[Federal Register Volume 60, Number 148 (Wednesday, August 2, 1995)]
[Notices]
[Pages 39430-39462]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-10802]
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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 July 7, 1995, through July 21, 1995. The
last biweekly notice was published on Wednesday, July 19, 1996 (60 FR
37084).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By September 1, 1995, 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
[[Page 39431]]
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, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, 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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendments request: July 3, 1995
Description of amendments request: The proposed Technical
Specification (TS) amendment temporarily adds new ACTION Statements
3.8.1.1.f and 3.8.1.1.g to TS 3.8.1.1, ``A.C. Sources - Operating,'' to
provide a method of responding to sustained degraded switchyard
voltage. Bases 3/4.8.1, ``A.C. Sources,'' 3/4.8.2, ``D.C. Sources,''
and 3/4.8.3, ``Onsite Distribution Systems,'' are also being revised to
provide guidance on how and why degraded offsite power voltage and the
number of startup transformers in service affect compliance with GDC 17
and to give the basis for the additional ACTION statements.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not significantly increase the
probability of an accident previously evaluated in the Updated Final
Safety Analysis Report (UFSAR). The safety function of the
Electrical Distribution System (EDS) is to provide sufficient
capacity and capability to assure that 1) specified acceptable fuel
design limits and design conditions of the reactor coolant pressure
[[Page 39432]]
boundary are not exceeded as a result of anticipated operational
occurrences and 2) the core is cooled and containment integrity and
other vital functions are maintained in the event of postulated
accidents. In addition, it shall have sufficient independence,
redundancy, and testability to perform its safety function assuming
a single failure. The proposed ACTIONs will restore the EDS to
conformance with General Design Criterion (GDC) 17 of Appendix A to
10 CFR 50. Once in conformance with GDC 17, the system will be
capable of performing its safety function as analyzed in Chapters 6
and 15 of the UFSAR. The proposed temporary change has no effect on
the probability of accident initiation, therefore, the probability
of an accident previously evaluated has not been significantly
increased.
The consequences of an accident previously evaluated in the
UFSAR will not be significantly increased. Restoring one train to
OPERABLE, by blocking Fast Bus Transfer (FBT), within one hour is
consistent with the response time of Technical Specification (TS)
ACTION 3.0.3. The second train will be restored to OPERABLE by
having its Emergency Diesel Generator (EDG) started, loaded, and
separated from offsite power within two hours or FBT will be blocked
within two hours. Action within two hours is consistent with the
plants TS since TS ACTION 3.8.2.1.a, ``D. C. Sources - Operating,''
would be the most limiting requirement with one train of inoperable
electric power. In a degraded voltage event, the ability of the
Class 1E 125VDC battery chargers to perform their function is
indeterminate, therefore, the Class 1E 125VDC batteries must be
assumed to provide the 125VDC control power to the Class 1E
Engineered Safety Features (ESF) circuit breakers for both of their
sequences. The battery capacity calculations assume only one
sequence. Once one train is restored to OPERABLE and the other
trains EDG demonstrated to be OPERABLE by loading and separating
from the grid, ACTION 3.8.1.1.a, for one INOPERABLE offsite power
supply, allows operation to continue for up to seventy-two hours. If
both trains are blocked, then both trains are OPERABLE.
The proposed change will ensure that the train that blocks FBT
will be in conformance with GDC 17 should a subsequent accident
occur. As such, that train of ESF equipment will be supplied Class
1E preferred and standby power in the manner assumed by Chapters 6
and 15 analyses. Starting, loading, and separating the other trains
EDG from offsite power ensures that the second train of ESF
equipment is prepared to respond to any subsequent accident. This
configuration presents one OPERABLE offsite circuit and two OPERABLE
EDGs to any subsequent accident, and would be capable of
withstanding the single failures in the UFSAR Table 15.0-0, ``Single
Failures.'' Optionally, with both trains blocked, both are OPERABLE
and would be capable of withstanding the single failures in the
UFSAR Table 15.0-0, ``Single Failures.''
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Given the current licensing basis, the proposed temporary TS
change does not create the possibility of an accident of a new or
different kind. The plant is currently licensed to have both trains
of FBT blocked when low switchyard voltages exist in order to
prevent the loss of power generated by the nuclear power unit from
causing the loss of the preferred power circuits. The proposed
temporary TS ACTIONs 3.8.1.1.f and 3.8.1.1.g are being added as
ACTIONs to prevent a double sequencing event from occurring. The
train that is blocked is consistent with previous UFSAR Chapter 6
and Chapter 15 safety analyses since it will conform to GDC 17 prior
to the onset of the accident. Under this condition it will be able
to contribute to the mitigation of an accident and withstand the
effects of any single failure equal to its ability when initially
analyzed and licensed. The EDG which is loaded and isolated from
offsite power also contributes to GDC 17 compliance since the entire
system can withstand a Loss of Offsite Power (LOP) and a single
failure of an EDG. With both trains blocked, the EDS is in
compliance with GDC 17 and is analyzed.
It is understood that an accident of a different kind will exist
if a degraded voltage condition occurs coincident with an accident
(e.g., LOCA [versus the analyzed LOP + LOCA]). Should such an
accident occur, the manual action described in the proposed ACTION
statements could not be credited to protect the plant. However, the
purpose of proposed ACTIONs 3.8.1.1.f and 3.8.1.1.g is to provide an
appropriate response to degraded voltage prior to an accident by
eliminating the malfunction of a different type (double sequencing)
and an accident of a different type (e.g., degraded voltage + LOCA)
for one train within one hour and for the second train within two
hours. This duration of response is consistent with the required
responses currently in the TSs 3.0.3, 3.8.2.1.a, and 3.8.1.1.a.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety has not been reduced in that the train
which has FBT blocked prior to the onset of an accident will be in
conformance with GDC 17 (which is the basis to TS 3/4.8.1). Since
the blocked train is in conformance with GDC 17 prior to the onset
of an accident, it will support the single failure analyses and the
safety analyses to the extent previously analyzed and licensed. The
train not blocked will have its EDG started, loaded, and separated
from offsite power prior to the end of the second hour. Action
within two hours is consistent with TS 3.8.2.1.a. The proposed
action recovers one train of A.C. sources in one hour and places the
plant in a configuration of one less power source than is required
by LCO 3.8.1.1 within two hours. Currently, TS ACTION 3.8.1.1.a (one
power source inoperable) has a duration of seventy-two hours. The
proposed ACTION requires responses within time frames consistent
with TSs 3.0.3, 3.8.2.1.a, and 3.8.1.1.a, and therefore, does not
reduce the margin of safety. Optionally, restoration of the second
train by blocking FBT within two hours is also consistent with
response times required by TS 3.0.3 and 3.8.2.1.a and therefore,
also does not reduce the margin of safety. TS 3.8.1.1.a would not be
required with both trains of FBT blocked as all four AC power
sources would then be OPERABLE.
Regulatory Guide 1.93, ``Availability of Electric Power
Sources,'' Revision 0, December 1974 recognizes that under certain
conditions it may be safer to continue operation at full or reduced
power for a limited time than to effect an immediate shutdown based
on the loss of some of the required electric power sources. In an
effort to minimize the risk to the health and safety of the public,
the proposed ACTIONs 3.8.1.1.f and 3.8.1.1.g balance the risk of a
forced shutdown against the risk of remaining at power with a
degraded switchyard voltage.
Probabilistic Risk Analysis (PRA) has compared the probability
of a core melt event for 1) blocking fast bus transfer in one train
after one hour for the next seventy-one hours, and in the second
train after two hours for the next seventy hours; 2) blocking fast
bus transfer in one train after the first hour for the next seventy-
one hours, and supplying power to the other train from the EDG after
the second hour for seventy hours; and 3) a normal shutdown assuming
the plant is in a normal configuration and no other transients or
accidents except an uncomplicated reactor trip occurs during the
shutdown process. Seventy-two hours was chosen for comparison
purposes as the proposed ACTIONs would allow operation for up to
seventy-two hours with one offsite circuit INOPERABLE.
The PRA has shown that the probability of a core melt event
during power operation with FBT blocked in one train after one hour
for the next seventy-one hours, and in the second train after two
hours for the next seventy hours is approximately 1.91E-6. The
probability of a core melt event during power operation with FBT
blocked in one train after one hour for the next seventy-one hours
and the EDG powering the opposite train after the second hour for
the next seventy hours (the proposed configuration) is between
approximately 1.91E-6 and 1.93E-6. A range is provided because the
current PRA model can only model blocking both trains or the EDGs
supplying both trains. The risk lies somewhere between the two
values. The probability of a core melt event due to a normal
shutdown assuming the plant is in a normal configuration and no
other transients or accidents except an uncomplicated reactor trip
occurs during the shutdown process is 2.4E-6. The risk can not be
calculated for a forced shutdown with degraded switchyard voltage
present but it is expected to be higher. Therefore, the analysis
provided is conservative.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
[[Page 39433]]
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: William H. Bateman
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: July 14, 1995
Description of amendment request: The proposed amendment would
change the scram insertion times, Section 3.3.C, Minimum Critical Power
Ratio section, Section 4.11.C and the associated bases in Section 2.1.1
and 3/4.3.
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:
Section 2.1 Bases - Safety Limits
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because equivalent fuel cladding protection (99.9 percent
of all fuel rods do not experience transition boiling following a
design basis transient) is provided.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed change does not affect the function of any
structure, system or component.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety because the utilization of current General Electric
fuel designs provides an equivalent margin of safety. As stated
previously, equivalent fuel cladding protection is provided and
ensures that 99.9 percent of all fuel rods will not experience
transition boiling following a design basis transient.
Section 3.3.C - Scram Insertion Times
1. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant increase in the
probability of consequences of an accident previously evaluated. The
correlation of the scram insertion times with the actual notch
position will simplify the surveillance procedure while maintaining
the accuracy of the test.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because no physical modifications are associated with the proposed
change and it does not affect the function of any structure, system
or component.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety. The notch positions were chosen to coincide with
the relative insertion values specified in the Technical
Specifications. Use of the proposed combination of notch positions
and scram insertion times will maintain the existing margins of
safety that 99.9 percent of all fuel rods will not experience
transition boiling following a design basis transient.
Section 4.11.C - Minimum Critical Power Ratio (MCPR) Calculation
Method
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the method used to calculate the measured scram
speed distribution is consistent with the PNPS [Pilgrim Nuclear
Power Station] licensing basis.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed change does not affect the function of any
structure, system or component.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in the
margin of safety because the proposed changes provide equivalent
fuel
cladding protection which ensures that 99.9 percent of all fuel
rods will not experience transition boiling following a design basis
transient.
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: 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: Ledyard B. Marsh
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station,
Units 1 and 2, Rock Island County, Illinois
Date of application for amendment requests: September 17, 1993, as
supplemented July 20, 1995
Description of amendment requests: As a result of findings by a
Diagnostic Evaluation Team inspection performed by the NRC staff at the
Dresden Nuclear Power Station in 1987, Commonwealth Edison Company
(ComEd, the licensee) made a decision that both the Dresden Nuclear
Power Station and sister site Quad Cities Nuclear Power Station needed
attention focused on the existing custom Technical Specifications (TS)
used.
The licensee made the decision to initiate a Technical
Specification Upgrade Program (TSUP) for both Dresden and Quad Cities.
The licensee evaluated the current TS for both Dresden and Quad Cities
against the Standard Technical Specifications (STS) contained in NUREG-
0123, ``Standard Technical Specifications General Electric Plants BWR/
4.'' The licensee's evaluation identified numerous potential
improvements such as clarifying requirements, changing TS to make them
more understandable and to eliminate interpretation, and deleting
requirements that are no longer considered current with industry
practice. As a result of the evaluation, ComEd has elected to upgrade
both the Dresden and Quad Cities TS to the STS contained in NUREG-0123.
The TSUP for Dresden and Quad Cities is not a complete adaption of
the STS. The TSUP focuses on (1) integrating additional information
such as equipment operability requirements during shutdown conditions,
(2) clarifying requirements such as limiting conditions for operation
and action statements utilizing STS terminology, (3) deleting
superseded requirements and modifications to the TS based on the
licensee's responses to Generic Letters (GL), and (4) relocating
specific items to more appropriate TS locations.
The September 17, 1993, and July 20, 1995, applications proposed to
upgrade only Section 3/4.7 (Containment Systems) of the Dresden and
Quad Cities TS.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis.
Implementation of these changes will provide increased reliability
of equipment assumed to operate in the current safety analysis, or
provide continued assurance that specified parameters remain within
their acceptance limits, and as such, will not significantly
increase the probability or consequences of a previously evaluated
accident.
Some of the proposed changes represent minor curtailments of the
current
[[Page 39434]]
requirements which are based on generic guidance or previously approved
provisions for other stations. The proposed amendment for Dresden
and Quad Cities Station's Technical Specification Section 3/4.7 is
based on STS guidelines or later operating BWR plants' NRC accepted
changes. Any deviations from STS requirements do not significantly
increase the probability or consequences of any previously evaluated
accidents for Dresden or Quad Cities Stations. The proposed
amendment is consistent with the current safety analyses and has
been previously determined to represent sufficient requirements for
the assurance and reliability of equipment assumed to operate in the
safety analysis, or provide continued assurance that specified
parameters remain within their acceptance limits. As such, these
changes will not significantly increase the probability or
consequences of a previously evaluated accident.
The associated systems that make up the Containment Systems are
not assumed in any safety analysis to initiate any accident sequence
for Dresden or Quad Cities Stations; therefore, the probability of
any accident previously evaluated is not increased by the proposed
amendment. In addition, the proposed surveillance requirements for
the proposed amendments to these systems are generally more
prescriptive than the current requirements specified within the
Technical Specifications. The additional surveillance requirements
improve the reliability and availability of all affected systems
and, therefore, reduce the consequences of any accident previously
evaluated, as the probability of the systems outlined within Section
3/4.7 of the proposed Technical Specifications performing their
intended function is increased by the additional surveillances.
Create the possibility of a new or different kind of accident
from any previously evaluated because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis. Other
changes represent minor curtailments of the current requirements
which are based on generic guidance or previously approved
provisions for other stations. These changes do not involve
revisions to the design of the station. Some of the changes may
involve revision in the operation of the station; however, these
provide additional restrictions which are in accordance with the
current safety analysis, or are to provide for additional testing or
surveillances which will not introduce new failure mechanisms beyond
those already considered in the current safety analyses.
The proposed amendment for Dresden and Quad Cities Station's
Technical Specification Section 3/4.7 is based on STS guidelines or
later operating BWR plants' NRC accepted changes. The proposed
amendment has been reviewed for acceptability at the Dresden or Quad
Cities Nuclear Power Stations considering similarity of system or
component design versus the STS or later operating BWRs. Any
deviations from STS requirements do not create the possibility of a
new or different kind of accident previously evaluated for Dresden
or Quad Cities Stations. No new modes of operation are introduced by
the proposed changes. Surveillance requirements are changed to
reflect improvements in technique, frequency of performance or
operating experience at later plants. Proposed changes to action
statements in many places add requirements that are not in the
present technical specifications. The proposed changes maintain at
least the present level of operability. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any previously evaluated.
The associated systems that make up the Containment Systems are
not assumed in any safety analysis to initiate any accident sequence
for Dresden or Quad Cities Stations. In addition, the proposed
surveillance requirements for affected systems associated with the
Containment Systems are generally more prescriptive than the current
requirements specified within the Technical Specifications;
therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
Involve a significant reduction in the margin of safety because:
In general, the proposed amendment represents the conversion of
current requirements to a more generic format, or the addition of
requirements which are based on the current safety analysis. Other
changes represent minor curtailments of the current requirements
which are based on generic guidance or previously approved
provisions for other stations. Some of the later individual items
may introduce minor reductions in the margin of safety when compared
to the current requirements. However, other individual changes are
the adoption of new requirements which will provide significant
enhancement of the reliability of the equipment assumed to operate
in the safety analysis, or provide enhanced assurance that specified
parameters remain with their acceptance limits. These enhancements
compensate for the individual minor reductions, such that taken
together, the proposed changes will not significantly reduce the
margin of safety.
