monitoring action and the surveillance requirements added
for the Automatic Depressurization System (ADS) pneumatic supply
help assure the continued operability of ADS for the mitigation of
accidents involving high reactor vessel pressure and the loss of the
high pressure core spray system. This helps assure ADS is maintained
in a ready status. The previous TS SRs only tested the
instrumentation, and did not verify the parameter remained within
limits. Therefore, the margin of safety is not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library,
Illinois Valley Community College, Oglesby, Illinois 61348.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan.
[[Page 8796]]
Date of amendment request: January 10, 1996.
Description of amendment request: The proposed amendment would
revise test requirements for the containment emergency escape airlock.
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 evaluation supports the finding that operation of
the facility in accordance with the proposed change to the Technical
Specifications would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not alter any plant operating
conditions, operating practices, equipment design, equipment
settings, or equipment capabilities. Therefore, operation of the
facility in accordance with the proposed change will not involve an
increase in the probability of an accident. This determination is
made because the full pressure test and the seal contact check
provides reasonable assurance that the Emergency Escape Airlock
doors will act as designed to maintain containment integrity.
Procedures are established to test seal integrity with full pressure
airlock test and to verify the seal contact following the test.
Acceptance criteria are established for each evolution. Failure to
meet the acceptance criteria would result in corrective action to
restore the Emergency Escape Airlock to the intended condition.
The proposed change defines the pressure tests required for the
Emergency Escape Airlock and specifies the method used to restore
the airlock door seals after full pressure testing. Due to the
design of the airlock, the doors must be opened after testing. This
change recognizes the practice of verifying the final integrity of
the airlock by verifying door seal contact. Since the pressure test
does not load the door seals in the same direction as a design basis
accident, this seal contact check provides better assurance that the
door is sealed than alternative pressure tests. The Emergency Escape
Airlock continues to be capable of performing its design function
and the consequences of those accidents previously evaluated will
not increase.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change does not alter any plant operating
conditions, operating practices, equipment design, equipment
settings, or equipment capabilities. Therefore, operation of the
facility in accordance with the proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change requires testing of the Emergency Escape
Airlock at full pressure (greater than or equal to Pa) rather
than a reduced pressure between-the-seals test. This reduced
pressure test is allowed by the existing Technical Specifications
when the door is opened during periods when containment integrity is
required. The door seal contact check and restoration will provide
assurance that the Emergency Escape Airlock is capable of performing
its design function after the doors are opened during recovery from
full pressure testing. Implementation of these test requirements and
meeting the acceptance criteria will ensure that containment
integrity with respect to the Emergency Escape Airlock will be
maintained. Therefore, there will be no reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: John N. Hannon.
Duke Power Company, Docket Nos. 50-269, 270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina.
Date of amendment request: February 5, 1997 (TSC 96-11)
Description of amendment request: The proposed changes would
reflect replacement of the existing nuclear instrumentation with an
enhanced wide range nuclear instrumentation system that provides more
channels and continuous coverage from the source to above the power
range. As a result: (1) The various references to Intermediate Range of
nuclear instrumentation would be eliminated and replaced with reference
to Wide Range instrumentation; (2) the minimum number of operable
Source and Wide Range Nuclear Instrumentation channels that are
available and that are required to be operable in Table 3.5.1-1 would
be increased; (3) the minimum power level specified in Note (c) of
Table 3.5.1-1 would be changed from 10-10 amps on the intermediate
range instrument channels to 4 x 10-4% rated power on the wide
range instrument channels; and (4) entries that specify the Wide Range
Nuclear Instrumentation, the number of Required Operable Channels,
reference to a new Action Statement, and Applicability would be added
to Table 3.5.6-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed amendment to the Oconee Technical
Specifications is associated with the implementation of an enhanced
nuclear instrumentation system. The new Gamma Metrics system
provides twice the number of channels of neutron detectors for use
during both normal plant operations and post-accident monitoring.
The proposed change will make Oconee's Technical Specifications
consistent with a nuclear instrumentation system that meets the
reliability and redundancy requirements of Regulatory Guide 1.97.
Additionally, the new Technical Specifications will be more
conservative in terms of stating the minimum number of operable
channels required, since there are now a greater number of redundant
channels available. Assuring that the nuclear instrumentation at
Oconee is more reliable and more redundant, does not affect the
probability of an occurrence of an accident, since this system is a
monitoring system and not an accident initiator. However, these
characteristics (increased reliability and redundancy) could provide
additional capability to deal with the consequences of post-accident
situations.
(2) Will the change create the possibility of a new or different
kind of accident from any [kind of accident] previously evaluated?
No. The proposed amendment to Oconee Technical Specifications
involves the implementation of an enhanced nuclear instrumentation
system. By implementing a nuclear instrumentation system that meets
the provisions of Regulatory Guide 1.97, Oconee's ability for
neutron monitoring is enhanced during normal operations and post-
accident recovery. The Source Range nuclear instrumentation system
is utilized for monitoring purposes only, while the Wide Range
provides a control rod withdrawal interlock based on high startup
rate. The new Gamma Metrics detectors have been shown to be more
reliable, accurate, and redundant than Oconee's original detectors.
Therefore, changing the Oconee Technical Specifications to be
consistent with the current nuclear instrumentation arrangement, as
proposed in this amendment request, has no effect on the possibility
of any type of accident: new, different, or previously evaluated.
(3) Will the change involve a significant reduction in a margin
of safety?
No. Margin of safety is associated with confidence in the
ability to maintain the fission product barriers (i.e., fuel and
fuel cladding, Reactor Coolant System pressure boundary, and
containment structure) to limit the level of radiation dose to the
public. The proposed Technical Specifications amendment will
establish operability requirements for an enhanced nuclear
instrumentation system at Oconee. By
[[Page 8797]]
implementing a more reliable and redundant nuclear instrumentation
system, Oconee's post-accident monitoring capability is enhanced.
Therefore, the ability to protect the public from radiation dose is
further assured, and no reduction in any existing margin of safety
will occur.
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 location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036.
NRC Project Director: Herbert N. Berkow.
Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi.
Date of amendment request: October 22, 1996.
Description of amendment request: The proposed amendment would
revise Figure 3.4.11-1, ``Minimum Reactor Vessel Metal Temperature vs.