The proposed amendment to Technical Specification Section 3/4.7
implements present requirements, or the intent of present
requirements in accordance with the guidelines set forth in the STS.
Any deviations from STS requirements do not significantly reduce the
margin of safety for Dresden or Quad Cities Stations. The proposed
changes are intended to improve readability, usability, and the
understanding of technical specification requirements while
maintaining acceptable levels of safe operation. The proposed
changes have been evaluated and found to be acceptable for use at
Dresden or Quad Cities based on system design, safety analysis
requirements and operational performance. Since the proposed changes
are based on NRC accepted provisions at other operating plants that
are applicable at Dresden or Quad Cities and maintain necessary
levels of system or component reliability, the proposed changes do
not involve a significant reduction in the margin of safety.
The proposed amendment for Dresden and Quad Cities Stations will
not reduce the availability of systems associated with the
Containment Systems when required to mitigate accident conditions;
therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Public
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Connecticut Yankee Atomic Power Company, and Northeast Nuclear
Energy Company, et al., Docket Nos.50-213, 50-245, 50-336, and 50-
423 Haddam Neck Plant, and Millstone Nuclear Power Station, Units
1,2, and 3, Middlesex County and New London County, Connecticut
Date of amendment request: June 6, 1995
Description of amendment request: The proposed amendment will
modify the size of the Plant Operations Review Committee (PORC) which
will collectively have the experience and expertise in various areas of
plant operation, and will clarify the composition of the Site
Operations Review Committee (SORC).
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration
(SHC), which is presented below:
These proposed changes do not involve an SHC because the changes
do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The PORC is an oversight group and helps to ensure that the
units are operated in a safe manner. To accomplish this the PORCs
provide their recommendations on the safety related activities to
the Vice President - Haddam Neck Plant for Haddam Neck and to the
respective Nuclear Unit Directors for Millstone. Each Millstone
Unit has its own PORC. It is proposed that the members of the
PORC be selected by the respective Nuclear Unit Director based
on their knowledge and
[[Page 39435]]
expertise in specific key plant functions. The Millstone Station has
one SORC. The SORC is also an oversight group whose charter is to
advise the Senior Vice President - Millstone Station on all matters
related to nuclear safety at the Millstone site. The Haddam Neck
Plant, being a single unit site, has one PORC, which advises the
Vice President - Haddam Neck Plant. The members of the Haddam Neck
Plant PORC will be selected by the Vice President - Haddam Neck
Plant based on their knowledge and expertise in specific key plant
functions. The PORC and SORC add to the defense-in-depth concept
provided by the design, operation, maintenance, and quality
oversight by promoting excellence through the conduct of their
affairs and by maintaining a diligent watch over their
responsibilities.
These administrative changes will revise the composition section
of the technical specifications for the PORC members. Millstone Unit
individuals will be appointed by the Nuclear Unit Directors if the
individual meets one or more of the following areas of expertise:
Plant Operations, Engineering, Reactor Engineering, Maintenance,
Instrumentation and Controls, Health Physics, Chemistry, Work
Planning and Control, and Quality Services. The Haddam Neck Plant,
due to its broader scope of review also include[s] an individual
experienced in Security and specific experience in Electrical
Maintenance and Mechanical Maintenance. The individuals who will
serve on PORC shall continue to meet the criteria of ANSI N18.1-
1971. This approach is consistent with the standard technical
specifications and NUREG 0800, Section 13.4. For SORC at the
Millstone Station, the method of identifying who shall serve as Vice
Chairperson has been modified for clarity. The Site Services
Director position is proposed to be eliminated since this position
no longer exists. The functions previously performed by this
individual have been assumed by those individuals who currently
serve on SORC. Finally, [the TS relating to] the individual who
shall represent Quality and Assessment Services shall be modified to
allow a qualified member of Quality and Assessment Services to serve
on SORC.
The remaining portions of the technical specifications related
to PORC and SORC are not being revised.
These modifications broaden the unit committee participation and
reflect current organizational positions and will not increase the
probability of occurrence or the consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed administrative enhancements to the composition of
the PORC and Millstone Station SORC will not affect the way in which
the units are physically operated. These administrative changes to
PORC and SORC continue to meet the guidelines of ANSI N18.7-1976.
The modifications to PORC and SORC continue to allow these groups to
provide a thorough review of activities at the units.
The proposed modification does not impact any initiating events,
and, therefore, cannot create the possibility of any new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
These proposed administrative changes will not impact the margin
of safety provided by PORC and SORC. The PORC and SORC will continue
to be staffed by qualified individuals experienced in the operation
of the plants. These administrative changes will modify how the
composition of the PORC and SORC members are presented in the
technical specifications, but will not adversely impact their
ability to review and comment on operations at the units.
These changes do not impact any protective boundaries nor do
they impact the safety limits for the protective boundaries. These
proposed changes are administrative in nature. Therefore, there is
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: Russell Library, 123 Broad
Street Middletown, Copnnecticut 06457, for the Haddam Neck Plant, and
the Learning Resources Center, Three Rivers Community-Technical
College, 574 New London Turnpike, Norwich, CT 06360, for Millstone 1,
2, and 3.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: July 5, 1995
Description of amendment request: The proposed amendment would
change the Administrative Controls section of the Palisades Technical
Specifications. The changes involve deleting training requirements in
the Administrative Controls section, revising the Plant Review
Committee composition, and revising the function and composition of the
plant safety and licensing staff review requirements.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
A. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
This change does not affect the probability or consequences of
an accident. The changes are administrative, deleting an unnecessary
specification on staff training requirements, eliminating the
specific references to the Nuclear Engineering and Construction
Organization (NECO) staff, and requiring that the Plant Review
Committee (PRC) chairman, alternate chairman, and members be
designated in administrative procedures by the Plant General
Manager. Further administrative changes clarify the function of the
Plant Safety and Licensing organization and eliminate the numerical
requirement for five staff members to fulfill the organization
function.
The removal of an obsolete staff training requirement does not
diminish the regulatory requirement to have an adequately trained
staff. The accredited training programs for the plant staff ensure
an appropriate level of training is conducted to maintain an
appropriate skill and knowledge base for the staff. The requirements
of 10CFR55 provide the necessary rules for operator licenses. Since
a trained staff will be maintained, there will [be] no increase in
the probability or consequences of an accident as a result of this
change.
The composition of the PRC will not be affected by this change
as it will, at a minimum, be comprised of personnel from the
operations, engineering, radiological services and maintenance
departments as required by the Technical Specifications. The
composition of the Plant Safety and Licensing organization as a
whole may change. The function of the organization as it relates to
these Technical Specifications, however, will not be affected. These
changes have no affect on the plant accident analyses. Qualified
personnel will continue to conduct the PRC and Plant Safety and
Licensing reviews. Therefore, the changes do not increase the
probability or consequences of an accident.
B. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed changes are administrative and do not create the
possibility of a new or different kind of accident. Staff training
will continue to meet the accreditation requirements of the National
Academy for Nuclear Training Accreditation Board and the
requirements for the Systematic Approach to Training. Operators'
license training will continue to meet the regulatory requirements
of 10CFR55. Activities conducted by the Plant Review Committee and
the Plant Safety and Licensing staff will continue to be
accomplished by a staff which meets the qualification requirements
of the Technical Specifications. These administrative changes will
not affect the operation of the plant or the safety function of
plant equipment nor will it affect the quality of the review
activities. Therefore, there will be no possibility that a new or
different kind of accident will be created.
C. Involve a significant reduction in a margin of safety.
The changes do not affect installed plant equipment nor do they
affect plant
[[Page 39436]]
operations. These administrative changes have not affected the
probability or consequences of a previously analyzed accident or
created the possibility of a new or different kind [of] accident
from any previously evaluated. Therefore, they do not involve any
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
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley
Power Station, Unit No. 1, Shippingport, Pennsylvania
Date of amendment request: October 11, 1994, as supplemented June
23, 1995.
Description of amendment request: The proposed amendments would
revise Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-
2) Technical Specifications (TSs) 1.18, ``Quadrant Power Tilt Ratio,''
3/4.2.4, ``Quadrant Power Tilt Ratio,'' the Table Notation of TS Table
3.3.-1, ``Reactor Trip System Instrumentation,'' and associated Bases
to incorporate the guidance provided in the NRC's Improved Standard
Technical Specifications (NUREG-1431) applicable to these TSs. The
proposed amendments would clarify the requirements of the subject TSs
with regard to the use of excore power range neutron flux detectors to
monitor quadrant power tilt ratio when an excore power range neutron
flux instrument is inoperable. The proposed change would also make
several minor editorial changes in the subject TSs.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The existing quadrant power tilt ratio (QPTR) definition and
Surveillance Requirement (SR) 4.2.4.c are inconsistent concerning
reactor power limitations when performing QPTR surveillance
requirements. The proposed change modifies these and related
requirements to improve the understanding and consistency by
generally incorporating the Improved Standard Technical
Specification (ISTS) requirements of NUREG-1431.
Editorial changes have been incorporated throughout the proposed
specifications to address ISTS or plant specific convention and do
not affect the accident analyses. The QPTR definition has been
modified to reflect the ISTS wording and eliminate the inconsistency
with SR 4.2.4.c. This change does not reduce the QPTR testing
requirements or affect the accident analyses assumptions. The
current action statements require power reduction along with a
reduction in power range high neutron flux trip setpoints when the
QPTR exceeds the limit. This ensures the core conditions are
consistent with the accident analyses assumptions. With the modified
action statements and the QPTR exceeding the limit, power reduction
is also required along with performing a flux map to verify the
peaking factors are within the accident analyses assumptions. In
addition, the safety analyses must be re-evaluated to confirm the
results remain valid prior to increasing power with an indicated
tilt condition. The new action statements provide methods different
from the current requirements. However, they satisfy the same
objective, to ensure the conditions assumed in the accident analyses
are maintained. Therefore, these changes will not involve
significant increase in the probability or consequences of an
accident previously evaluated.
The current surveillance requirements define the methods and
frequencies for verifying the QPTR is within the limit specified in
the limiting condition for operation. The proposed SRs include
associated notes that allow separation of a power range channel into
two portions made-up of the Nuclear Instrumentation System (NIS) and
the excore detector portion. If an excore detector portion of a
power range channel is inoperable, then the power range channel is
inoperable since the detector provides input to the NIS which inputs
to the solid state protection system. However, if the excore
detector is operable and the NIS is inoperable, then the power range
channel is inoperable but the ability to monitor the QPTR is
unaffected. When the NIS portion of a channel is inoperable,
appropriate actions are applied in accordance with Specification
3.3.1. The new SRs continue to require the same testing and
frequencies as the current SRs along with reducing the need to
interpret the requirements when special conditions exist. Therefore,
the proposed SRs will not affect the accident analyses or
significantly increase the probability or consequences of an
accident previously evaluated.
Table 3.3-1 Action 2 applies when a power range channel is
inoperable. This action has been reformatted to incorporate changes
similar to those adopted in the QTPR SR which allow separation of a
power range channel into the NIS portion and the excore detector
portion. Proposed Action 2.a applies to an inoperable power range
high neutron flux channel and Action 2.b applies to ``all other
channels'' which includes the Low Setpoint function along with the
High Positive and High Negative Rate functions. The new action is
modified by Note (3) to allow bypassing the inoperable channel for
surveillance testing and setpoint adjustment and by Note (4) that
only requires performing SR 4.2.4 when the power range high neutron
flux channel input to QPTR is inoperable. The new action does not
require reducing the power range neutron flux setpoint like the
current action since the proposed action is to perform the QPTR
surveillance or shutdown which is more conservative than the current
action requirement, otherwise, the new action requires essentially
the same steps to be performed as the current action. Therefore, the
proposed action will not affect the accident analyses or involve a
significant increase in the probability or consequences of an
accident previously evaluated.
These changes are proposed to allow flexibility in plant
operations by modifying the QPTR action and surveillance
requirements to allow separation of a power range channel into the
NIS portion and the excore detector portion. The modified action and
surveillance requirements continue to provide monitoring of those
parameters required to ensure the core is operating safely. Since
these changes are not significantly different from the current
requirements and no change is being introduced that would affect the
accident analyses assumptions, we have concluded that the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes incorporate modifications generally
consistent with the ISTS QPTR requirements to ensure the core power
distribution is adequately monitored. The revised action statements
provide for peaking factor verification as a logical compensatory
measure to ensure the core is operating within required limits. This
is more conservative than the current requirements and provides
additional assurance that Specification 3.2.4 will continue to
govern the QPTR limitations in a manner consistent with the accident
analyses assumptions. The revised SR provides clear and
understandable testing requirements to reduce confusion concerning
how the QPTR is to be monitored based on plant conditions. The
proposed change does not introduce any new mode of plant operation
or require any physical modification to the plant, therefore, this
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The QPTR limit ensures that the gross radial power distribution
is maintained within the assumptions used in the safety analyses.
The QPTR is one of the variables that is monitored to ensure the
core operates within the bounds used in the safety analyses. When
the QPTR is maintained below 1.02 it provides an indication that the
peaking factors are within the limiting values by preventing and
undetected change in the
[[Page 39437]]
gross radial power distribution. The proposed changes ensure the
required parameters are verified during the applicable conditions
and on a consistent basis, therefore, these changes will not reduce
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas
Date of amendment request: May 19, 1995
Description of amendment request: The proposed amendments revise
the specifications to permit the reactor building personnel airlock
doors to remain open during fuel handling.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed change would allow the containment personnel
airlock doors to remain open during fuel movement and core
alterations. These doors are normally closed during this time period
in order to prevent the escape of radioactive material in the event
of a fuel handling accident. These doors are not initiators of any
accident. The probability of a fuel handling accident is unaffected
by the position of the containment personnel airlock doors.
The proposed change alters assumptions made in evaluating the
radiological consequences of a fuel handling accident inside the
reactor containment building. Allowing the containment personnel
airlock doors to remain open during fuel movement and core
alterations does increase, however not significantly, the
consequences of a fuel handling accident inside containment.
Previously, the fuel handling accident inside containment was
bounded by the fuel handling accident analysis in the spent fuel
pool area of the auxiliary building. Part of the dose increase has
been offset by the increase in the minimum decay time before
irradiated fuel may be moved inside the reactor containment
building. Extending the minimum decay time actually decreases the
consequences of a fuel handling accident by reducing the radioactive
inventory of the irradiated fuel which could possibly be released
during a fuel handling accident. The revised fuel handling accident
analysis results in maximum offsite doses of 43.4 Rem and 41.8 Rem
to the thyroid and 0.616 Rem and 0.598 Rem to the whole body for
ANO-1 and ANO-2, respectively. The calculated offsite doses are well
within the limits of 10CFR Part 100. Also, the calculated doses are
larger than the actual doses which would be expected during a fuel
handling accident because the calculation does not incorporate the
closing of at least one of the personnel airlock doors following
evacuation of containment. The proposed change would significantly
reduce the dose to workers in the containment in the event of a fuel
handling accident by expediting the containment evacuation process.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed change does not involve the addition or
modification of any plant equipment. Also, the proposed change would
not alter the design, configuration, or method of operation of the
plant.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
This proposed change has the potential for an increased dose at
the site boundary due to a fuel handling accident; however, the dose
remains within acceptable limits. The margin of safety as defined by
10CFR Part 100 has not been significantly reduced. There is an
increase in the calculated offsite dose resulting from a fuel
handling accident; however, the increase is not significant and is
well within the limits specified in 10 CFR Part 100. The overall
significance will be offset by the increased minimum decay time, the
decreased potential radiation dose to workers, and the increased
availability of the personnel airlock door in the event of a fuel
handling accident. Closing at least one of the personnel airlock
doors following an evacuation of containment, further reduces the
offsite doses in the event of a fuel handling accident which
partially compensates for the higher offsite doses calculated as a
result of this proposed change.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. BecknerEntergy Operations, Inc.,
Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County,
Arkansas
Date of amendment request: March 17, 1995
Description of amendment request: The proposed amendment revises
requirements associated with channel functional tests of the core
protection calculator following a high temperature alarm.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The core protection calculators (CPCs) are not accident
initiators, therefore this change does not increase the probability
of an accident previously evaluated.
The core protection calculators (CPCs) are dedicated
minicomputers that receive key parameters necessary to calculate the
departure from nucleate boiling ratio (DNBR) and local power density
(LPD) and issue a reactor trip command prior to reaching plant
conditions that may damage the fuel in the reactor. Subjecting a
computer to elevated temperatures may affect the reliability of the
computer calculations. This change in the Arkansas Nuclear One-Unit
2 (ANO-2) Technical Specifications (TS) will require a verification
of the CPC operability, by the performance of a channel functional
test, in the event a cabinet high temperature switch is actuated.
This is a more accurate indication of the operating environment of
the CPCs than the current requirement to perform the test based upon
room temperature. The ability of the CPCs to monitor DNBR and LPD
and issue a trip command when appropriate will not be affected in
any way by this change, therefore the consequences of an accident
previously evaluated are not increased.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
Because the proposed changes do not alter the design,
configuration, or method of operation of the plant, they do not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
These proposed changes do not alter the acceptance criteria of
any surveillance requirements. The changes do not alter any
assumptions used in accident analysis, change any actuation
setpoints, nor allow
[[Page 39438]]
operations in any configuration not previously analyzed. This change
will trigger a verification of affected CPC operability based on
cabinet temperature instead of room temperature, which is a more
accurate indication of the operating environment of the CPC
computer. Therefore, this change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: April 4, 1995
Description of amendment request: The proposed amendment revises
operating criteria and requirements associated with containment
personnel air locks.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The containment air locks are passive components integral to the
containment structure and are not evaluated to be accident
initiators, therefore, the proposed amendment does not involve an
increase in the probability of an accident previously evaluated.
Each air lock door is rated for and tested to full design
pressure of the containment building. If one door were inoperable in
each air lock, the remaining door, since required to remain closed
and locked, would provide the necessary fission product barrier to
prevent an uncontrolled release, therefore the amendment allowance
for an inoperable air lock door in each air lock does not increase
the consequences of any previously evaluated accident.
During a situation where one containment air lock door is
inoperable and the operable door is opened, a breech in containment
integrity would essentially exist while the operable door remains
open. The time required for a containment air lock door to be open
for ingress or egress does not exceed two to three minutes. The
amendment provision to allow unlocking and opening an operable air
lock door for ingress and egress to facilitate air lock maintenance
necessary to restore operability does not increase the consequences
of any previously evaluated accident since the time necessary for
the door to be open is bounded by the existing one hour time
allowance for an actual breech of containment integrity (TS
3.6.1.1.)
The containment air lock interlock functions to prevent
simultaneous opening of both air lock doors thereby creating a
breech in containment integrity. A dedicated individual stationed at
the air lock to administratively control door operations, or locking
closed an operable door will adequately assure containment
integrity. The addition of this technical specification action
statement, therefore, does not increase the consequences of any
previously evaluated accident.
Performance of the overall air lock leakage test requires
opening the outer air lock door for installation of the mechanical
dogging devices on the inner door. The current technical
specifications make no provisions for this entry and thus would
require a plant shutdown if the inner door was inoperable in an air
lock. The proposed amendment removes the requirement to shut down
when the barrel leak rate is due. The time required for the
containment air lock doors to be opened for dog installation would
be the same as for ingress and egress as discussed above, therefore
this change does not increase the consequences of any previously
evaluated accident.
10 CFR 50, Appendix J contains containment leakage testing
requirements, including specific requirements for containment
building air locks. Changing the TS surveillance requirements to
refer to 10 CFR 50, Appendix J for these test requirements will not
degrade these tests, therefore this change does not increase the
consequences of any previously evaluated accident.
The air lock door seal pressure test is performed any time the
air lock is used for containment access during modes of operation
when containment integrity is required. The door seal test is
intended to be a gross test to verify that the door seals were not
damaged during the opening and closing cycle(s). This test does not
replace the required overall barrel leakage test. Based on
information provided by the air lock vendor, a test pressure of 10
psig is sufficient to perform this gross seal verification. A change
in the allowable leakage rate is requested to remove a specific
numerical value from the TS surveillance requirements section and
replace it with a fraction of LaG. This new acceptable leakage
rate remains relatively insignificant and is bounded by the overall
air lock leakage rate. Based on these facts this change in test
pressure and associated acceptance criteria does not increase the
consequences of any previously evaluated accident.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
Because the proposed changes do not change the design,
configuration, or method of operation of the plant, they do not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed changes to ANO-2 TS involve allowing brief breaches
in containment integrity for the purpose of repairing inoperable air
lock components or performing surveillances required by 10 CFR 50,
Appendix J. These cases are adequately bounded by the one hour
allowable outage time afforded by TS 3.6.1.1.
The addition of a specific action statement addressing an
inoperable air lock interlock provides those actions necessary to
assure the maintenance of containment integrity. This is achieved by
locking an operable door in the affected air lock when not in use
and stationing a dedicated individual at the air lock, during
periods of ingress and egress, whose sole responsibility is to
insure only one air lock door is opened at a time thereby
duplicating the function of the mechanical interlock.
The proposed changes also consist of administrative changes
removing an outdated exemption to 10 CFR 50, Appendix J and removing
specific surveillance requirements from the specifications, instead
referring to the controlling requirements of 10 CFR 50, Appendix J.
This is consistent with the provisions of NUREG 1432 ``Revised
Standard Technical Specifications for Combustion Engineering
Plants,'' Rev. 0.
None of the proposed changes increase the allowable overall air
lock leakage rate, nor affect the acceptance criteria of the overall
integrated containment leakage rate. All of the changes are bounded
by existing analyses for all evaluated accidents and do not create
any situations that alter the assumptions used in these analyses.
Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: May 19, 1995
[[Page 39439]]
Description of amendment request: The proposed amendment adds
criteria to address optional inspections of steam generator tubes.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
Steam generator tubes are inspected on a periodic basis to
reduce the probability of a steam generator tube rupture or tube
leakage. Five special interest groups are being added for optional
inspections in addition to the general tube inspections currently
required by the technical specifications. These special interest
groups define areas of tubes where known or potential degradation
mechanisms may exist for which additional inspection, above that
currently required in the technical specifications, may be
beneficial. Inspection of these special interest groups may utilize
probes which more readily detect indications which may be found in
the special interest areas. The increased detection capability will
reduce the probability that a structurally significant flaw will go
undetected during an inspection. The minimum sample size and
expansion criteria (should a flaw be found) for inspections of
special interest groups are based on percentages of tubes
potentially affected by the specific degradation mechanisms for
which the special inspection is being performed. The percentages
used are the same as used for the current general tube inspections.
The expansion criteria allow expansion within the area of interest
without affecting the expansions of any general tube inspection. By
expanding within the area of interest, a more complete inspection
for the defects caused by a specific degradation mechanism can be
performed than if the expansion were conducted in tubes not
necessarily affected by the degradation mechanism, which is possible
with the current technical specifications. Therefore, this change
does not involve a significant increase in the probability of an
accident previously considered.
The proposed change does not increase the amount of radioactive
material available for release or modify any systems used for
mitigation of such releases during accident conditions. The steam
generator tubing will continue to be examined on the frequency
currently specified in the technical specifications. This change
will allow steam generator examinations to focus on known areas of
interest without requiring unnecessary expansion. The integrity of
the steam generators will continue to be assured at an equivalent
level. Therefore, the change does not involve a significant increase
in the consequences of any accident previously evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
Special inspections such as the ones being added to the
technical specifications have been conducted in the past at ANO-2.
The method of inspection, pushing or pulling a probe through the
steam generator tubes from the primary side, is the same method
employed for the current technical specification required
inspections. Inspection methodology is not being changed by
incorporation of these special interest groups into the technical
specifications. No design or operational characteristics of the
plant are changed by the proposed amendment.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed amendment adds special interest groups for optional
inspection into the technical specifications. These inspections
concentrate on areas of interest using inspection methodology that
is equivalent or better at finding specific types of flaws than the
methodology used for the currently required general tube
inspections. If the special interest groups are not inspected, the
existing technical specification requirements for inspection still
apply.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: May 19, 1995
Description of amendment request: The proposed amendment increases
the allowed outage time for an emergency diesel generator from 72 hours
to seven days. Additionally, the amendment authorizes one, ten-day
diesel generator maintenance outage every fuel cycle.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The emergency diesel generators (EDGs) are backup alternating
current power sources designed to power essential safety systems in
the event of a loss of offsite power. EDGs are not an accident
initiator in any accident previously evaluated. Therefore, this
change does not involve an increase in the probability of an
accident previously evaluated.
The EDGs provide backup power to components that mitigate the
consequences of accidents. The proposed changes to allowed outage
times (AOTs) do not affect any of the assumptions used in
deterministic safety analysis.
In order to fully evaluate the EDG AOT extension, probabilistic
safety analysis methods were utilized. The results of these analyses
indicate no significant increase in the consequences of an accident
previously evaluated. These analyses are detailed in CE NPSD-996,
Combustion Engineering Owners Group ``Joint Applications Report for
Emergency Diesel Generators AOT Extension.''
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
This proposed change does not alter the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed changes do not affect the technical specification
limiting conditions for operation or their bases which support the
deterministic analyses used to establish the margin of safety.
Evaluations used to support the requested technical specification
changes have been demonstrated to be either risk neutral or risk
beneficial. These evaluations are detailed in CE NPSD-996.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,
Unit No. 2, Pope County, Arkansas
Date of amendment request: May 19, 1995
Description of amendment request: The proposed amendment increases
the allowed outage time for an inoperable
[[Page 39440]]
Safety Injection Tank (SIT) from one hour to 24 hours. Additionally,
the amendment limits power operation to 72 hours when certain SIT
related instrument functions are inoperable.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The Safety Injection Tanks (SITs) are passive components in the
Emergency Core Cooling System. The SITs are not accident initiators
in any accident previously evaluated. Therefore, this change does
not involve an increase in the probability of an accident previously
evaluated.
SITs were designed to mitigate the consequences of Loss of
Coolant Accidents (LOCA). These proposed changes do not affect any
of the assumptions used in deterministic LOCA analysis. Therefore,
the consequences of accidents previously evaluated do not change.
In order to fully evaluate the effect of the SIT Allowable
Outage Time (AOT) extension, probabilistic safety analysis (PSA)
methods were utilized. The results of these analyses show no
significant increase in the core damage frequency. As a result,
there would be no significant increase in the consequences of an
accident previously evaluated. These analyses are detailed in CE
NPSD-994, Combustion Engineering Owners Group ``Joint Applications
Report for Safety Injection Tank AOT/STI Extension.''
The change pertaining to SIT inoperability based solely on
instrumentation malfunction does not involve a significant increase
in the consequences of an accident as evaluated and endorsed by the
NRC in NUREG-1366, ``Improvements to Technical Specifications
Surveillance Requirements.''
Therefore, this change does not involve an increase in the
probability or a significant increase in the consequences of any
accident previously evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
This proposed change does not change the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed changes do not affect the limiting conditions for
operation or their bases that are used in the deterministic analyses
to establish the margin of safety. PSA evaluations were used to
evaluate these changes. These evaluations demonstrated that the
changes are either risk neutral or risk beneficial. These
evaluations are detailed in CE NPSD-994.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear
One, Unit No. 2, Pope County, Arkansas
Date of amendment request: May 19, 1995
Description of amendment request: The proposed amendment increases
the allowed outage time for one train of low pressure safety injection
from 72 hours to seven days.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The low pressure safety injection system (LPSI) is part of the
Emergency Core Cooling System subsystem. Inoperable LPSI components
are not considered to be accident initiators. Therefore, this change
does not involve an increase in the probability of an accident
previously evaluated.
The LPSI system was designed to mitigate the consequences of a
large loss of coolant accident (LOCA). These proposed changes do not
affect any of the assumptions used in deterministic LOCA analysis.
In order to fully evaluate the LPSI AOT extension, probabilistic
safety analysis methods were utilized. The results of these analyses
indicate no significant increase in the consequences of an accident
previously evaluated. These analyses are detailed in CE NPSD-995,
Combustion Engineering Owners Group ``Joint Applications Report for
Low Pressure Safety Injection System AOT Extension.''
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
This proposed change does not change the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed changes do not affect the technical specification
limiting conditions for operation or their bases which support the
deterministic analyses used to establish the margin of safety.
Probabilistic evaluations used to support the requested technical
specification changes have been demonstrated to be either risk
neutral or risk beneficial. These evaluations are detailed in CE
NPSD-995.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: June 26, 1995
Description of amendment request: The amendment revises the snubber
visual inspection intervals to match the schedule developed by the NRC
staff for use with a 24 month refueling interval. This schedule was
documented in Generic Letter 90-09. The licensee has made wording
changes not contained in Generic Letter 90-09. These changes are as
follows:
a) Section 4.5.Q.1 - GL 90-09 wording ''...performance of the
following augmented inservice inspection program in addition to the
requirements of Section 4.0.5.''
Proposed Technical Specification wording ''...performance of the
following inspection program.''
b) Section 4.5.Q.1.a - GL 90-09 wording ''...based on the criteria
of Table 4.7.2 and the first inspection interval determined using the
criteria shall be based upon the previous inspection interval
established by the requirements in effect before Amendment (*).
``Proposed Technical Specification wording ''...based on the criteria
provided in Table 4.5.1.''
c) Section 4.5.Q.1.b - GL 90-09 wording ''...All snubbers found
connected to an inoperable common hydraulic fluid reservoir shall be
[[Page 39441]]
counted as unacceptable for determining the next inspection interval.''
Proposed Technical Specification deletes this sentence.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
The proposed amendment would revise the basis for the snubber
visual inspection to be consistent with the bases described in Generic
Letter 90-09.
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not affect the probability of
occurrence nor does it affect the consequences of an accident
previously evaluated as the requested visual inspection interval has
been determined generically to be a safe and acceptable alternative
to the existing visual inspection requirements as documented by the
NRC in Generic Letter 90-09. With the completion of over 25 years of
operating experience and only detecting one visual inspection
failure, GPU Nuclear agrees that the existing intervals are overly
conservative and can be extended to those described in the generic
letter.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
As the requested change deals only with the frequency of visual
inspection and not with the content, scope, or acceptance criteria
of the inspection, no new or different type of accident has been
created.
3. Involve a significant reduction in the margin of safety.
The margin of safety as defined in the bases of the Technical
Specifications is not reduced as the requested requirements provide
the same degree of confidence in snubber operability at the existing
requirements.
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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Phillip F. McKee
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: May 31, 1995
Description of amendment request: The proposed amendment would
modify (by relocation to the Technical Requirements Manual) Technical
Specification (TS) 3/4.1.2.1, Boration Systems/Flow Paths - Shutdown,
TS 3/4.1.2.2, Boration Systems/Flow Paths - Operating, TS 3/4.1.2.3,
Charging Pumps - Shutdown, TS 3/4.1.2.4, Charging Pumps - Operating, TS
3/4.1.2.5, Borated Water Sources - Shutdown, TS 3/4.1.2.6, Borated
Water Sources - Operating, TS 3/4.4.2.1, Safety Valves - Shutdown, and
the associated Bases.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change to the subject Technical Specifications is
of an administrative nature in that the subject Technical
Specifications and Bases will be relocated in their entirety to the
Technical Requirements Manual. Future changes to the relocated
requirements will be in accordance with 10CFR50.59 and approved
station procedures.
Whether the listed Technical Specifications and Bases are
located in Technical Specifications or the Technical Requirements
Manual has no effect on the probability or consequences of an
accident previously evaluated.