Reactor Vessel Pressure,'' in Limiting Condition for Operation 3.4.11,
``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,''
of the Technical Specifications (TSs). The existing curve is valid only
up to 10 Effective Full Power Years (EFPYs) and would be revised to be
valid up to 32 EFPYs.
The proposed curves, pages 1 through 5 of Figure 3.4.11-1, have
been drawn for five different EFPY periods: 16, 20, 24, 28 and 32.
There are two sets of curves attached to the licensee's application.
The first set of curves (Attachment 3) would replace the existing curve
in TS Figure 3.4.11-1. The second set of curves (Attachment 4) are
duplicates of the Attachment 3 curves except that these curves also
contain detailed information used in development of the curves and
would be included in the next update of the Updated Final Safety
Analysis Report (UFSAR) for information.
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 its application for the proposed amendment, which is
presented below:
(A) The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
Regulatory Guide 1.99, Revision 2 is currently used to prepare
the pressure-temperature limit curves and is inherently conservative
for Boiling Water Reactors (BWRs). [Grand Gulf Unit 1 is a BWR.] The
proposed Technical Specification Figure 3.4.11-1 was prepared in
accordance with the requirements of 10CFR50 [10 CFR Part 50],
Appendix G [(Fracture Toughness Requirements)], and using NRC
approved methodology outlined in NRC Regulatory Guide 1.99, Revision
2, ``Radiation Embrittlement of Reactor Vessel Materials.''
Operation of the plant within the limitations of the proposed figure
will ensure that the Requirements of 10CFR50 [10 CFR Part 50],
Appendix G are met up to and including 32 Effective Full Power Years
(EFPY) of operation. The proposed changes assure that the existing
safety limits are not exceeded due to changing Reactor Vessel
conditions by continued incorporation of the effect of neutron
radiation embrittlement of vessel materials into the proposed
curves.
The curves have also been editorially enhanced by removal of
phrases used for validation of the curves. Having the phrases on the
TS (Technical Specification) curves distracts from the intended
purpose which is to maintain operation of the reactor to the right
of the curves. Operators, in performance of their job function, do
not need this information to comply with TS Limiting Condition for
Operation (LCO) 3.4.11. This change also revises the curve labeling
consistent with the terminology used in Table 1 of 10CFR50 [10 CFR
Part 50], Appendix G. These enhancements and revisions have no
impact on the operation of the plant since they are editorial in
nature and do not change the technical content of the curves.
Therefore, the proposed change does not significantly increase
the probability or consequences of an accident previously evaluated.
(B) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The pressure-temperature curves are controlled by the Technical
Specifications and are determined using the conservative methodology
in NRC Regulatory Guide 1.99, Revision 2, ``Radiation Embrittlement
of Reactor Vessel Materials.'' The proposed pressure-temperature
limit curves are inherently conservative, therefore, the possibility
of failure of the reactor vessel is not increased. The proposed
curves establish new periods of applicability (16, 20, 24, 28, and
32 EFPY) for the current pressure-temperature limitations based on
NRC methodology in Regulatory Guide 1.99 and actual fluence
measurements. These limitations are appropriate up to and including
32 EFPY exposure and operation of the plant within the figure's
limitations will ensure that the requirements of 10CFR50 [10 CFR
Part 50], Appendix G are met for that time frame. No physical plant
modifications or new operating configurations result from these
changes. These changes do not adversely affect the design or
operation of any system or component important to safety, rather
they establish limits to assure that operations remain within
acceptable safety boundaries.
The curves have also been editorially enhanced by removal of
phrases used for validation of the curves. Having the phrases on the
TS curves distracts from the intended purpose which is to maintain
operation of the reactor to the right of the curves. Operators, in
performance of their job function, do not need this information to
comply with TS Limiting Condition for Operation (LCO) 3.4.11. This
change also revises the curve labeling consistent with the
terminology used in Table 1 of 10CFR50 [10 CFR Part 50], Appendix G.
These enhancements and revisions have no impact on the operation of
the plant since they are editorial in nature and do not change the
technical content of the curves.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(C) The proposed change does not involve a significant reduction
in a margin of safety.
The proposed curves were developed using the methodology of
Regulatory Guide 1.99, Revision 2, ``Radiation Embrittlement of
Reactor Vessel Materials.'' This methodology includes an allowance
for margin that is to be included in the upper-bound values of the
adjusted reference temperature (ART). The proposed changes maintain
the existing margins of safety by modifying the operating limits
based on the most limiting of the actual reference temperature
shifts. These new limits consider the most limiting pressure vessel
material. The revised analysis demonstrates that the existing
Technical Specification [TS] pressure-temperature limit curves are
applicable for periods of 16, 20, 24, 28, and 32 EFPY. Using the
methodology in NRC Regulatory Guide 1.99 Revision 2 and fluence
based on actual exposure provides for additional conservatism, and
therefore [,] further assures the existence of current margins of
safety. The proposed pressure-temperature limit curves are
inherently conservative and provide sufficient margin to ensure the
integrity of the reactor vessel.
The curves have also been editorially enhanced by removal of
phrases used for validation of the curves. Having the phrases on the
TS curves distracts from the intended purpose which is to maintain
operation of the reactor to the right of the curves. Operators, in
performance of their job function, do not need this information to
comply with TS Limiting Condition for Operation (LCO) 3.4.11. This
change also revises the curve labeling consistent with the
terminology used in Table 1 of 10CFR50 [10 CFR Part 50], Appendix G.
These enhancements and revisions have no impact on the operation of
the plant since they are editorial in nature and do not change the
technical content of the curves.
Continuing commitment to the methodology contained in NRC
Regulatory Guide 1.99, Rev. 2, will ensure that the most limiting
plate or beltline weld material will be utilized in the
determination of the pressure-temperature limits for any future
curve changes.
[[Page 8798]]
Therefore, the proposed change does not result in a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana.