The proposed change does not alter the assumptions previously
made in the listed Technical Specifications. The proposed change
allows the Commission and the South Texas Project more effective use
of personnel resources to control requirements that meet the four
Criteria in the Final Policy Statement. The proposed change will not
change the dose to workers.
Since the probability of an accident is unaffected by
administratively relocating the subject Technical Specification, and
the doses are not affected and do not exceed acceptance limits, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change to the subject Technical Specifications is
of an administrative nature in that the subject Technical
Specifications and Bases will be relocated in their entirety to the
Technical Requirements Manual. Future changes to the relocated
requirements will be in accordance with 10CFR 50.59 and approved
station procedures. Whether the listed Technical Specifications and
Bases are located in Technical Specifications or the Technical
Requirements Manual has no effect on any previously evaluated
accident. It does not represent a change in the configuration or
operation of the plant and, therefore, does not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in
the margin of safety?
The proposed change to the subject Technical Specifications is
of an administrative nature in that the subject Technical
Specifications and Bases will be relocated in their entirety to the
Technical Requirements Manual. Future changes to the relocated
requirements will be in accordance with 10CFR50.59 and approved
station procedures. The margin of safety is not reduced when the
requirements are relocated to a Licensee-controlled document because
the requirements to change a License Basis Document via the
10CFR50.59 process ensure the same questions concerning the margin
of safety required for license amendments are asked. Therefore, this
proposed change does not significantly reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior
College, J. M. Hodges, Learning Center, 911 Boling Highway, Wharton,
Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: William D. Beckner
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: June 28, 1995
Description of amendment request: The proposed amendment would
revise technical specifications related to the standby liquid control
(SLC) system. The proposed changes include increasing the required
reactor pressure vessel boron concentration and modifying the SLC pump
operability testing surveillance frequency from monthly to quarterly.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the
[[Page 39442]]
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 current analysis requires the SLC system to be
capable of bringing the reactor 3% delta - k subcritical assuming a
cold xenon free condition. The increase in SLC storage tank boron
concentration limits will ensure this capability is maintained for
future reload cores using the same 3% delta - k shutdown reactivity
margin without imposing restrictions in cycle exposure for current
and future anticipated core configurations. The change in the
surveillance frequency for SLC pump operability testing to once each
three months is in agreement with the ASME Code. The relaxation of
the testing interval for the SLC pumps decreases pump degradation,
and eliminates an unnecessary burden on personnel resources without
compromising plant safety. In addition, the administrative changes
only correct typographical and editorial errors.
Since these proposed changes do not affect precursors for any
accident or transient analyzed in Chapter 14 of the USAR, there is
no increase in the probability of any accident previously evaluated.
Furthermore, since these changes will ensure the ability of the SLC
system to mitigate the consequences of an accident for future
anticipated core designs, they do not involve a significant increase
in the consequences of any accident previously evaluated.
2. The proposed changes will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The change in the SLC storage tank boron concentration
limits will ensure that a cold xenon-free reload core can be brought
to a subcritical condition as previously analyzed. The change in the
frequency of the SLC pump operability testing to once each three
months is in agreement with the ASME Code. The relaxation in the
testing interval for the SLC pumps decreases pump degradation, and
eliminates an unnecessary burden on personnel resources without
compromising plant safety. In addition, the administrative changes
only correct typographical and editorial errors.
These proposed changes do not affect the design, function, or
operation of the SLC or any other system. Also, these changes do not
introduce any new modes of operation or modify existing equipment
design. Therefore, they do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed changes will not create a significant reduction
in the margin of safety. The proposed increase in the required boron
concentration in the reactor pressure vessel will ensure the SLC
system will be capable of bringing a cold xenon-free reload core
subcritical while maintaining the 3% delta - k shutdown reactivity
margin as specified in the previous operating cycle. The change in
the frequency of SLC pump operability testing to once each three
months is in agreement with the ASME Code. The relaxation in the
testing interval for the SLC pumps decreases pump degradation, and
eliminates an unnecessary burden on personnel resources without
compromising plant safety. In fact, it increases SLC system
availability. In addition, the administrative changes only correct
typographical and editorial errors. Therefore, it is concluded that
the requested changes do not create a significant reduction in the
existing margin of safety as defined in the Technical
Specifications.
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: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499
NRC Project Director: William D. Beckner North Atlantic Energy
Service Corporation, Docket No. 50-443, Seabrook Station, Unit No. 1,
Rockingham County, New Hampshire
Date of amendment request: June 16, 1995
Description of amendment request: The proposed amendment would
change the minimum boron concentration specified for the refueling
water storage tank (RWST) in Limiting Condition for Operation (LCO) in
Technical Specification (TS) 3.1.2.5 and would replace the minimum
specified concentration for boron with an acceptable range of boron
concentration for the RWST and the accumulators in the LCOs for TS
3.1.2.6, 3.5.1.1, and 3.5.4.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration. The
NRC staff has reviewed the licensee's analysis against the standards of
10 CFR 50.92(c). The NRC staff's review is presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10 CFR
50.92(c)(1)) because the changes are proposed to assure that the post-
event shutdown margin required by the Technical Specifications will
continue to be met and the consequences of a boron dilution event will
remain as previously evaluated. The changes do not affect the design or
manner of operation of any structure, system, or component important to
safety.
B. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because they do not affect the manner by which the
facility is operated and do not involve a change to any structure,
system, or component important to safety. The proposed changes merely
assure that station will be operated within original design limits.
C. The changes do not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)) because the proposed changes merely
assure that the station will continue to be operated within the
original design limits. Therefore, the acceptance criteria for
previously evaluated accidents will continue to be met.
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: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One
International Place, Boston MA 02110-2624.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: July 11, 1995
Description of amendment request: The proposed amendment modifies
Technical Specification 3.5.F.7 to also allow the use of pull-to-lock
switches to defeat the automatic initiation of the emergency core
cooling system (ECCS) while in the refuel condition. The proposed
amendment also makes administrative changes and makes changes to the
associated Bases section.
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:
NNECO has reviewed the proposed change in accordance with 10 CFR
50.92 and concluded that the change does 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 change does not
[[Page 39443]]
involve an SHC because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
This change to LCO [Limiting Condition for Operation] 3.5.F.7(e)
will allow an alternative means of de-energizing power to the
selected ECCS pump motors during refueling. The current
technical specification already allows these motors to be de-
energized. Use of the pull-to-lock switches provides a safer method
of achieving this condition. The pull-to-lock condition of the
switches is annunciated in the control room. Therefore, the switches
will not be inadvertently left in the pull-to-lock position.
Deletion of the statement that the 4160 volt supply breakers are
racked in does not affect the requirement of LCO 3.5.F.7 to ensure
the specified ECCS subsystems are OPERABLE.
Therefore, there is no change in the probability or consequences
of an accident previously analyzed due to this change.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The use of an alternative means of de-energizing power from the
selected ECCS pump motors does not create a possibility of a new or
different kind of accident. Using the control room pull-to-lock
switch to disable the pump motor circuit breaker has the same effect
on the ECCS pump as the removal of the circuit breaker from the
switchgear.
Deletion of the statement that the 4160 volt supply breakers are
racked in does not affect the requirement of LCO 3.5.F.7 to ensure
the specified ECCS subsystems are OPERABLE.
3. Involve a significant reduction in the margin of safety.
The proposed change to the Millstone Unit No. 1 Technical
Specifications does not reduce the margin of safety. By using the
control room pull-to-lock switches to disable the ECCS pump motors,
instead of racking out the pump motor circuit breakers, it is
possible to reenergize the ECCS pumps more quickly in an emergency,
should one occur. The time savings can be translated into added
safety margin from a shutdown risk perspective. The ability to
disable and enable the pumps from the control room, instead of the
switchgear area, also contributes to this added safety margin.
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.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: July 18, 1995
Description of amendment request: The proposed amendment request
will add operability and surveillance requirements for reactor pressure
vessel (RPV) overfill protection instrumentation. The proposed
amendment will also add the associated Bases.
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:
NNECO has reviewed the proposed change in accordance with 10 CFR
50.92 and concluded that the change does 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 change does not involve an SHC because the change would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The new LCO [Limiting Condition for Operation] and surveillance
requirements ensure that the reactor high water level feedwater pump
trip instrumentation is available. This technical specification
change does not involve the addition of new equipment or logic. This
change does not add new surveillance requirements for the
instrumentation. This change simply establishes requirements for the
operation and surveillance of
reactor high water level feedwater pump trip instrumentation in
the technical specifications. The implementation of this technical
specification change will decrease the likelihood of an RPV
overfill. No other postulated event is affected by the addition of
this instrumentation to the technical specifications.
Thus, adding the proposed requirements to the technical
specifications will not increase the probability or consequences of
any previously evaluated transients or accidents.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
No new failure modes are introduced by the addition of the
reactor high water level feedwater pump trip instrumentation LCO and
surveillance requirements. Modifying the technical specifications to
formally add surveillance requirements already being performed in
accordance with plant procedures will not modify plant response to
any operational or transient event. Increasing the surveillance
interval of the LITS [level indicating transmitter switches] from
annual to once per operating cycle will not significantly affect
reliability. Ensuring the operability of installed instrumentation
does not add new or different kinds of accidents.
Therefore, the new LCO and surveillance requirements do not
create the possibility of a new or different kind of accident.
3. Involve a significant reduction in the margin of safety.
The surveillance requirements being added in this change are
consistent with current surveillances being performed for this
instrumentation, with the exception that the LITS are currently
calibrated on an annual rather than operating cycle basis. These
surveillance and shutdown requirements ensure that protection from
RPV overfill is maintained as assumed in the safety analyses.
Therefore, there is no impact on 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: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: July 7, 1995
Description of amendment request: The proposed change to technical
specification 3/4.7.6 is being made to: 1) increase the allowable
control room air conditioning (CRAC) system in-leakage from 100 cubic
feet per minute (cfm) to 130 cfm; 2) provide a more conservative value
for the maximum differential pressure across the high efficiency
particulate air (HEPA) filters and charcoal adsorbers; 3) clarify that
when the CRAC system is shifted to ``recirculation,'' this will be
performed from the normal mode; and 4) modify the corresponding basis
to reflect the above changes and to note that there are certain
infrequent situations during which the control room emergency
ventilation system (CREVS) will not automatically operate.
Basis for proposed no significant haz- ards consideration
determination: As
[[Page 39444]]
required by 10 CFR 50.91(a), the licensee has provided its analysis of
the issue of no significant hazards consideration (SHC), which is
presented below:
...The proposed changes do not involve an SHC because the
changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The CRAC system in the recirculation mode is used to mitigate
the effects of an accident. Surveillance Requirement 4.7.6.1.e.2 has
been modified to clarify that the system will automatically switch
from the normal mode into a recirculation mode. This change and the
proposed modifications to the acceptance criterion for the
differential pressure across the HEPA filters and charcoal adsorbers
and the increase in the control room in-leakage have no [e]ffect on
the probability of an accident previously evaluated. The
consequences of the accidents that have been previously evaluated
have been reviewed to determine the impact of these proposed
modifications. The increase in the in-leakage will affect the
results of previously generated accident analysis. The accidents
evaluated, namely the Millstone Unit No. 1 MSLB [main steam line
break] and LOCA [loss-of-coolant accident], Millstone Unit No. 2
LOCA, both high and low wind speed case, and Millstone Unit No. 3
LOCA have been reviewed. The Millstone Unit No. 1 LOCA doses to the
Millstone Unit No. 2 control room were qualitatively determined to
be bounded by the Millstone Unit No. 2 LOCA cases. Therefore the
Millstone Unit No. 1 LOCA was not performed. The remaining accidents
were performed. The resultant doses are nearly identical to the
existing doses found in the Millstone Unit No. 2 Final Safety
Analysis Report and are all within the regulatory limits. To perform
these revised control room dose calculations, NNECO used certain new
assumptions which NNECO believes better model the control room and
the effects the accident will have on the control room. The most
significant change with the assumptions is the use of ICRP 30 in
lieu of Regulatory Guide 1.109, Revision 1 for iodine dose
conversion factors. The NRC has used ICRP 30 over the past 5 years
for other applications and its use in this instance is appropriate.
The change in the acceptance criterion for the differential
pressure across the HEPA filter and charcoal adsorbers is a
conservative modification in that the value given is a plant
specific value and will be more indicative of blocked or clogged
filters in actual plant conditions. These proposed changes do not
have any negative impact on the consequences of any accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed modifications to Surveillance Requirement 4.7.6.1
will clarify a portion of a surveillance requirement and will modify
the differential pressure across the HEPA filters and the charcoal
adsorbers. These changes will not create the possibility of a new or
different kind of accident from any previously evaluated. The
increase in the allowable control room in-leakage value from it[s]
current level of 100 cfm to its new value of 130 cfm also does not
create the possibility of a new or different kind of accident. The
CRAC system is used to mitigate the consequences of an accident.
3.Involve a significant reduction in the margin of safety.
The proposed modifications do not decrease the margin of safety
provided. Using the new accident assumptions, the limiting accidents
were re-calculated to determine the impact on the Millstone Unit No.
2 control room. These values are similar to the values found in the
Millstone Unit No. 2 Final Safety Analysis Report and the Millstone
Unit No. 2 Safety Evaluation Report and are within the regulatory
limits established for the control room operators. Since the re-
calculated doses have been shown to be within limits, it has been
concluded that there is 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: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: June 8, 1995
Description of amendment request: The Millstone Unit No. 3
Technical Specification Section 3/4.8.4.3 requires removal of
electrical power to the safety injection accumulator isolation valves
in Modes 1, 2, 3, and 4 in order to protect the containment electrical
penetrations and penetration conductors. Bases Section 3/4.8.4 states
that containment electrical penetrations and penetration conductors are
protected by either deenergizing circuits not required during normal
plant operation (Modes 1 through 4) or by demonstrating the operability
of primary and backup overcurrent protection circuit breakers during
performance of periodic surveillances. It is proposed that Section 3/
4.8.4.3 will be deleted since the containment electrical penetration
and penetration conductors for these circuits are protected by primary
and backup penetration circuit breakers which are demonstrated to be
operable by periodic surveillance 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
(SHC), which is presented below:
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 revised Technical Specification Section 3.5.1 requirements
will provide guidance to ensure that power to the accumulator
isolation valves is removed when the accumulators are required to be
operable and will clarify these requirements.
Removal of the electrical penetration protection requirements of
Section 3/4.8.4.3 is justified since Section 3/4.8.4.1 (Containment
Penetration Conductor Overcurrent Protective Devices) will provide
guidance to ensure that two breakers in series protect the
electrical penetrations and penetration conductors against an
overcurrent condition and the single failure of a circuit breaker.
The two breakers in series also protect the Class 1E buses against a
variety of overcurrent conditions including electrical faults which
may be introduced due to the possible submergence of the accumulator
isolation valves during a LOCA [loss-of-coolant accident].
Therefore, the proposed amendment will 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.
The amended Technical Specification Section 3.5.1 requirements
will provide guidance to ensure that the accumulator isolation
valves are deenergized when the accumulators are required to be
operable. Deletion of the Technical Specifications Section 3.5.1
requires that electrical power to the safety injection accumulator
isolation valves (3SIL*MV8808A, B, C, D) be removed for the
accumulators to be operable. This requirement prevents the
inadvertent closure of these isolation valves which would block the
safety function of the accumulators. Section 4.5.1.c requires
demonstrating accumulator operability by ``At least once per 31 days
when the RCS [reactor coolant system] pressure is above 1000 psig by
verifying that power to the isolation valve operator is disconnected
by removal of the breaker from the circuit.'' The surveillance
requirements for verifying removal of power to the accumulator
isolation valves for Section 4.5.1.c will be changed to ``At least
once per 31 days when the RCS pressure is above 1000 psig by
verifying that the associated circuit breakers are locked in a
deenergized position or removed.''
The proposed change will clarify requirements for securing these
breakers in
[[Page 39445]]
the off (tripped) position in the applicable modes. In addition, index
page xi has been revised to reflect the deletion of Section 3/
4.8.4.3. Attachments 1 and 2 provide the mark-up and retyped pages
of the Millstone Unit No. 3 Technical Specifications, respectively
and reflect the currently issued version of the pages.
Millstone Unit No. 3 Technical Specifications Section 3/4.8.4.3
will not create the possibility of a new or different kind of
accident from any accident previously evaluated since two breakers
in series protect against an overcurrent condition and a single
failure of a circuit breaker. The proposed amendment will not result
in physical plant changes and there are no new credible failure
modes. Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The revised Technical Specification Section 3.5.1 will require
that the accumulator isolation valves have their power deenergized
when the accumulators are required to be operable. This requirement
will maintain accumulator operability by assuring the accumulator
isolation valves remain open.
The removal of the Millstone Unit No. 3 Technical Specification
Section 3/4.8.4.3 is safe since redundant circuit breakers in series
for the accumulator isolation valves will provide assurance that the
electrical penetration and penetration conductors are protected
against overcurrent conditions. This will provide assurance that the
containment boundary is intact.
The proposed amendment will not adversely impact the physical
protective boundaries (fuel matrix/cladding, RCS pressure boundary
and containment) and therefore will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: June 9, 1995
Description of amendment request: The proposed amendment relocates
Surveillance Requirement 4.6.6.1.d.3 for attaining a negative pressure
in the secondary containment to Specification 3.6.6.2, Secondary
Containment. The Action Statement of Section 3.6.6.1 is revised to
decouple Sections 3.6.6.1 and 3.6.6.2. In addition, Definition 1.12,
``Secondary Containment Boundary'' is deleted and included in the Bases
Section 3/4.6.6, Secondary Containment. Bases Section 3/4.6.6.2,
Secondary Containment is expanded using the guidance of the improved
standard technical specifications (STS) for Westinghouse plants (NUREG-
1431).
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration
(SHC), which is presented below:
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 to LCO [limiting condition for operation]
3.6.1.2, LCO 3.6.6.1 and LCO 3.6.6.2 Action Statements, relocation
of Surveillance Requirement 4.6.6.1.d.3 to Specification 3.6.6.2,
changes to Bases Section 3/4.6.6.1, 3/4.6.6.2, and 3/4.6.6.3, and
deletion of Definition 1.12 will resolve the conflict that currently
exists between Specifications 3.6.6.1 and 3.6.6.2. Specifically, the
requirement to establish and maintain a negative pressure in the
secondary containment boundary included in Specification 3.6.6.1
belongs to Specification 3.6.6.2. In the event Secondary Containment
operability is not maintained, the Action Statement for LCO 3.6.6.2
requires that Secondary Containment operability must be restored
within 24 hours. Twenty-four hours is a reasonable completion time
considering the limited leakage design of containment and the low
probability of a DBA [design basis accident] occurring during this
time period. Therefore, it is considered that there exists no loss
of safety function. The proposed changes do not modify the LCO or
surveillance acceptance criterion, nor do they change the frequency
of the surveillances. The proposed changes do not involve any
physical changes to the plant, do not alter the way any structure,
system, or component functions. Therefore, the structures, systems,
or components will perform their intended function when called upon.
The proposed changes do not affect the probability of any previously
evaluated accident. Additionally, the proposed changes are
consistent with the new, improved STS for Westinghouse plants
(NUREG-1431).
Based on the above, the proposed changes do 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.
The proposed changes do not make any physical or operational
changes to existing plant structures, systems, or components. The
proposed changes do not introduce any new failure modes. The
proposed changes simply resolve a conflict which currently exists
between Specifications 3.6.6.1 and 3.6.6.2. Thus, the proposed
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 a margin of safety.
The proposed changes do not have any adverse impact on the
accident analyses. Also, the proposed changes resolve a conflict
which currently exists between Specifications 3.6.6.1 and 3.6.6.2.
The structures, systems, or components covered under Specifications
3.6.6.1 and 3.6.6.2 will performed [sic] their intended safety
function when called upon.
Based on the above, there is no significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: June 20, 1995
Description of amendment request: The proposed amendment relocates
the applicable requirements of Specification 3.6.3 for the main steam
line isolation valves (MSIVs) to Specification 3.7.1.5, ``Main Steam
Line Isolation Valves.'' In addition, the Applicability section of
Specification 3.7.1.5 is revised to indicate that Specification 3.7.1.5
is applicable in Mode 1 and in Modes 2, 3 and 4, except where all MSIVs
are closed and deactivated (i.e., in Modes 2, 3, and 4, Specification
3.7.1.5 is applicable only if the MSIVs are open). Also, the Action
Statement for the Limiting Condition for Operation (LCO) 3.7.1.5 has
been revised using the guidance of the improved standard technical
specifications (STS) for Westinghouse plants (NUREG-1431).
[[Page 39446]]
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration
(SHC), which is presented below:
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 to the Applicability section, Action
Statements, and Surveillance Requirements of Specification 3.7.1.5
and the proposed changes to Specification 3.6.3 preserve the
assumptions in the existing safety analysis. The proposed changes to
the Applicability Section of Specification 3.7.1.5 will require the
MSIVs to be operable in Mode 1 and in Modes 2, 3, and 4, except when
closed and deactivated. The closure of the MSIVs in Modes 2, 3, or 4
is acceptable because when they are closed, they are already
performing their safety function. Since the MSIV closure time has
not been changed, there is no adverse impact on the accidents
previously evaluated.
The proposed changes do not involve any physical changes to the
plant, and do not alter the way any structure, system, or component
functions. Therefore, the proposed changes do not affect the
probability of any previously evaluated accident. Additionally, the
proposed changes are consistent with the new, improved STS for
Westinghouse plants (NUREG-1431).
Based on the above, the proposed changes do 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.
The proposed changes do not make any physical changes to
existing plant structures, systems, or components. When the MSIVs
are closed and deactivated, they are already in the safe position;
therefore, the proposed changes do not introduce a new failure mode.
Additionally, the MSIV closure time (i.e., surveillance acceptance
criterion) is not changed. The purpose of the surveillance is to
ensure that the MSIVs can perform their safety function, and this
requirement is preserved.
Thus, the proposed 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 a margin of safety.
The proposed changes do not revise the closure time of the
MSIVs. This provides assurance that the MSIVs will perform their
design safety function to mitigate the consequences of an accident.
In addition, when they are closed in Modes 2, 3, and 4, they are
already performing their safety function. Therefore, there is no
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: June 26, 1995
Description of amendment request: This proposed amendment would
revise Technical Specification 2.3 to extend the allowed outage time
(AOT) from 24 hours to 7 days for an inoperable low-pressure safety
injection pump. This amendment request is a collaborative effort of
participating Combustion Engineering Owners Group members and is based
on an integrated assessment of plant operations and deterministic and
probabilistic analyses.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The low pressure safety injection (LPSI) system is part of the
emergency core cooling system. Inoperable LPSI components are not
accident initiators in any accident previously evaluated. Therefore,
these changes do not involve an increase in the probability of an
accident previously evaluated.
The LPSI system is primarily designed to mitigate the
consequences of a large loss of coolant accident (LOCA). These
proposed changes do not affect any of the assumptions in the
deterministic LOCA analysis. Hence the consequences of accidents
previously evaluated do not change.
In order to fully evaluate the LPSI allowed outage time (AOT)
extension, probabilistic safety analysis (PSA) methods were
utilized. The results of these analyses show no significant increase
in the core damage frequency. As a result, there would be no
significant increase in the consequences of an accident previously
evaluated. These analyses are detailed in CE NPSD-995, ``Combustion
Engineering Owners Group Joint Applications Report for Low Pressure
Safety Injection System AOT Extension.''
The CEOG report reviewed the risk factors that are impacted by
extending the AOT for a single LPSI pump from 24 hours to seven (7)
days, and demonstrates that the increase in risk is negligible. In
order to perform a more complete assessment of the overall change in
risk, an accounting for avoided risks associated with reducing power
and going to hot or cold shutdown was also considered. This
``transition risk'' is important in understanding the trade-off
between the risk of shutting down the plant compared with restoring
a LPSI pump to operability while at power.
In assessing overall plant risk, the risk avoided based on LPSI
system maintenance while in cold shutdown must also be considered.
Every time the plant is placed in cold shutdown, the LPSI system is
required for decay heat removal when in the shutdown cooling mode of
operation. Maintenance performed on the LPSI system during shutdown
cooling operations may add to the risk of a loss of shutdown cooling
event. Therefore, performing LPSI system maintenance with the unit
on-line, when the LPSI system is not normally in demand, represents
a decrease in shutdown risk.
The CE study concluded that the change in core damage frequency
due to increasing the LPSI AOT from 24 hours to seven (7) days is
insignificant. Additionally, when the reduction in transition and
shutdown risks are considered, it can be shown that there is an
overall reduction in plant risk. Thus, it is the conclusion of the
study that the overall plant impact will either be risk beneficial
or risk neutral.
Therefore, the proposed changes would not increase the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There will be no physical alterations to the plant
configuration, changes to setpoint values, or changes to the
implementation of setpoints or limits as a result of the proposed
changes. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
These proposed changes do not affect the limiting conditions for
operation or their bases used in the deterministic analyses to
establish the margin of safety. PSA evaluations were used to
evaluate this change. These evaluations demonstrate that the changes
are either risk neutral or risk beneficial. These evaluations are
detailed in CE NPSD-995. Therefore, the proposed changes do not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
[[Page 39447]]
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502
NRC Project Director: William H. Bateman
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: June 27, 1995
Description of amendment request: This proposed amendment would
revise Technical Specification 2.2 on the chemical and volume control
system to reformat, clarify the requirements, and be more consistent
with Combustion Engineering Standard Technical Specifications (STS) as
presented in NUREG-0212, Revision 2.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes incorporate required actions, restrictions, and
surveillance requirements for the Chemical and Volume Control System
(CVCS) similar to Combustion Engineering Standard Technical
Specifications (NUREG-0212 Revision 2).
Technical Specification (TS) 2.2(1) specifies the requirements
for borated water sources and flow paths when the reactor is
subcritical and fuel is in the reactor. In order for a flow path to
be operable, a charging or high pressure safety injection pump is
required to be operable to inject the boric acid solution into the
Reactor Coolant System. Currently this specification does not state
any operability requirements for boric acid transfer pumps, charging
pumps or high pressure safety injection pumps. In addition, this
specification does not state any required actions to be taken if the
borated water source or flow path is not operable.
Therefore, the proposed changes incorporate requirements for the
CVCS during shutdown into separate Limiting Conditions for
Operations (LCOs) that will address the requirements for borated
water sources, boric acid flow paths, charging pumps, and boric acid
transfer pumps.
The proposed changes delete operability and surveillance
requirements for level instrumentation on the boric acid storage
tanks. Level instrumentation by itself does not fulfill a safety
function. The proposed changes will still require verification of
tank level.
Additionally, level instrumentation on the boric acid storage
tanks does not meet any of the four criteria for inclusion into
Technical Specifications as presented in the Final Policy Statement
on Technical Specifications Improvements. This instrumentation is
not installed instrumentation used to detect a significant
degradation of the RCS boundary, a design feature or operating
restriction that is an initial condition of a Design Basis Accident,
a component that is part of the primary success path or actuates to
mitigate a DBA, nor is it a component that has been shown to be
significant to public health and safety. Therefore, testing and
maintenance of the level instrumentation will be controlled outside
of the TS.
TS 2.2(3) specifies the Modifications of Minimum Requirements
that are allowed during Power Operation. This specification is
inconsistent with TS 2.2(2) which states the minimum requirements
and is incomplete as it does not address components during Modes 3,
4, and 5. The proposed changes incorporates consistent allowed
outage times for the various components, and additional required
actions for component inoperability during Modes 4 and 5 when fuel
is in the reactor.
The proposed changes incorporate additional operability
requirements for the CVCS and required actions to be taken for CVCS
component inoperability during Modes 4 and 5 when fuel is in the
reactor. The proposed changes delete inconsistencies and clarify
operability requirements for the CVCS whenever the reactor coolant
temperature (Tcold) is greater than or equal to 210 degrees F,
and ensures that operation of the system is consistent with its
design bases. The proposed changes also revise the allowed outage
time for CVCS components from 24 hours to 72 hours based on Standard
Technical Specifications. This change is insignificant based on the
FCS plant specific probabilistic risk assessment. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There will be no physical alterations to the plant
configuration, changes to setpoint values, or changes to the
implementation of setpoints or limits as a result of this proposed
change. No new modes of operation are proposed. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously analyzed.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes incorporate additional operability
requirements, delete inconsistencies, and clarify operability
requirements for the CVCS to ensure that operation of the system is
consistent with its design bases. Therefore, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502
NRC Project Director: William H. Bateman
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: July 11, 1995
Description of amendment request: The proposed amendment would
allow up to 24 hours to restore Safety Injection Tank (SIT) operability
if the SIT is inoperable due to level and/or pressure outside
prescribed limits or if the associated isolation valve is in other than
the full open position. The proposed change would also allow up to 72
hours to restore SIT operability if the SIT is inoperable due to boron
concentration outside prescribed limits.
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 safety injection tanks (SITs) are passive components in the
emergency core cooling system. The SITs are not an accident
initiator in any accident previously evaluated. Therefore, this
change does not involve an increase in the probability of an
accident previously evaluated.
SITs were designed to mitigate the consequences of a loss of
coolant accident (LOCA). These proposed changes do not affect any of
the assumptions used in deterministic LOCA analysis. Hence the
consequences of accidents previously evaluated do not change.
In order to fully evaluate the affect of the SIT allowable
outage time (AOT) extension, probabilistic safety analysis (PSA)
methods were utilized. The results of these analyses show no
significant increase in the core damage frequency. As a result,
there would be no significant increase in the consequences of an
accident previously evaluated. These analyses are detailed in CE
NPSD-994, ``Combustion Engineering Owners Group Joint Applications
Report for Safety Injection Tank AOT/STI Extension.''
The AOT extension based upon boron concentration outside the
prescribed limits
[[Page 39448]]
does not involve a significant increase in the consequences of an
accident as evaluated and approved by the NRC in NUREG-1432,
``Standard Technical Specifications for Combustion Engineering
Plants.'' This proposed change is applicable to FCS.
Therefore, the proposed changes would not increase the
probability or consequences of any accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There will be no physical alterations to the plant
configuration, changes to setpoint values, or changes to the
implementation of setpoints or limits as a result of these proposed
changes. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the limiting conditions for
operation or their bases that are used in the deterministic analyses
to establish the margin of safety. PSA evaluations were used to
evaluate these changes. These evaluations demonstrated that the
changes are either risk neutral or risk beneficial. These
evaluations are detailed in CE NPSD-994. Therefore, the proposed
changes do not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: James R. Curtiss, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502
NRC Project Director: William H. Bateman
Philadelphia Electric Company, Docket No. 50-353, Limerick
Generating Station, Unit 2, Montgomery County, Pennsylvania
Date of amendment request: June 23, 1995
Description of amendment request: This Technical Specifications
(TS) Change Request involves a one-time (i.e., temporary) change
affecting the Allowed Outage Time (AOT) for the Emergency Service Water
(ESW) System; Residual Heat Removal Service Water (RHRSW) System; the
Suppression Pool Cooling, the Suppression Pool Spray, and Low Pressure
Coolant Injection (LPCI) modes of the Residual Heat Removal (RHR)
System; and Core Spray System to be extended from 3 and 7 days to 14
days during the Limerick Generating Station (LGS), Unit 1, sixth
refueling outage scheduled to begin January, 1996. This proposed
extended AOT will allow adequate time to install isolation valves and
cross-ties on the ESW and RHRSW Systems to facilitate future
inspections or maintenance.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. The proposed Technical Specifications changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed one-time TS changes will not increase the
probability of an accident since it will only extend the time period
that the 'A' ESW and RHRSW loops and the affected equipment can be
out-of-service. The extension of the time duration that certain
equipment is out-of-service has no direct physical impact on the
plant. The proposed inoperable systems are normally in a standby
mode while the unit is in OPCON 1 or 2 and are not directly
supporting plant operation. Therefore, they can have no impact on
the plant that would make an accident more likely to occur due to
their inoperability.