Date of amendment request: January 10, 1997.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) for reactor pressure vessel
pressure and temperature (P-T) limits to replace the curves for 2
effective full power years (EFPY) with curves for 12 EFPY. The P-T
curves are used for heatup, cooldown, and inservice leak and
hydrostatic testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Pressure-temperature (P-T) limits (RBS Technical Specifications
Figure 3.4.11-1) are imposed on the reactor coolant system to ensure
that adequate safety margins against nonductile or rapidly
propagating failure exist during normal operation, anticipated
operational occurrences, and system hydrostatic tests. The P-T
limits are related to the nil-ductility reference temperature,
RTNDT, as described in ASME Section III, Appendix G. Changes in
the fracture toughness properties of [Reactor Pressure Vessel] RPV
beltline materials, resulting from the neutron irradiation and the
thermal environment, are monitored by a surveillance program in
compliance with the requirements of 10 CFR [Part] 50, Appendix H.
The effect of neutron fluence on the nil-ductility reference
temperature of pressure vessel steel is predicted by methods given
in Regulatory Guide [RG] 1.99, Rev. 2.
The revised P-T limits of this amendment request were
established based on adjusted reference temperatures developed in
accordance with the procedures prescribed in Reg. Guide [RG] 1.99,
Rev. 2, Regulatory Position C.1. Calculation of adjusted reference
temperature by these procedures includes a margin term to ensure
conservative, upper-bound values are used for the calculation of the
P-T limits. Stress intensity factors used to compute the pressures
were calculated in accordance with, and include the required safety
factors given in ASME Section III, Appendix G. The limits
established by the lower portion of the P-T curves, which cover the
discontinuity (non-beltline) regions of the vessel (e.g., flanges,
nozzles, etc.), were retained throughout this current analysis. The
limits established by the lower portion of these curves do not
change as they are not affected significantly by the neutron
fluence.
This change is not related to any accidents previously
evaluated. The proposed change will provide for approved P-T limit
curves which are valid through 12 EFPY. This change will not affect
any Safety Limits, Power Distribution Limits, or Limiting Conditions
for Operation. The proposed change will not affect reactor pressure
vessel [RPV] performance as no physical changes are involved and RBS
vessel P-T limits will remain conservative in accordance with Reg.
Guide [RG] 1.99, Rev. 2 and ASME Section III, Appendix G
requirements. The proposed change will not cause the reactor
pressure vessel [RPV] or interfacing systems to be operated outside
of their design or testing limits. Also, the proposed change will
not alter any assumptions previously made in evaluating the
radiological consequences of accidents. The proposed change ensures
that adequate margins against brittle fracture of the vessel are
maintained through 12 EFPY of reactor operations. Therefore, the
probability or consequences of accidents previously evaluated will
not be increased by the proposed change.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change is a revision of Technical Specification
Figure 3.4.11-1 to show P-T limit curves valid through 12 EFPY. The
revised P-T limits have been established in accordance with
applicable NRC regulations and the ASME Code. This proposed change
does not involve a modification of the design of plant structures,
systems, or components. The proposed change will not impact the
manner in which the plant is operated as plant operating and testing
procedures will not be affected by the change. The proposed change
will not degrade the reliability of structures, systems, or
components important to safety (ITS) as equipment protection
features will not be deleted or modified, equipment redundancy or
independence will not be reduced, supporting system performance will
not be downgraded, the frequency of operation of ITS equipment will
not be imposed. No new accident types or failure modes will be
introduced as a result of the proposed change. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from that previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
As stated in the River Bend SER, ``Appendices G and H of 10 CFR
50 describe the conditions that require pressure-temperature [P-T]
limits and provide the general bases for these limits. These
appendices specifically require that pressure-temperature [P-T]
limits must provide safety margins at least as great as those
recommended in the ASME Code, Section III, Appendix G. * * * Until
the results from the reactor vessel surveillance program become
available, the staff will use RG 1.99, Revision 1 [now Revision 2]
to predict the amount of neutron irradiation damage. * * * The use
of operating limits based on these criteria--as defined by
applicable regulations, codes, and standards--will provide
reasonable assurance that nonductile or rapidly propagating failure
will not occur, and will constitute an acceptable basis for
satisfying the applicable requirements of GDC 31.''
Bases for RBS Technical Specification 3.4.11 states: ``The P/T
[P-T] limits are not derived from Design Basis Accident (DBA)
analyses. They are prescribed during normal operation to avoid
encountering pressure, temperature, and temperature rate of change
conditions that might cause undetected flaws to propagate and cause
nonductile failure of the RCPB, a condition that is unanalyzed. * *
* Since the P/T [P-T] limits are not derived from any DBA, there are
no acceptance limits related to the P/T [P-T] limits. Rather, the P/
T [P-T] limits are acceptance limits themselves since they preclude
operation in an unanalyzed condition.''
This amendment request proposes P-T limit curves which will be
valid through 12 EFPY. The proposed P-T limits were established
based on adjusted reference temperatures for vessel beltline
material calculated in accordance with Regulatory Position 1 of Reg.
Guide [RG] 1.99, Rev. 2 and pressures calculated in accordance with
ASME Section III, Appendix G requirements. Required margins and
safety factors were included to ensure that conservative, upper-
bound values were used in calculation of the P-T limits. The
proposed change will not affect any Safety Limits, Power
Distribution Limits, or Limiting Conditions for Operation. The
proposed change does not represent a change in initial conditions,
or in a system response time, or in any other parameter affecting
the course of an accident analysis supporting the Bases of any
Technical Specification. The proposed P-T limits provide adequate
safety margins against brittle failure of the reactor vessel through
12 EFPY of power operations. For these reasons, the proposed changes
do not involve a reduction in any margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 8799]]
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Project Director: William D. Beckner.
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana.
Date of amendment request: January 20, 1997.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to allow the use of flow
control spectral shift strategies to increase cycle energy; an
estimated additional 30 days at full power. The request is based on a
General Electric (GE) Maximum Extended Load Line Limit (MELLL) analysis
for the River Bend Station.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated.
Abnormal operational transients or accidents analyzed in the SAR
have been examined for any impact caused by MELLL operation. The
limiting abnormal operation transients, including the Generator Load
Rejection with No Bypass (LRNBP) event and the Feedwater Controller
Failure (FWCF) maximum demand event, have been evaluated in detail.