During transients or events which require these systems to be
operating, there is sufficient capacity in the operable loops to
support plant operation or shutdown, in-so-much that failures that
are accident initiators will not occur more frequently than
previously postulated.
In addition, the consequences of an accident previously
evaluated in the SAR [Safety Analysis Report] will not be increased.
With the 'A' loops of ESW and RHRSW inoperable, a known quantity of
equipment is either inoperable or the equipment is not fully capable
of fulfilling its design function under all design conditions due to
certain support systems not being operable. Based on the support
functions of the ESW and RHRSW systems, a review of the plant was
performed to determine the impacts that the inoperable ESW and RHRSW
'A' loops would have on other systems. The impacts were identified
for each system, as discussed in the preceding Safety Assessment,
and it was determined whether there were any adverse affects on the
systems. It was then determined how the adverse affects would impact
each system's design basis and overall plant safety. The
consequences of any postulated accidents occurring on Unit 2 during
this AOT extension was found to be bounded by the previous analyses
as described in the SAR.
The existing AOTs limit the amount of time that the plant can
operate with certain equipment inoperable, where single failure
criteria is still met. The minimum equipment required to mitigate
the consequences of an accident and/or safely shutdown the plant
will be operable or the plant will be shutdown. Therefore, by
extending certain AOTs and extending the assumptions concerning the
combinations of events and single failures for the longer duration
of each extended AOT, we conclude, based on the evaluations above,
that at least the minimum equipment required to mitigate the
consequences of an accident and/or safely shutdown the plant will
still be operable during the extended AOT. Therefore, the
consequences of an accident previously evaluated in the SAR will not
be increased.
Therefore, these proposed one-time TS changes will not result in
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed one-time TS changes will not create the possibility
of a different type of accident since it will only extend the time
period that the 'A' ESW and RHRSW loops and the affected equipment
can be out-of-service. The extension of the time duration that
certain equipment is out-of-service has no direct physical impact on
the plant and does not create any new accident initiators. The
systems involved are either accident mitigation systems, safe
shutdown systems or systems that support plant operation. All of the
possible impacts that the inoperable equipment may have on its
supported systems were previously analyzed in the SAR and are the
basis for the present TS ACTION statements and AOTs. The impact of
inoperable support systems for a given time duration was previously
evaluated and any accident initiators created by the inoperable
systems was evaluated. The lengthening of the time duration does not
create any additional accident initiators for the plant.
Therefore, the proposed one-time TS changes will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The ESW and RHRSW systems and their supported systems are
designed with sufficient independence and redundancy such that the
removal from service of a component/subsystem will not prevent the
systems from performing their required safety functions. Since
removal of an ESW and a RHRSW loop from service with one unit in
operation and the other unit in a refueling outage is allowed by the
current Technical Specifications, then the concern is the reduced
margin of safety incurred by extending the affected AOTs.
The present ESW and RHRSW AOT limits were set to ensure that
sufficient safety-related equipment is available for response to all
accident conditions and that sufficient decay heat removal
capability is available for a LOCA/LOOP [Loss-of-Coolant Accident/
Loss-of-Offsite Power] on one unit and simultaneous safe shutdown of
the other unit. A slight reduction in the margin of safety is
incurred during the proposed extended AOT due to the increased risk
that an event could occur in a fourteen day period versus a three or
seven day period. This increased risk is judged to be minimal due to
the low probability of an event occurring
[[Page 39449]]
during the extended AOT and based on the following discussion of
minimum ECCS [Emergency Core Cooling System]/decay heat removal
requirements.
The reduction in the margin of safety is not significant since
the remaining operable ECCS equipment is adequate to mitigate the
consequences of any accident. This conclusion is based on the
information contained in the UFSAR [Updated Final SAR] reference
documents NEDO-24708A and NEDC-30936-A. These documents describe the
minimum requirements to successfully terminate a transient or LOCA
initiating event (with scram), assuming multiple failures with
realistic conditions were used to justify certain TS AOTs per UFSAR
sections 6.3.1.1.2.o and 6.3.3.1. The minimum requirements for short
term response to an accident would be either one LPCI pump or one
Core Spray loop in conjunction with ADS [Automatic Depressurization
System], which would be adequate to re-flood the vessel and maintain
core cooling sufficient to preclude fuel damage. For long term
response, the minimum requirements would be one loop of RHR for
decay heat removal, along with another low pressure ECCS loop. These
minimum requirements will be met since implementation of the
proposed TS changes will require the operability of HPCI [High
Pressure Coolant Injection], ADS, two LPCI subsystems (or one LPCI
subsystem and one RHR subsystem during decay heat removal) and one
Core Spray subsystem be maintained during the 14 day period. A
Special Procedure will be written to ensure the operability of
specified components and that other appropriate compensatory
measures are implemented.
Compensatory measures will be taken prior to or during the
proposed extended AOT for those fire regions that rely on one or
more safe shutdown methods which would all be unable to safely
shutdown the plant with inoperable loops of the ESW and RHRSW
systems or the inoperable systems that ESW or RHRSW support. These
compensatory measures will offset the increased risk of a fire event
occurring in the vulnerable areas, during the fourteen day versus
three day AOT period. Therefore, the proposed extended AOT does not
adversely affect the approved level of fire protection as described
in UFSAR Appendix 9A (Fire Protection Evaluation Report).
A Special Procedure will be written to administratively control
the requirement to maintain the operability of specified components
and implementation of any appropriate compensatory measures which
are deemed necessary during the proposed AOT. In addition,
operations personnel are fully qualified by normal periodic training
to respond to and mitigate a Design Basis Accident, including the
actions needed to ensure decay heat removal while LGS Unit 1 and
Unit 2 are in the operational configurations described within this
submittal. Accordingly, procedures are already in place that cover
safe plant shutdown and decay heat removal for situations applicable
to those in the proposed AOTs.
A Probabilistic Safety Assessment (PSA) Study was performed for
an ESW and RHRSW loop being out-of-service for 14 days on an
operating unit. The Core Damage Frequency (CDF) increased by
3.14x10-6, from 5.11x10-6 /reactor-year to 8.25x10-6/
reactor-year. In absolute terms, this is not a significant increase
in risk. In addition, the modifications to be installed during this
proposed extended AOT will allow for future maintenance and
inspections to be performed on the ESW and RHRSW loops without
removing an entire loop from service, which will reduce risk in the
future. For example, if the ESW loop unavailability, due to testing
or maintenance, is reduced by half, the CDF will decrease by more
than four percent. It will also minimize the potential need for
future AOT extensions on these systems.
Therefore, the implementation of the proposed one-time TS
changes will not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J. W Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: September 29, 1994
Description of amendment request: The proposed Technical
Specification changes represent revisions to Sections 3/4.3.7.2
``Seismic Monitoring Instrumentation'' and 3/4.3.7.3 ``Meteorological
Instrumentation.'' The proposed revisions remove the requirements from
the Technical Specifications and relocates the appropriate descriptive
information and testing requirements to the Hope Creek Updated Final
Safety Analysis Report (UFSAR).
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes involve no hardware changes, no changes to
the operation of any systems or components, and no changes to
existing structures. Neither the relocation of the seismic/
meteorological specifications to the UFSAR nor the elimination of
the Special Report requirements represent changes that affect plant
safety or alter existing accident analyses.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed changes are procedural in nature concerning the
operability and surveillance of instrumentation that are not safety
related and will not impact the operation of any plant safety
related component or equipment. Therefore, these changes will not
create a new or unevaluated accident or operating condition.
3. Will not involve a significant reduction in a margin of
safety.
In accordance with the guidance provided by the NRC regarding
the improvement of Technical Specifications, SECY-93-067, the
proposed changes relocate the seismic and meteorological
instrumentation portions of the Technical Specifications, with the
exception of the Special Report requirements, to the UFSAR. These
instruments are not safety related and do not have any associated
safety margins which could be affected by this change.
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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: November 23, 1994
Description of amendment request: The proposed changes to the
Technical Specifications (TSs) would revise TS 4.8.2.1, ``Electrical
Power Systems D. C. Sources, Surveillance Requirements,'' and
associated Bases Section B 3/4.8.2. The proposed changes would (1)
increase the terminal voltage acceptance criteria for the battery
discharge test from 106 to 108 VDC, (2) delete a ``one time only'' test
that is no longer applicable, (3) delete the battery load profile from
the TS, and (4) revise TS Table 4.8.2.1-1, ``Battery Surveillance
Requirements,'' to agree more closely
[[Page 39450]]
with the BWR4 Standard Technical Specification format.
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:
....will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed changes restore the conservatism to the battery
voltage requirements by raising the minimum acceptable terminal
voltage for the 125 VDC system in order to support proper operation
of the connected loads. This change will cause no change in the
probability of any accident and will, by providing increased support
for connected loads, provide assurance [that] the consequences of
previously evaluated accidents remain within limits. Removal of the
load profile table does not affect the surveillance test loading
which is contained in the station procedures. The (*) footnote
deletion is purely editorial and has no safety bearing. Table
changes agree with the format and wording of the improved BWR4
Standard Technical Specifications.
2....will not create the possibility of a new or different kind
of accident from any previously evaluated.
The revision of the battery sizing calculations did not change
the design base requirement to supply the designed load for a duty
cycle of 4-hours. The proposed change to the minimum acceptable
battery terminal voltage for the 125 VDC system ensures proper
voltages at the battery loads. No other changes to the physical
plant or to the manner in which it is operated are caused by the
proposed amendment; therefore, there is no new or different kind of
accident created by this change.
3....will not involve a significant reduction in a margin of
safety.
The revision of the battery sizing calculations did not change
the design base requirement to supply the designed load for a duty
cycle of 4-hours; however, battery capacity sizing parameter of end
cell voltage was changed to a more conservative value to account for
minimum load voltage requirements. Load profiles for these batteries
were slightly modified to incorporate more precise yet conservative
load current values. These batteries were evaluated using a 25%
additional capacity margin for aging as required by IEEE-450. In
addition, the batteries have a design margin of 5 to 10% for load
growth and/or less than optimum operating condition of the battery;
thereby, maintaining safety margins. Additionally, changes are
comparable to the format and ACTIONS of the improved BWR4 STS.
Permitting 31 days to restore a battery to within CATEGORY A and/or
B limits per the improved BWR4 STS does not involve a reduction in
any margin of safety since the battery, in Category C, remains
operable, as discussed in the BASES.
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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: November 28, 1994
Description of amendment request: The proposed Technical
Specification (TS) revisions provide as follows: (1) The setpoints and
allowable values for the Average Power Range Monitor (APRM) flow-biased
upscale scram/control rod block would be modified to improve operating
margin in the Extended Load Line Limit Analysis (ELLLA) region; (2) The
proposed changes to the Rod Block Monitor (RBM) trip function would
transfer control of the setpoint and allowable value for the RBM -
upscale rod block to the Core Operating Limits Report (COLR); (3) For
the Reactor Coolant System (RCS) recirculation flow upscale trip
function, the proposed changes would revise the trip setpoint and
allowable value to reflect 105% of rated core flow.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
A. Changes to APRM Flow-Biased Scram/Control Rod Block
The proposed changes to the Average Power Range Monitor (APRM)
flow-biased scram/control rod block setpoints and allowable values
were evaluated using NRC approved procedures and methods. The
results of this evaluation are demonstrated in NEDC-31487.
Application of this change in APRM flow-biased scram/control rod
block setpoints and allowable values to Reload 5/Cycle 6 is
confirmed in General Electric Document No. 23A7219.
Analysis presented in NEDC-31487 demonstrate that performance in
the ELLLA region is within design limits for overpressure
protection, stability, loss-of-coolant, containment, reactor
internals, flow-induced vibration, and reactor internal pressure
difference. Impact of ELLLA operation on anticipated transients
without scram is evaluated in Section 7.6.1.7.2 of the UFSAR.
Application of ELLLA region extension to Reload 5/Cycle 6 has been
confirmed in GE Document No. 23A7219.
Because operation with the APRM flow-biased scram/control rod
block setpoints and allowable values is within the bases reviewed
and approved by the NRC in the UFSAR [Updated Final Safety Analysis
Report], this change does not significantly increase the possibility
or consequences of an accident previously evaluated.
B. Transfer of RBM Setpoint Control to the COLR
The proposed changes would transfer control of the setpoint and
allowable value for the rod block monitor (RBM) - Upscale rod block
to the Core Operating Limits Report (COLR). Technical Specification
6.9.1.9, ``Core Operating Limits Report,'' requires that the
analytical methods used to determine core operating limits be those
previously reviewed and approved by the NRC and that the core
operating limits be determined such that all applicable limits of
the safety analysis are met.
The setpoint and allowable value incorporate a controlling value
which will be specified in the COLR and noted as such by reference
in the Technical Specifications. Therefore, the setpoint and
allowable value would continue to be controlled in a manner that
would ensure that safety analysis limits are met and implementation
of the proposed changes would not reduce the level of assurance
provided by the existing Technical Specifications. Based upon the
above information, we conclude that implementation of the proposed
change would not significantly increase the probability or
consequences of an accident previously evaluated.
C. RCS Recirculation Flow Revisions
The original analysis used to support operation up to 105% of
rated core flow is contained in NEDC-31487. NEDC-31487 addresses the
full range of transient and accident events associated with
operation up to 105% of rated core flow. The affects of operation
with the revised RCS recirculation flow upscale trip setpoint and
allowable value are bounded by the analysis presented in NEDC-31487.
In addition, cycle specific analysis performed for Reload 5/
Cycle 6, have incorporated the assumption of operation up to 105% of
rated core flow and have confirmed that operation is within
allowable design limits.
Based on the above information, we conclude that the proposed
change would not significantly increase the probability or
consequences of an accident previously evaluated.
2. Would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
A. Changes to APRM Flow-Biased Scram/Control Rod Block
The proposed changes to the APRM flow-biased scram/control rod
block setpoints and allowable values would not alter the function of
the APRM system nor involve any type of
[[Page 39451]]
plant modification. In addition, operation with the revised APRM flow-
biased scram/control rod block setpoints and allowable values would
not create any new operating modes, accident scenarios, equipment
failure modes, or fission product release paths. Based upon the
above information, we conclude that the proposed changes would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
B. Transfer of RBM Setpoint Control to the COLR
The proposed transfer of control of the RBM setpoint and
allowable value to the COLR would not alter the function of the RBM
system nor involve any type of plant modification. In addition,
operation with the revised setpoint and allowable value would not
create any new operating modes, accident scenarios, equipment
failure modes, or fission product release paths. Based upon the
above information, we conclude that the proposed changes would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
C. RCS Recirculation Flow Revisions
The proposed changes would not alter the function of the RCS
recirculation flow upscale trip function nor involve any type of
plant modification. In addition, operation with the revised RCS
recirculation flow upscale trip setpoint and allowable value would
not create any new operating modes, accident scenarios, equipment
failure modes, or fission product release paths. Based upon the
above information, we conclude that the proposed changes would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Would not involve a significant reduction in a margin of
safety.
A. Changes to APRM Flow-Biased Scram/Control Rod Block
The Bases for Hope Creek Technical Specification 2.2.1 state
that the APRM setpoints were selected to provide adequate margin for
the safety limits while allowing operating margins that reduce the
possibility of unnecessary shutdowns.
The proposed changes would ensure that these objectives are met.
The Minimum Critical Power Ratio (MCPR) operating limit specified in
the Hope Creek COLR was determined using the APRM flow-biased scram/
control rod block setpoints and allowable values proposed in this
amendment application and has been chosen to ensure that the
cladding safety limit would not be violated during normal plant
operations and anticipated transients. Since the operating limit
MCPR is chosen such that the cladding safety limit is maintained,
adequate margins for the safety limits are ensured. The proposed
changes would also serve to ensure that the objective of avoiding
unnecessary shutdowns is met by furnishing greater margin between
the operating envelope and the setpoint at lower flows.
Based on the above information, we conclude that the proposed
changes would not significantly reduce a margin of safety.
B. Transfer of RBM Setpoint Control to the COLR
The proposed transfer of control of the RBM setpoint and
allowable value to the COLR would not affect the methodology for
establishing the core operating limits. The setpoint and allowable
value are modified to incorporate a controlling value which will be
included in the COLR and indicated as such by reference in the
Technical Specifications. Therefore, the setpoint and allowable
value would continue to be controlled in a manner that would ensure
that safety analysis limits are met. We conclude that implementation
of the proposed changes would not significantly reduce a margin of
safety.
C. RCS Recirculation Flow Revisions
The HCGS was licensed to operate up to 105% of rated core flow
as part of Amendment 15. The analysis used to justify operation up
to 105% of rated core flow is contained in NEDC-31487. NEDC-31487
addresses the full range of transient and accident events associated
with operation up to 105% of rated core flow. The affects of
operation with the revised RCS recirculation flow upscale trip
setpoint and allowable value are bounded by the analysis presented
in NEDC-31487.
In addition, cycle specific analysis performed for Reload 5/
Cycle 6, have incorporated the assumptions of operation up to 105%
of rated core flow and have confirmed that operation is within
allowable design limits.
Based on the above information, we conclude that the proposed
changes would not significantly reduce 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: Pennsville Public Library,190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: January 11, 1995
Description of amendment request: The proposed Technical
Specification (TS) revision provides changes to TS Section 3/4.3.8
``Turbine Overspeed Protection System.'' The proposed revision removes
these requirements from the TS and relocates the Bases to the Hope
Creek Updated Final Safety Analysis Report (UFSAR) and the Surveillance
Requirements to the applicable surveillance procedures. The Limiting
Conditions for Operation (LCOs) would be eliminated.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes involve no hardware changes, no changes to
existing structures, and no changes to the operation of any systems
or components. Specifically, the deletion of the LCO's by this
submittal will not alter established turbine overspeed protection
system operation. Procedural guidance will be provided in the event
of an inoperable control, stop, or intermediate valve to place the
system in a safe condition. The relocation of this specification to
the UFSAR and surveillance procedures will continue to ensure that
the probability of unacceptable damage to safety-related structures,
systems, and components from turbine missiles remains acceptably
low. Relocation of this specification's Bases and Surveillance
Requirements to the UFSAR and surveillance procedures, respectively,
and the deletion of the LCO's represents changes that do not affect
plant safety and do not alter existing accident analyses.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed changes are procedural in nature concerning the
location of the descriptive information and surveillance
requirements for the turbine overspeed protection system. Removing
these specifications from the Technical Specifications and placing
them in the UFSAR and surveillance procedures will not alter the
operation of the turbine overspeed protection system or its ability
to perform its intended function. Procedural guidance will be
provided to assist in placing the system in a safe condition while
maintenance and testing of this system will continue in accordance
with the turbine manufacturers recommendations taking into
consideration plant operating experience and ASME guidance.
Therefore, these changes will not create a new or unevaluated
accident or operating condition.
3. Will not involve a significant reduction in a margin of
safety.
The proposed changes relocate the Turbine Overspeed Protection
System portion of the Technical Specifications to the UFSAR and
surveillance procedures in accordance with guidance provided by the
NRC Final Policy Statement regarding the improvement of Technical
Specifications. The requirements that will reside in the UFSAR for
the turbine overspeed protection system will ensure that the system
remains capable of protecting against excessive turbine overspeed.
Therefore, the proposed changes will not involve a significant
reduction in any margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 39452]]
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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: January 20, 1995
Description of amendment request: The proposed Technical
Specification (TS) revision represents changes to TS Section 3/4.11.2.6
``Explosive Gas Mixture,'' TS Table 3.3.7.11-1 ``Radioactive Gaseous
Effluent Monitoring Instrumentation,'' and TS Table 4.3.7.11-1
``Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance
Requirements.'' The proposed revision would remove these TS from the
Technical Specifications and relocate the Bases to the Hope Creek
Updated Final Safety Analysis Report (UFSAR) and the Surveillance
Requirements to the applicable surveillance procedures. The Limiting
Conditions for Operation (LCOs) would be eliminated.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes involve no hardware changes, no changes to
the operation of any systems or components, and no changes to
existing structures. The relocation of this specification to the
UFSAR and surveillance procedures will continue to ensure that the
entrainment of hydrogen in the main condenser is monitored and
controlled. Relocation of this specification's Bases and
Surveillance Requirements to the UFSAR and surveillance procedures,
respectively, and the deletion of the LCO's represent changes that
do not affect plant safety and do not alter existing accident
analyses.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed changes are procedural in nature concerning the
location of the descriptive information and surveillance
requirements for the explosive gas mixture monitoring
instrumentation. Removing these specifications from the Technical
Specifications and placing them in the UFSAR and surveillance
procedures will not alter the operation of the explosive gas
monitors or their ability to perform intended functions. Maintenance
and testing of these monitors will continue based upon the
manufacturers' recommendations taking into consideration plant
operating experience. Therefore, these changes will not create a new
or unevaluated accident or operating condition.
3. Will not involve a significant reduction in a margin of
safety.
The proposed changes relocate the Explosive Gas Mixture
specifications from the Technical Specifications to the UFSAR and
surveillance procedures in accordance with guidance provided by the
NRC Final Policy Statement regarding the improvement of Technical
Specifications. The requirements that will reside in the UFSAR and
surveillance procedures for the explosive gas mixture monitoring
instrumentation will ensure that the ability to determine main
condenser hydrogen concentrations is properly maintained. Therefore,
the proposed changes will not involve a significant reduction in any
margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: January 20, 1995
Description of amendment request: The proposed change to the
Technical Specifications (TS) would revise TS 4.1.3.1.2.b, ``Control
Rods - Surveillance Requirement'' to change the required action to be
taken when a control rod becomes immovable due to excessive friction or
mechanical interference from ``at least once per'' 24 hours to
``within'' 24 hours. The other control rods would be tested within 24
hours and every 7 days thereafter, as opposed to the current
requirement of testing every 24 hours.
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 not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change involves no hardware changes, no changes to
the operation of any systems or components, and no changes to
existing structures. The revision of the control rod movement test
frequency represents a change that does not affect plant safety and
does not alter existing accident analyses.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed change is procedural in nature concerning the
frequency of control rod movement tests for all withdrawn control
rods after a control rod has been determined to be immovable due to
excessive friction or mechanical interference. The methodology for
determining additional immovable control rods remain unchanged. The
proposed change while slightly increasing the possibility of an
undetected immovable control rod will not create a new or
unevaluated accident or operating condition.
3. Will not involve a significant reduction in a margin of
safety.
The proposed change is in accordance with recommendations
provided by the NRC regarding the improvement of Technical
Specifications. This change will result in the perpetuation of
current safety margins while reducing regulatory burden and
decreasing equipment degradation.
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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: John F. Stolz
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: June 1, 1995
Description of amendment request: The proposed change would revise
the Technical Specifications to make them more restrictive regarding
control rod drive (CRD) scram time testing. CRD scram time testing
would be required following maintenance prior to considering the CRD
operable, and could be performed at any reactor
[[Page 39453]]
pressure. Additional testing would be required when reactor coolant
pressure is greater than or equal to 950 psig and prior to 40 percent
rated thermal power.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration which
is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change provides more stringent requirements for
operation of the facility. These more stringent requirements do not
result in operation that will increase the probability of initiating
an analyzed event and do not alter assumptions relative to
mitigation of an accident or transient event. The more restrictive
requirements continue to ensure process variables, structures,
systems and components are maintained consistent with the safety
analysis and licensing basis. Therefore, this change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in the methods governing normal plant operation. The
proposed change does impose different requirements. However, these
changes are consistent with assumptions made in the safety analysis
and licensing basis. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The imposition of more restrictive requirements either has no
impact on or increase in the margin of plant safety. As provided in
the discussion of the change, each change in this category is by
definition providing additional restrictions to enhance plant
safety. The change maintains requirements within safety analyses and
licensing bases. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
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 amendment request: April 10, 1995
Description of amendment request: The proposed amendment would
revise the following Technical Specifications (TS) and their associated
Bases: TS 3/4.7.1.2, ``Auxiliary Feedwater System,'' to clarify Action
``a'' by inserting ``or both'' steam generators'' and to
remove references to pressure indicators and specific pressure readings
and adding performance based requirements; TS 3/4.7.1.3, ``Condensate
Storage Tanks,'' to modify the Limiting Condition for Operation (LCO)
to more closely conform to standard TS; and TS 3/4.7.1.7, ``Motor
Driven Feedwater Pump System,'' to consolidate the requirements of 2
current surveillance requirements and clarify the operability
requirements when local manual valves are realigned for testing
purposes.
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:
Toledo Edison has reviewed the proposed changes and determined
that asignificant hazards consideration does not exist because
operation of the Davis-besse Nuclear Power Station, Unit Number 1,
in accordance with these changes would:
a. Not involve a significant increase in the probability of an
accident previously evaluated because no change is being made to any
accident initiator. No previous analyzed accident scenario is
changed, and initiating conditions and assumptions remain as
previously analyzed. The proposed changes are clarifications and the
incorporations of the guidance provided by NUREG-1430. Therefore, it
can be concluded that the proposed changes do not involve a
significant increase in the probability of an accident previously
evaluated.
b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
affect accident conditions or assumptions used in evaluating the
radiological consequences of an accident. The proposed changes do
not alter the source term, containment isolation or allowable
radiological releases.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not change the way the plant is operated and, no new or
different failure modes have been defined for any plant system or
component important to safety, nor has any limiting single failure
been identified as a result of the proposed changes. No new or
different types of failures or accident initiators are introduced by
the proposed changes.
3. Not involve a significant reduction in a margin of safety
because the proposed changes are clarifications and the
incorporations of the guidance provided by NUREG-1430, and continue
to ensure the availability and capability of the Auxiliary Feedwater
System, Service Water System and the Motor Driven Feedwater Pump
System when called upon to perform their functions. The proposed
changes will not adversely impact any safety analysis assumptions.
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 Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus
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 amendment request: June 1, 1995
Description of amendment request: The proposed amendment would
change the allowed outage time from 72 hours to 7 days for one
unavailable emergency diesel generator (EDG) as detailed in Technical
Specification 3.8.1.1, ``AC Power Sources, Operating,'' and its
associated Bases 3.0.5.
Basis for proposed no significant hazards consideration
determination:As required by 10 CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:
Toledo Edison has reviewed the proposed change and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station (DBNPS), Unit No.
1, in accordance with this change would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the proposed change to
increase the allowed outage time for one emergency diesel generator
from three (3) days to seven (7) days does not make a change to any
accident initiator, initiating condition or assumption. The accident
previously evaluated in the DBNPS Updated Safety Analysis Report
(USAR) Section 15.2.9, Loss of All AC Power to the Station
[[Page 39454]]
Auxiliaries (Station Blackout), is not affected by this proposed
change. The proposed change does not involve a significant change to
the plant design or operation, only to the allowed outage time, and
based on a review of the available alternate A.C. power sources, the
effect on probabilistic risk at power, the effect on shutdown risk,
and maintenance planning and scheduling, this change has been
determined to be acceptable.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed change does not
invalidate assumptions used in evaluating the radiological
consequences of an accident, does not alter the source term or
containment isolation and does not provide a new radiation release
path or alter potential radiological releases.
2. Not create the possibility of a new or different kind of
accident from any previously evaluated because the proposed change
does not introduce a new or different accident initiator or
introduce a new or different equipment failure mode or mechanism.
3. Not involve a significant reduction in the margin of safety
because the proposed change does not significantly reduce the margin
to safety which exists in the present Technical Specification action
statements. The DBNPS USAR Section 15.2.9 evaluates the
acceptability of the loss of all A.C. power to the station,
including the loss of both EDGs, and the margin of safety in this
analysis is not affected by the proposed change. in addition, since
the issuance of the original DBNPS Operating License Technical
Specifications Toledo Edison has installed a Station Blackout Diesel
Generator (SBODG), comparable in continuous rating to the EDGs and
capable of providing emergency A.C. power in the event all three
offsite 345 kV transmission lines and the two EDGs are unavailable.
This has positive effect on maintaining the margin to safety which
exists in the Technical Specifications with a three day allowed
outage time, which was established prior to installation of the
SBODG.
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 Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus
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 amendment request: June 7, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.9.4, Refueling Operations -
Containment Penetrations, and associated Bases 3/4.9.4, Containment
Penetrations. The proposed changes include revising the Limiting
Condition for Operation (LCO) 3.9.4.b to allow both doors of the
containment personnel airlock to be open during core alterations or
movement of irradiated fuel within the containment, provided that
certain specified conditions are meet. Additional changes are proposed
to revise or clarify TS LCO 3.9.4.c, TS Action 3.9.4.a, and TS
Surveillance Requirement 4.9.4, and modify the Bases to reflect the
requested 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:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Plant (DBNPS), Unit No. 1, in
accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no Updated Safety Analysis
Report (USAR) accident initiators are affected by the proposed
changes.
The proposed change to TS LCO 3.9.4.b would allow both doors of
the containment personnel air lock to be open during core
alterations or movement of irradiated fuel within the containment,
provided that certain specified conditions are met. The containment
personnel air lock is not an initiator to any accident. Whether the
containment personnel air lock doors are open or closed during fuel
movement and core alterations has no effect on the probability of
any accident previously evaluated.
The proposed clarification of TS LCO 3.9.4.c, changing the term
``outside atmosphere'' to ``atmosphere outside containment,'' and
the proposed change to TS LCO 3.9.4.c.1, confirming that, in
addition to a manual or automatic isolation valve, or a blind
flange, equivalent means may be used to close a containment
penetration, have no bearing on the probability of an accident
previously evaluated.
The proposed changes to TS Action 3.9.4.a, TS Surveillance
Requirement (SR) 4.9.4, and TS Bases 3/4.9.4 are administrative
changes and have no bearing on the probability of an accident
previously evaluated.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
invalidate accident conditions or assumptions used in evaluating the
radiological consequences of any accident.
he analysis results for a fuel handling accident inside
containment, as presented in Section 15.4.7.3 of the DBNPS USAR, are
well within the 10 CFR 100 guideline values. Since the analysis does
not take credit for containment isolation, the status of the
personnel air lock has no impact on the acceptability of the
results. In the event of a fuel handling accident, release of
radioactive material will continue to be minimized since the air
lock door will remain capable of being closed. Further, the proposed
change could significantly reduce the dose to workers in the
containment in the event of a fuel handling accident by speeding the
containment evacuation process.
Since an engineering evaluation described in proposed Bases 3/
4.9.4 will ensure that a particular containment penetration closure
technique is capable of restricting the release of radioactive
material from a fuel handling accident, the proposed change to TS
LCO 3.9.4.c.1, confirming that an equivalent means may be used to
close a containment penetration, has no adverse effect on the
consequences of an accident previously evaluated.
The proposed clarification of TS LCO 3.9.4.c, and the proposed
changes to TS Action 3.9.4.a, TS SR 4.9.4, and TS Bases 3/4,9.4 are
administrative changes and have no effect on the consequences of an
accident previously evaluated.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because there are no
new failure modes or mechanisms associated with the proposed
changes, nor do the proposed changes involve any modification of
plant equipment or changes in plant operational limits.
As described above, the analysis results for a fuel handling
accident inside containment does not take credit for containment
isolation. Thus the proposed change to TS LCO 3.9.4.b to allow both
doors of the containment personnel air lock to be open during core
alterations or movement of irradiated fuel within the containment
could affect the release path for radioactive material released
during a fuel handling accident, however no new or different kind of
accident will result.