The LOCA [Loss-of-Coolant Accident], Fuel Loading Error (FLE), rod
drop accident, rod withdrawal error, and the Anticipated Transient
Without Scram (ATWS) analyses have also been evaluated for the
effects of MELLL operation. The flow and power dependent [Minimum
Critical Power Ratio] MCPR curves for off-rated and rated conditions
and the [Maximum Average Planar Linear Heat Generation Rate] MAPLHGR
criteria establish limits on power operation. These limits ensure
that the core is operated within the assumptions and initial
conditions of the transient or accident analyses. Operation within
these limits will ensure that the consequences of a transient or
accident remain within the acceptable limits of the analyses.
The [Average Power Range Monitor] APRM scram in the Technical
Specifications [TSs] and affected rod block setpoints are revised to
ensure that operation remains within the analyzed MELLL region. This
restriction ensures the consequences of abnormal operation and
accidents are acceptable. The probability of an accident is not
affected by the proposed Technical Specification [TS] changes since
no systems or equipment which could initiate an accident are
affected. Therefore, the proposed changes do not significantly
increase the probability or consequences of any previously evaluated
accident.
2. The request does not create the possibility of occurrence of
a new or different kind of accident from any accident previously
evaluated.
Operation in the MELLL domain expands the current power/flow
along the 121% rod line to 100% power at 75% rated core flow and
improves flexibility and capacity factor. Abnormal operation
transients or accidents have been evaluated and the most limiting
cases have been analyzed for applicability for operation in the
MELLL region. The proposed Technical Specification [TS] changes
prohibit power operation outside the MELLL region and do not
constitute or require any system or equipment changes that might
create an accident of a different type then previously evaluated.
The MAPLHGR, the power and flow dependent MCPR and [Liner Heat
Generation Rate] LHGR and the revised Technical Specifications [TSs]
will continue to assure that plant operation is consistent with the
assumptions, initial conditions and assumed power distribution and
therefore will not create a new type of accident. The proposed
Technical Specification [TS] changes do not introduce any new modes
of plant operation nor involve new system interactions. Therefore,
the proposed changes do not create the possibility of a new or
different kind of accident from any previous analyzed.
3. The request does not involve a significant reduction in a
margin of safety.
The proposed Technical Specifications [TSs] prohibit power
operation outside the allowable MELLL region. The transients and
accidents described in the SAR are evaluated for operation in the
MELLL region. NEDC-32611, ``MELLL Analysis for River Bend Station
Reload 6 Cycle 7,'' shows that the OLMCPR for operation in the MELLL
region is bounded by the OLMCPR established for current conditions
(100% power/107% flow). The thermal limits MCPR and LHGR curves and
the MAPLHGR limits establish limits on power operation and thereby
ensure that the core is operated within the assumptions and initial
conditions of the transient and accident analyses.
As demonstrated in the analysis provided in Attachment 4, [the
proposed amendment request] operation within these limits, using the
MCPR limits, LHGH limits and MAPLHGR criteria, will ensure that the
margin of safety will be maintained to the same level described in
the Technical Specifications Bases and the SAR and the consequences
of the postulated transient or accidents are not increased. The MCPR
safety limit, mechanical performance limits and overpressure limit
are not exceeded during any transient or postulated accident.
Therefore, the proposed Technical Specifications [TSs] to allow
operation in the MELLL region do not involve a significant reduction
in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005.
NRC Project Director: William D. Beckner.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine.
Date of amendment request: February 7, 1997.
Description of amendment request: The proposed amendment would
modify Technical Specification 3.12 to require both 115 kV incoming
lines to be operable when the reactor is critical; allow continued
operations for up to 72 hours with one 115 kV incoming line inoperable;
allow continued operations for up to 24 hours with both 115 kV incoming
lines inoperable; apply the increased operability requirements
described above to another affected remedial action; incorporate minor
editorial changes to uniformly apply the usage of the term
``operable;'' and change the basis section to be consistent with the
proposed changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to Specification 3.12.B do not involve a
physical change to the plant or the maintenance of the plant. The
proposed changes increase the operating requirements associated with
the operability of the 115 kV incoming lines beyond that currently
required by Technical Specifications. For those accidents previously
evaluated, the more restrictive operability requirements associated
with maintaining both 115 kV incoming lines operable and the more
restrictive remedial action times result in increased assurance that
station service power will be available when required. This
increased availability will be achieved because elective maintenance
on the offsite power system will be significantly restricted and the
restoration of inoperable 115 kV incoming lines will be treated with
greater urgency. The increased
[[Page 8800]]
assurance of availability will result in a decrease in the
probability or consequences of these postulated accidents.
However, the more restrictive remedial action times decrease the
restoration period and consequently increase the possibility that
successful restoration may not be achieved, given an outage of the
115 kV power system. A unit shutdown without offsite power would
then be commenced. This evolution would involve a unit shutdown
without the availability of equipment such as the reactor coolant
pumps, condensate pumps and main feedwater pumps. Although none of
these components are credited as available for the mitigation of the
consequences of accidents previously evaluated, the probability of
the occurrence of certain accidents is increased without them.
Although the combination of these considerations could involve
an increase in the probability of accidents previously evaluated,
the increase would not be significant due to the low probability of
independent failures or common cause failures of both of the 115 kV
incoming lines. There is no increase in the consequences of any
accident previously evaluated as a result of these proposed
Technical Specification changes. The proposed Technical
Specification changes are consistent with the Standard Technical
Specifications approved by the NRC. The proposed changes, therefore,
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed Technical Specification change does not involve a
change to the physical plant or to the physical configuration of the
offsite power system. The effect of the proposed change will be to
increase the availability of the offsite power system when required.
In addition, the proposed change will increase the possibility of a
unit shutdown without offsite power operable. However, the accidents
previously evaluated assume a simultaneous loss of offsite power,
design basis accident and worst case single failure as part of the
design basis. The proposed changes do not result in the creation of
a unique operating condition or a configuration that has not been
previously evaluated. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
This proposed change modifies Technical Specification 3.12 to be
consistent with the Standard Technical Specifications. The proposed
Technical Specification change maintains the current margin of
safety which is based upon supplying power to engineered safeguards.
Adequate sources of power remain available for the operation of the
engineered safeguards equipment. Therefore, the proposed change
would not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Wiscasset Public Library, High
Street, P.O. Box 367, Wiscasset, ME 04578.
Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic
Power Company, 329 Bath Road, Brunswick, ME 04011.
NRC Project Director: Patrick D. Milano, Acting.
Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-
336, and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and
3, New London, Connecticut.
Date of amendment request: February 3, 1997.
Description of amendment request: The licensee has proposed to
revise Section 6, ``Administrative Controls,'' of the Millstone Unit
Nos. 1, 2, and 3 Technical Specifications to reflect organizational
changes that have been implemented in the Nuclear Division.
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 a [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No design basis accidents are affected by these proposed
changes. The proposed changes are administrative in nature and are
being proposed to reflect the organizational changes which become
effective on February 3, 1997. The unit level responsibilities of
the Executive Vice President--Nuclear are assigned to the Officers
for the individual Millstone units. The site level responsibilities
of the Executive Vice President--Nuclear are shared by the Senior
Vice President and CNO [Chief Nuclear Officer]--Millstone and the
President and Chief Executive Officer. The changes to the SORC [Site
Operations Review Committee] and the three unit[s'] PORC [Plant
Operations Review Committee] reflect changes in job function or job
position titles only.
No safety systems are adversely affected by the proposed
changes, and no failure modes are associated with the changes.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Because there are no changes in the way plants are operated due
to this administrative change, the potential for an unanalyzed
accident is not created. There is no impact on plant response, and
no new failure modes are introduced. These proposed administrative
and editorial changes have no impact on safety limits or design
basis accidents, and they have no potential to create a new or
unanalyzed event. The changes to the SORC and the three unit[s']
PORC reflect changes in job function or job position titles only.
3. Involve a significant reduction in a margin of safety.
The changes do not directly affect any protective boundaries nor
do they impact the safety limits for the protective boundaries.
These proposed changes are administrative and editorial in nature.
Therefore, there is no reduction in the margin of safety. These
changes do not reduce the margin of safety provided by the PORC and
the SORC review and approval of changes to the operations of the
Millstone Unit Nos. 1, 2, and 3.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut.
Date of amendment request: February 5, 1996.
Description of amendment request: The amendment would delete a
clause from Technical Specification 4.0.5.a. Specifically, this change
would delete the clause ``(g), except where specific written relief has
been granted by the Commission pursuant to 10 CFR Part 50, Section
50.55a(g)(6)(i).'' The amendment would also make the appropriate
changes to the Bases section. In addition, NNECO made changes to Bases
Section 3/4.7.7 and 3/4.7.8 to add design basis information and provide
clarification of system design and operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
[[Page 8801]]
consideration, which is presented below:
Pursuant to 10 CFR 50.92, NNECO has reviewed the proposed
changes to Technical Specification 4.0.5a and Bases Section 3/4.4.10
and has concluded that the changes do not involve a significant
hazards consideration (SHC). The basis for this conclusion is that
the three criteria of 10 CFR 50.92(c) are not compromised. The
proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes would remove the wording ``* * * (g),
except where specific written relief has been granted by the
Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).''
The Inservice Inspection and Testing Programs are described in the
technical specifications pursuant to 10 CFR 50.55a. In addition, the
proposed changes, in accordance with NUREG-1431 and NUREG-1482,
would provide relief to the ASME Code requirement in the interim
between the time of submittal of a relief request until the NRC has
issued a safety evaluation and granted the relief. The changes being
proposed are administrative in nature and do not affect assumptions
contained in plant safety analyses, the physical design and/or
operation of the plant, nor do they affect any technical
specification that preserves safety analysis assumptions. Any relief
from the approved ASME Section XI Code requirements will require a
10 CFR 50.59 evaluation to ensure no technical specification changes
or unreviewed safety questions exist. Therefore, operation of the
facility in accordance with the proposed changes would not affect
the probability or consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed changes would remove the wording ``* * * (g),
except where specific written relief has been granted by the
Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).''
The Inservice Inspection and Testing Programs are described in the
technical specifications pursuant to 10 CFR 50.55a. In addition, the
proposed changes, in accordance with NUREG-1431 and NUREG-1482,
would provide relief to the ASME Code requirement in the interim
between the time of submittal of a relief request until the NRC has
issued a safety evaluation and granted relief. The changes being
proposed are administrative in nature and will not change the
physical plant or the modes of operation defined in the facility
license. The changes do not involve the addition or modification of
equipment nor do they alter the design or operation of plant
systems. Any relief from the approved ASME Section XI Code
requirements will require a 10 CFR 50.59 evaluation to ensure no
technical specification changes or unreviewed safety questions
exist. Therefore, operation of the facility in accordance with the
proposed changes would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed changes would remove the wording ``* * * (g),
except where specific written relief has been granted by the
Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).''
The Inservice Inspection and Testing Programs are described in the
technical specifications pursuant to 10 CFR 50.55a. In addition, the
proposed changes, in accordance with NUREG-1431 and NUREG-1482,
would provide relief to the ASME Code requirement in the interim
between the time of submittal of a relief request until the NRC has
issued a safety evaluation and granted relief. The changes being
proposed are administrative in nature and will not alter the bases
for assurance that safety-related activities are performed correctly
or the basis for any technical specification that is related to the
establishment or maintenance of a safety margin. Any relief from the
approved ASME Section XI Code requirements will require a 10 CFR
50.59 evaluation to ensure no technical specification changes or
unreviewed safety questions exist. Therefore, operation of the
facility in accordance with the proposed changes would not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
United States Department of Commerce, National Institute of
Standards and Technology, Docket No. 50-184, NIST (formerly known as
National Bureau of Standards) Test Reactor or NBSR.
Date of amendment request: January 17, 1997.
Description of amendment request: The National Institute of
Standards and Technology (NIST) is planning to change the name of the
Reactor Radiation Division to the NIST Center for Neutron Research to
be headed by a Director. The requested amendment involves a name change
only. All functions, responsibilities, and personnel remain the same.
The Technical Specification references to the ``Chief, Reactor
Radiation Division'' will be changed to Director, NIST Center for
Neutron Research in Sections 7.1, 7.2, and 7.3. The Organization Chart
in Figure 7.1 will also reflect this change. The Technical
Specification references to the ``Reactor Radiation Division'' will be
changed to ``NIST Center for Neutron Research'' in Section 7.2.
Basis for proposed no significant hazards consideration
determination: The Commission has provided standards for determining
whether a significant hazards consideration exists (10 CFR 50.92(c)). A
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 change being proposed is a change in the title of the
organization and the title of the head of the organization that directs
the operation of the reactor. As noted previously, all functions,
responsibilities and personnel remain the same. The staff agrees with
the licensee's no significant hazards consideration and finds that the
mere title changes render a negative response to the three criteria
outlined in 10 CFR 50.92(c).
Local Public Document Room location: N/A.
Attorney for licensee: N/A
NRC Project Director: Seymour H. Weiss.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont.
Date of amendment request: December 10, 1996.
Description of amendment request: The proposed amendment would move
fire protection requirements from the Vermont Yankee Technical
Specifications to the Fire Protection Plan and the final safety
analysis report (FSAR), in accordance with the guidance in NRC Generic
Letters 86-10 and 88-12.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated:
[[Page 8802]]
The proposed changes are administrative in nature and are
consistent with the guidance provided in NRC Generic Letters 86-10
and 88-12. These changes do not affect the initial conditions or
precursors assumed in the FSAR safety analyses. These proposed
changes also do not decrease the effectiveness of equipment relied
upon to mitigate the previously evaluated accidents. Programmatic
controls will continue to assure that fire protection program
changes do not reduce the effectiveness of the program to achieve
and maintain safe shutdown in the event of a fire.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from an accident previously
evaluated:
The proposed changes do not modify any plant equipment, there is
no reduction in fire protection requirements, there is no change in
operating procedure and surveillance requirements and no reduction
in administrative control or equipment reliability. Therefore,
implementation of the proposed change will not affect the design
function or configuration of any component, introduce any new
operating scenarios, failure modes or accident initiators.
3. The proposed amendment will not involve a significant
reduction in a margin of safety:
The proposed amendment does not involve a reduction to the Fire
Protection Program. The fire protection requirements are simply
being relocated to other controlled documents. There are no
equipment modifications being proposed, only the location of fire
protection requirements, which is administrative in nature.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: R. K. Gad, III, Ropes and Gray, One
International Place, Boston, MA 02110-2624.
NRC Project Director: Patrick D. Milano, Acting Director.
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.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina.
Date of amendment request: January 10, 1997.
Brief description of amendment: The proposed change would revise
Technical Specification 4.8.1.1.2 to clarify pressure testing
requirements for the isolable and non-isolable portions of the diesel
fuel oil piping.
Date of publication of individual notice in Federal Register:
February 5, 1997 (62 FR 5490).
Expiration date of individual notice: March 6, 1997.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota.
Date of amendment request: November 6, 1996.
Description of amendment request: The proposed amendments would
revise the Technical Specifications governing the cooling water system.
The changes are proposed to improve plant operation based on
operational experience with the vertical motor-driven cooling water
pump. The changes are also proposed to incorporate information gathered
by the licensee during its self-assessment Service Water System
Operational Performance Inspection (SWSOPI) completed in late 1995.
Date of individual notice in the Federal Register: January 29, 1997
(62 FR 4338).
Expiration date of individual notice: February 28, 1997.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota.
Date of amendment requests: January 29, 1997.
Description of amendment requests: The proposed amendments would
change the Bases for Technical Specifications and the licensing basis
for the Operating Licenses relating to the cooling water system
emergency intake line flow capacity. The licensee determined through
testing that the emergency intake line flow capacity was less than the
design value stated in the Updated Final Safety Analysis Report. The
proposed changes reflect the use of operator actions to control cooling
water system flow following a seismic event. The proposed changes also
reclassify the intake canal for use during a seismic event, which would
be an additional source of cooling water during a seismic event.
Date of individual notice in the Federal Register: February 7, 1997
(62 FR 5857).
Expiration date of individual notice: March 10, 1997. NSHC
comments: February 24, 1997.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
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: January 31, 1997.
Brief description of amendment request: The amendment would make
changes to Technical Specification (TS) 3.4.3, ``Relief Valves,'' for
Salem Unit 1, and TS 3.4.5, ``Relief Valves,'' for Salem Unit 2, to
ensure that the automatic capability of the power operated relief
valves to relieve pressure is maintained when these valves are isolated
by closure of the block valves.
Date of publication of individual notice in Federal Register:
February 7, 1997 (62 FR 5861).
Expiration date of individual notice: March 10, 1997.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2: Hamilton County, Tennessee.
Date of application for amendments: October 18, 1996.
Description of amendments request: Amend Technical Specifications
to permanently incorporate new requirements associated with steam
generator tube inspections and repair. The requirements provide
alternate
[[Page 8803]]
steam generator tube plugging criteria at the tube support plate
intersections.
Date of publication of individual notice in the Federal Register:
February 11, 1997 (62 FR 6276).
Expiration date of individual notice: March 13, 1997.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks:
Manitowoc County, Wisconsin.
Date of amendment requests: September 19, 1996, as supplemented
November 18, 1996, and revised January 13 and January 27, 1997.
Description of amendment requests: The proposed amendments would
change Technical Specification requirements related to the low
temperature overpressure protection (LTOP) system. Specifically, the
reactor coolant system (RCS) temperature below which LTOP is required
to be enabled and the temperature below which one high pressure safety
injection pump is required to be rendered inoperable would be changed
from less than 275 degrees Fahrenheit to less than 355 degrees
Fahrenheit. Additionally, the restriction of ``less than the minimum
pressurization temperature for the inservice pressure test as defined
in Figure 15.3.1-1'' would be deleted and the specific temperature
limit of less than 355 degrees Fahrenheit would be specified. The
setpoint for the pressurizer power-operated relief valves (PORVs) would
be changed from less than or equal to 425 pounds per square inch gage
(psig) to less than or equal to 440 psig to allow for instrument
inaccuracies and increased margin allowed by the use of American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code
Case N-514. These modified requirements for LTOP ensure that RCS
materials meet the requirements of Title 10 of the Code of Federal
Regulations, Sec. 50.60, ``Acceptance Criteria for Fracture Prevention
Measures for Lightwater Nuclear Power Reactors for Normal Operation,''
(10 CFR 50.60) in accordance with 10 CFR Part 50, Appendices G and H,
and in accordance with the exemption granted on January 27, 1997, which
allows the use of ASME Code Case N-514 as an acceptable alternative.
Finally, editorial changes would be made to rename the ``Overpressure
Mitigating System'' to the ``Low Temperature Overpressure Protection
System.'' The September 19, 1996, application was previously noticed in
the Federal Register on October 1, 1996 (61 FR 51308).
Date of individual notice in the Federal Register: February 4, 1997
(62 FR 5256).
Expiration date of individual notice: March 6, 1997. NSHC comments
February 19, 1997.
Local Public Document Room location: 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, 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 26, 1996.
Brief description of amendments: The amendments adopt Option B of
10 CFR Part 50, Appendix J to require Type B and Type C containment
leakage testing to be performed on a performance-based testing
schedule.
Date of issuance: February 11, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 219 and 196.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
123).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated February 11, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts.
Date of application for amendment: April 25, 1996, as supplemented
December 23, 1996.
Brief description of amendment: The amendment will revise the
definition of Operable-Operability, revise Technical Specifications
(TSs) and associated Bases Section for TS 3.9.B.2 and 3.9.B.3,
``Auxiliary Electrical System,'' TS 3.4.B.1, ``Standby Liquid Control
System,'' TSs 3.7.b.1.a, c, and e, and 3.7.b.2.a, c, and e, ``Standby
Gas Treatment System and Control Room High Efficiency Air Filtration
System,'' and TSs. 4.5.F.1, ``Core and Containment Cooling Systems,''
and delete TS 3.7.b.1.f, ``Standby Gas Treatment System and Control
Room High Efficiency Air Filtration System.''
Date of issuance: February 10, 1997.
Effective date: February 10, 1997.
Amendment No.: 170.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 19, 1996 (61 FR
31172).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-
[[Page 8804]]
455, Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois.
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos.
1 and 2, Will County, Illinois.
Date of application for amendments: August 2, 1996.
Brief description of amendments: The amendments eliminate License
Condition 2.C.(16) from Facility Operating License NPF-37; License
Condition 2.C.(5) from Facility Operating License NPF-66; License
Condition 2.C.(6) from Facility Operating License NPF-72 and License
Condition 2.C.(5) from Facility Operating License NPF-77 that require
the licensee to conduct additional corrosion testing of sleeved steam
generator tubes.
Date of issuance: February 12, 1997.
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 85 to NPF-37, 85 to NPF-66, 77 to NPF-72, and 77 to
NPF-77.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revise the licenses.
Date of initial notice in Federal Register: September 25, 1996 (61
FR 50340).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 12, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina.
Date of application for amendments: November 26, 1996, as
supplemented December 17, 1996
Brief description of amendments: The amendments revise Technical
Specification 3.8.2.1 to allow a one-time change to replace the
existing 125-volt AT&T high specific gravity round cell battery banks
with the conventional low specific gravity cell battery banks.
Date of issuance: February 7, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 172 and 154.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 13, 1996 (61
FR 65605).
The December 17, 1996, letter did not change the scope of the
November 26, 1996, application and the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 7, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina 28223-0001.
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana.
Date of amendment request: August 29, 1996.
Brief description of amendment: The amendment revises the Technical
Requirements Manual (TRM) to change the reactor pressure vessel
surveillance capsule withdrawal schedule for the River Bend Station.
The first capsule will be withdrawn at 10.4 effective full power years
(EFPY) rather than at 6 EFPY.
Date of issuance: February 13, 1997.
Effective date: February 13, 1997.
Amendment No.: 92.
Facility Operating License No. NPF-47. The amendment revised the
Technical Requirements Manual.
Date of initial notice in Federal Register: October 23, 1996 (61 FR
55034) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 13, 1997.
No significant hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana.
Date of amendment request: June 27, 1996.
Brief description of amendment: The amendment modifies TS 3/
4.3.3.6, ``Accident Monitoring Instrumentation,'' to reflect the
Combution Engineering improved Standard Technical Specification (STS)
approved and issued as NUREG-1432. This amendment revises the TS to
include Accident Monitoring Instrumentation recommended in Regulatory
Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Plants to Assess Plant Conditions During and Following
an Accident,'' Revision 3.
Date of issuance: February 12, 1997.
Effective date: February 12, 1997, to be implemented within 90
days.
Amendment No.: 122.
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
40017).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 12, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana.
Date of amendment request: July 25, 1996, as supplemented by letter
dated January 27, 1997.
Brief description of amendment: The amendment changes the Appendix
A Technical Specifications by modifying TS 3/4.7.4, ``Ultimate Heat
Sink,'' to incorporate more restrictive fan operability requirements
and lower the maximum allowed basin temperature.
Date of issuance: February 13, 1997.
Effective date: February 13, 1997.
Amendment No.: 123.
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 19, 1996 (61
FR 58903).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 13, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.
Date of application for amendments: October 30, 1996.
Brief description of amendments: These amendments revise the St.
Lucie Technical Specifications to remove inconsistencies between the
definition of Core Alterations and the Applicability, Action and
Surveillance requirements of two specifications relating to water level
and containment
[[Page 8805]]
isolation systems during refueling operations.
Date of Issuance: February 10, 1997.
Effective Date: February 10, 1997.
Amendment Nos.: 148 and 87.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64386).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.
Date of application for amendment: October 28, 1996.
Brief description of amendment: The amendments consist of changes
to the Technical Specifications (TS) in response to your applications,
both dated October 28, 1996, regarding containment leakage tests and
removal of certain component lists from the TS.
Date of Issuance: February 10, 1997.
Effective Date: February 10, 1997.
Amendment Nos.: 149 and 88.
Facility Operating License No. NPF-16: Amendments revised the
Technical Specifications.
Date of initial notice in Federal Register: (61 FR 64386) December
4, 1996. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida.
Date of application for amendments: December 17, 1996.
Brief description of amendments: Revision to Technical
Specification (TS) 4.4.10 regarding reactor coolant pump flywheel
inspection intervals.
Date of issuance: February 11, 1997.
Effective date: February 11, 1997.
Amendment Nos.: 193 and 187.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: January 10, 1997 (62 FR
1476). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 11, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York.
Date of application for amendment: July 16, 1996.
Brief description of amendment: The amendment changes the Technical
Specifications to permit the use of 10 CFR Part 50, Appendix J, Option
B, Performance-Based Containment Leakage Rate Testing in accordance
with the implementation guidance in NRC's Regulatory Guide 1.163 dated
September 1995.
Date of issuance: February 10, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 159.
Facility Operating License No. DPR-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52965). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota.
Date of application for amendments: August 15, 1996.
Brief description of amendments: The amendments revise the
containment cooling systems limiting conditions for operation technical
specifications to bring them into conformance with recently completed
system analyses by no longer permitting both containment spray pumps to
be inoperable at the same time.
Date of issuance: February 10, 1997.
Effective date: February 10, 1997, with full implementation within
30 days.
Amendment Nos.: 125 and 117.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64388).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
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: November 25, 1996.
Brief description of amendments: These amendments revise the
wording in TS Section 4.8.1.1.2.e.2 and the associated TS Bases Section
3/4.8, to remove the specific reference to the Residual Heat Removal
(RHR) pump motor and its corresponding kW rating value, and replace it
with wording consistent with that specified in the Improved TS (i.e.,
NUREG-1433, Revision 1, ``Standard Technical Specifications General
Electric Plants,'' dated April 1995).
Date of issuance: February 4, 1997.
Effective date: Both units, as of date of issuance, to be
implemented within 30 days.
Amendment Nos.: 121 and 85.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66716).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 4, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 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: September 27, 1996.
Brief description of amendments: These amendments increase the
reactor enclosure secondary containment maximum inleakage rate, and
also impact secondary containment drawdown time and system flow rate
assumptions, thereby, affecting charcoal filter bed efficiency and post
accident dose analysis.
[[Page 8806]]
Date of issuance: February 11, 1997.
Effective date: Both units, as of the date of issuance, to be
implemented within 30 days.
Amendment Nos.: 122 and 86.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64392).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 11, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Philadelphia Electric Company, Docket No. 50-353, Limerick
GeneratingStation, Unit 2, Montgomery County, Pennsylvania.
Date of application for amendment: December 6, 1996, as
supplemented by letters dated January 15, and 28, 1997.
Brief description of amendment: This amendment modifies Technical
Specification (TS) Section 2.1 and its associated TS Bases to reflect
the change in the Minimum Critical Power Ratio safety limit due to the
use of GE13 fuel product line and the cycle-specific analysis performed
by General Electric Company (GE), for LGS, Unit 2, Cycle 5.
Date of issuance: February 12, 1997.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment No.: 87.
Facility Operating License No. NPF-85. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 23, 1996 (61
FR 67582).
The January 15, and 28, 1997, letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 12, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
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: June 10, 1996, as supplemented
June 24, July 1, August 13, September 20, and October 17, 1996.
Brief description of amendments: The amendments change Technical
Specifications 3/4.3.3.1, ``Radiation Monitoring Instrumentation,'' and
3/4.7.6, ``Control Room Emergency Air Conditioning System,'' to reflect
a control room design in which the common Unit 1 and Unit 2 control
room envelope is supplied by 2 one hundred percent capable Control Room
Emergency Air Conditioning System trains.
Date of issuance: February 6, 1997.
Effective date: Both units, as of date of issuance, to be
implemented within 30 days.
Amendment Nos.: 190 and 173.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 24, 1996 (61 FR
32468) The June 24, July 1, August 13, September 20, and October 17,
1996, letters provided clarifying information that did not change the
initial proposed no significant hazards consideration determination nor
expand the scope of the initial submittal as described in the initial
notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 6, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
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: May 31, 1996, as supplemented
December 23, 1996.
Brief description of amendments: The amendments change the
Technical Specification to (1) Revise the reactor vessel level
indication system action statements, (2) revise the channel calibration
definition, and (3) delete a requirement to install la jumper in the
auxiliary feedwater actuation logic.
Date of issuance: February 6, 1997.
Effective date: Both units, as of its date of issuance, to be
implemented within 60 days.
Amendment Nos.: 191 and 174.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 17, 1996 (61 FR
30641).
The December 23, 1996, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 6, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama.
Date of amendments request: November 15, 1996.
Brief Description of amendments: The amendments replace Containment
Systems TS 3.6.2.2 for the Spray Additive System, with a new Emergency
Core Cooling Systems (ECCS) TS 3.5.6 for the ECCS Recirculation Fluid
pH Control System.
Date of issuance: February 3, 1997.
Effective date: As of the date of issuance to be implemented prior
to Mode 4 for Unit 1 following the spring 1997 refueling outage; for
Unit 2 following the spring 1998 refueling outage.
Amendment Nos.: 123 and 118.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66718).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 3, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio.
Date of application for amendment: August 7, 1996.
Brief description of amendment: The amendment revises Technical
Specification (TS) 1.0, ``Definitions,'' by defining a refueling
interval to be [less than or equal to] 730 days; and revises TS 3/4.0,
``Applicability,'' TS 3/4.6.2.1, ``Containment Systems--
Depressurization and Cooling Systems--Containment Spray System,'' and
TS 3/4.6.3.1, ``Containment Systems--
[[Page 8807]]
Containment Isolation Valves,'' to reflect performing surveillance
tests during a refueling interval rather than every 18 months.
Date of issuance: February 10, 1997.
Effective date: February 10, 1997, to be implemented not later than
120 days after issuance.
Amendment No.: 213.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52970).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio.
Date of application for amendment: September 12, 1996.
Brief description of amendment: The amendment revised Technical
Specifications (TS) 3/4.1.3.4, ``Reactivity Control Systems--Rod Drop
Time,'' and TS 3/4.5.2, ``Emergency Core Cooling Systems--Tavg [greater
than or equal to] 280 deg.F,'' to change the surveillance test interval
from every 18 months to each refueling interval ([less than or equal
to] 730 days, nominally 24 months). Additionally, the amendment removed
a footnote for TS 4.5.2.b that is no longer applicable.
Date of issuance: February 11, 1997.
Effective date: February 11, 1997, to be implemented not later than
120 days over issuance.
Amendment No.: 214.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 9, 1996.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 11, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
Dated at Rockville, Maryland, this 19th day of February 1997.
For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects--III/IV, Office of Nuclear
Reactor Regulation.
[FR Doc. 97-4573 Filed 2-25-97; 8:45 am]
BILLING CODE 7590-01-P