3. Not involve a significant reduction in the margin of safety.
The analysis results for a fuel handling accident inside
containment, as presented in [Section 15.4.7.3 of] the DBNPS USAR,
are well within the 10 CFR 100 guideline values. Since the analysis
does not take credit for containment isolation, the status of the
personnel air lock has no impact on the acceptability of the
results.
The proposed change to TS LCO 3.9.4.c.1 regarding the use of
equivalent means of containment penetration closure has no adverse
impact on the margin of safety since an equivalent containment
penetration
[[Page 39455]]
closure technique will provide the same assurance of containment
closure during core alterations or movement of irradiated fuel
inside containment.
The various administrative changes and clarifications proposed
will not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus
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 amendment request: June 23, 1995
Description of amendment request: The proposed amendment would
relocate Technical Specifications (TS) 3/4.3.3.3 - Seismic
Instrumentation, TS 3/4.3.3.4 - Meteorological Instrumentation, and TS
3/4.4.11 - Reactor Coolant System Vents and associated Bases.
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:
Toledo Edison has reviewed the proposed changes and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station, Unit Number 1,
in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no change is being made to any
accident initiator. No previous analyzed accident scenario is
changed, and initiating conditions and assumptions remain as
previously analyzed.
The proposed changes are deletions and relocations of
specifications that do not meet the NRC Final Policy Statement [58
FR 39132, dated July 22, 1993] criteria for inclusion in TS.
Furthermore, these relocations and deletions are consistent with the
NRC guidance for TS provided by the ``Improved Standard Technical
Specifications for Babcock and Wilcox Plants,'' NUREG-1430, Revision
0. Therefore, it can be concluded that the proposed changes do not
involve a significant increase in the probability of an accident
previously evaluated.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
affect accident conditions or assumptions used in evaluating the
radiological consequences of an accident. The proposed changes do
not alter the source term, containment isolation or allowable
radiological releases.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not change the way the plant is operated, and no new or
different failure modes have been defined for any plant system or
component important to safety, nor has any limiting single failure
been identified as a result of the proposed changes. No new or
different types of failures or accident initiators are introduced by
the proposed changes.
3. Not involve a significant reduction in a margin of safety
because Seismic Instrumentation, Meteorological Instrumentation, and
Reactor Coolant System Vents are not inputs in the calculation of
any safety margin with regard to TS Safety Limits, Limiting Safety
System Settings, other TS Limiting Conditions for Operation, or
other previously defined margins for any structure, system, or
component important to 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 Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Gail H. Marcus
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: July 14, 1995
Description of amendment request: The proposed Technical
Specifications (TS) changes would provide a two-hour allowed outage
time (AOT) for one residual heat removal (RHR) pump to accommodate
plant safety and emergency power systems surveillance testing and
permit depressurizing safety injection (SI) accumulators in lieu of
accumulator isolation.
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:
Specifically, operation of the Surry Power Station in accordance
with the proposed change will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Surveillance and testing requirements are necessary to assure
that RHR and interfacing systems' reliability is maintained.
Existing analyses demonstrate that adequate shutdown cooling will be
maintained with one train of RHR Operable and in service. Analyses
also demonstrate that alternate shutdown cooling modes remain
available with adequate decay heat removal capability. Furthermore,
the opposite train of RHR remains available while in the two hour
surveillance AOT. The response time and operator actions required to
place the available RHR train in service are consistent with similar
operator response times and actions
required to place alternate shutdown cooling modes in service.
The administrative controls and procedures in place assure adequate
shutdown cooling capability is maintained as supported by existing
analyses.
The existing safety analyses demonstrate that Reactor Coolant
System [RCS] integrity will be maintained when SI accumulator
pressure is below the pressurizer PORV [power operated relief valve]
LTOPS [low temperature overpressure system] setpoint. Therefore, SI
accumulator isolation is not required to ensure Reactor Coolant
System integrity. With RCS temperature below the LTOPS enabling
temperature, automatic actuation of the pressurizer PORVs or other
TS specified relief paths ensure the assumed design basis reactor
vessel beltline flaw will not propagate under design basis low
temperature overpressurization accident conditions. System design
and configuration adequately mitigate an LTOPS actuation due to an
SI accumulator discharge with no negative consequences regarding RCS
structural integrity or SBLOCA [small break loss-of-coolant
accidents] concerns.
Therefore, the proposed Allowed Outage Time for an inoperable
RHR loop and the ability to depressurize the SI accumulator in lieu
of SI accumulator isolation do not increase the probability or
consequence of any previously analyzed accidents.
(2) Create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed two hour AOT for one train of the RHR System will
preclude the possibility of a Technical Specification violation for
conditions where a train of RHR is out of service for surveillance
testing. Calculations by Westinghouse with evaluations and
supporting analyses performed by Virginia Power, confirm the
adequacy of decay heat removal with one RHR train in service, and
multiple alternate shutdown cooling modes remain available. There
are no plant modifications required by this proposed TS change.
Further, the proposed change does not invalidate any
[[Page 39456]]
component design criteria or the assumptions of the UFSAR [updated
final safety analysis report] accident analyses. The RHR System is
being operated in a manner consistent with the design basis and
configuration of the system and is supported by existing analyses
and procedural controls.
There are no new failure modes or mechanisms associated with the
proposed change to allow the depressurizing of a SI accumulator to a
pressure value below the LTOPS setpoint. The LTOPS enabling
temperature remains unchanged. No operating limits or setpoints are
added or deleted by the proposed change. Reactor Coolant System
pressure relief paths are not affected.
Therefore, the possibility of a new or different kind of
accident is not being created by the proposed Allowed Outage Time
for an inoperable RHR loop and the ability to depressurize the SI
accumulator in lieu of SI accumulator isolation.
(3) Involve a significant reduction in margin of safety.
The proposed Technical Specifications change does not involve a
reduction in a margin of safety. The existing safety analyses
demonstrate that adequate shutdown cooling will be maintained when a
train of RHR is out of service for up to two hours for plant system
surveillance testing, while the operable train of RHR is operating.
Supporting analyses determined that the RHR System meets the design
cooldown requirements for a reactor core rating of 2546 MWth
[megawatt thermal] with either one or both trains of RHR in service.
Additionally, an evaluation of the technical basis for shutdown
operations for the proposed Surry core uprating to 2546 MWth
determined that the administrative controls and Abnormal Procedures
in place at Surry ensure adequate decay heat removal capability
during shutdown conditions. The administrative controls and
procedure revisions are supported by a detailed series of thermal-
hydraulic calculations for various loss of RHR scenarios. There is
no reduction in shutdown cooling capability due to the proposed TS
change, and no reduction in the capability to mitigate a loss of
decay heat removal event since the RHR train affected by the testing
is available and can be restored in a comparable time period to that
required to restore RHR to service in the event of loss of station
power or loss of the operating train of RHR. Consequently, system
design, plant configuration, and administrative controls remain
available to adequately mitigate a loss of RHR event with a single
train of RHR out of service for up to two hours during plant system
surveillance testing. It may be concluded that there is no reduction
in the margin of safety due to the proposed Technical Specification
change.
Existing safety analyses also demonstrate that Reactor Coolant
system integrity will be maintained in the event of an inadvertent
SI accumulator discharge when SI accumulator pressure is below the
pressurizer PORV LTOPS setpoint. Sufficient administrative controls
are maintained to ensure LTOPS is ``Enabled'' and SI accumulators
are isolated at the appropriate RCS conditions to minimize the
possibility of challenging RCS integrity. Technical Specifications
administrative controls that prevent inadvertent charging pump
operation, maintain adequate relief paths, and restrict Steam
Generator primary to secondary temperature differential remain in
place. Consequently, the Technical Specifications change ensures
that an inadvertent SI accumulator discharge cannot challenge RCS
structural integrity during LTOPS conditions when SI accumulator
pressure is below the pressurizer PORV LTOPS setpoint.
Therefore, the proposed Allowed Outage Time for an inoperable
RHR loop and the ability to depressurize the SI accumulator in lieu
of SI accumulator isolation does not reduce any margin of safety as
defined in the Technical Specifications.
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: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: David B. Matthews
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: June 14, 1995, as supplemented by letter
dated July 13, 1995.
Description of amendment request: This amendment request proposes
to revise Technical Specification (TS) 3.2.3, ``Nuclear Enthalpy Rise
Hot Channel Factor,'' TS 6.9.1.9, ``Core Operating Limits Report,'' and
the associated Bases sections. The revisions are needed to incorporate
changes associated with the planned implementation of advanced nuclear
and core thermal-hydraulic design methodologies licensed from
Westinghouse Electric Corporation for core reload design, starting with
Cycle 9.
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 probability of occurrence and the consequences of an
accident evaluated previously in the Updated Safety Analysis Report
(USAR) are not increased due to the proposed technical specification
changes. The Technical Specification changes being requested are to
reflect revised calculational methods to be used for core reload
design, starting with Cycle 9. There are no changes being made to
any licensed design parameters from previous cycles. Thus, it is
concluded that the probability and consequences of the accidents
previously evaluated in the USAR are not increased.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There is no new type of accident or malfunction being created.
The proposed changes only provide revised analysis methodologies to
support core reload design, starting with Cycle 9. The requested
changes do not change the method and manner of plant operation. The
safety design bases in the USAR have not been altered. Thus, the
requested changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not change the plant configuration in a
way that introduces a new potential hazard to the plant and do not
involve a significant reduction in the margin of safety. The
analyses and evaluations discussed in the safety evaluation
(Attachment I) [Attached to Wolf Creek Nuclear Operating
Corporation's letter number ET 95-0051, dated June 14, 1995]
demonstrates that all applicable design criteria continue to be met
for the changes. Therefore, it is concluded that the margin of
safety, as described in the bases to any technical specification, is
not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037
NRC Project Director: William H. Bateman
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the
[[Page 39457]]
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.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of amendment request: February 23, 1995
Description of amendment request: The amendment relates to
Commonwealth Edison Company's (ComEd) request to reflect the merger
between IIGEC, MidAmerican, Midwest Power Systems Inc., and Midwest
Resources, Inc. By letter dated November 21, 1994, Iowa-Illinois Gas
and Electric Company (IIGEC) requested approval, pursuant to Section
50.80 of Title 10 of the Code of Federal Regulations, of the transfer
of its ownership share of 25 percent of Quad Cities Nuclear Power
Station, Units 1 and 2, to MidAmerican Energy Company (MidAmerican).
Date of publication of individual notice in Federal Register: July
5, 1995 (60 FR 35054)
Expiration date of individual notice: August 4, 1995
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam
Neck Plant, Middlesex County, Connecticut
Date of amendment request: March 31, 1995
Description of amendment request: The proposed amendment will
delete Technical Specification (TS) Sections 1.38 and 1.39,
``Definitions, Fuel Assembly Types,'' revise TS Sections 3/4.9.3,
``Refueling Operations, Decay Time'' and TS 3/4.9.14, ``Refueling
Operations, Spent Fuel Pool - Reactivity Condition,'' replace TS
Sections 5.6.1.1, ``Spent Fuel,'' and TS 5.6.3, ``Capacity,'' and add a
new TS Section 3/4.9.15, ``Refueling Operations, Spent Fuel Pool
Cooling.'' These changes would support a rerack of the spent fuel pool
to expand the spent fuel pool's storage capacity from 1168 assemblies
to 1480 assemblies so as to accommodate a full-core-discharge through
the current validity date of the Haddam Neck operating license (2007).
Date of publication of individual notice in Federal Register: May
12, 1995 (60 FR 25746)
Expiration date of individual notice: June 12, 1995
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, CT 06457.
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station Units 1 and 2, Lake County, Illinois
Date of application for amendments: June 14, 1995
Brief description of amendments: The amendments revise the
requirement to perform an emergency diesel generator (EDG) automatic
start and sequence loading test immediately following the 24 hour EDG
endurance test.
Date of issuance: July 18, 1995
Effective date: July 18, 1995
Amendment Nos.: 166 and 154
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.Public comments requested as to
proposed no significant hazards consideration determination: Yes (60 FR
34308). This notice provided an opportunity to submit comments on the
Commission's proposed no significant hazards consideration
determination. No comments have been received. The notice also provided
for an opportunity to request a hearing by July 31, 1995, but indicated
that if the Commission makes a final no significant hazards
consideration determination any such hearing would take place after
issuance of the amendment. The Commission's related evaluation of the
amendments, finding of exigent circumstances and final no significant
hazards consideration determination is contained in a Safety Evaluation
dated July 18, 1995.
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: March 31, 1995
Brief description of amendments: The amendments revise Technical
Specification section 3.9.4 to allow, under certain conditions, both
containment personnel airlocks to be open during core alterations.
Date of issuance: July 12, 1995
Effective date: July 12, 1995
Amendment Nos.: 197 and 182
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29879)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 12, 1995.No significant
hazards consideration comments received: No.
[[Page 39458]]
Local Public Document Room location: Maud Preston Palenske
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station,Nemaha County, Nebraska
Date of amendment request: May 2, 1995
Brief description of amendment: The amendment revised Surveillance
Requirement 4.7.A.2.f.1 to allow a one-time extension for the
performance of Type B local leak rate testing of the drywell head and
manport from July 17, 1995, until startup from Refueling Outage 16,
scheduled to commence on October 13, 1995.
Date of issuance: July 11, 1995
Effective date: July 11, 1995
Amendment No.: 170
Facility Operating License No. DPR-46. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29879)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 11, 1995No significant hazards
consideration comments received: No.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, NE 68305.
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
MillstoneNuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: January 10, 1995
Brief description of amendment: The amendment revises the Technical
Specifications to delete the power range negative flux trip from Tables
2.2-1, 3.3-1, and 4.3-1, and delete the associated Bases Section 2.0.
Date of issuance: July 11, 1995
Effective date: As of the date of issuance to be implemented
within30 days.
Amendment No.: 116
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 1, 1995 (60 FR
11135)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 11, 1995. No significant hazards
consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
MillstoneNuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: March 29, 1995
Brief description of amendment: The amendment revises Technical
Specification 3.10.5 to allow more than one control bank to be fully
withdrawn from the core simultaneously in order to conduct rod drop
time response testing.
Date of issuance: July 11, 1995
Effective date: As of the date of issuance to be implemented
within60 days.
Amendment No.: 117
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29880) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 11, 1995. No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Northern States Power Company, Docket No. 50-263, Monticello
NuclearGenerating Plant, Wright County, Minnesota
Date of application for amendment: February 12, 1993, as
supplemented by letters dated March 22, 1993, and August 25, 1994
Brief description of amendment: The amendment increases the minimum
core spray pump flow to more conservatively account for emergency core
cooling systems bypass leakage paths. The amendment also makes various
typographical, editorial and administrative corrections and changes.
Date of issuance: July 12, 1995
Effective date: July 12, 1995
Amendment No.: 93
Facility Operating License No. DPR-22. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41508). The August 25, 1994 letter provided clarifying information
within the scope of the original submittal and did not change the
staff's initial proposed no significant hazards considerations
determination.The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 12, 1995.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.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: November 11, 1994, as supplemented by
letters dated April 7, 1995, and June 26, 1995
Brief description of amendment: The amendment implements
administrative changes to TS 5.2 and 5.5. These changes reflect
organizational changes in OPPD senior management, delete specific
titles of personnel on the Plant Review Committee (PRC), revise the
makeup of the PRC quorum, revise the membership of the Senior Audit and
Review Committee (SARC), delete SARC audit frequencies and add minor
clarifications to the descriptions of SARC reviews and audits.
Date of issuance: July 21, 1995
Effective date: July 21, 1995
Amendment No.: 168
Facility Operating License No. DPR-40. Amendment revised the
TechnicalSpecifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65819). The April 7, 1995, and June 26, 1995, letters provided
clarifying information and did not change the initial no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
July 21, 1995.No significant hazards consideration comments received:
No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: March 1, 1995
Brief description of amendment: The amendment revises TS 2.5, 2.8,
2.11, 3.2, and 3.10 and relocates administrative controls for the
emergency and security plans from TS 5.5 and 5.8 to the plans. The
relocation is in accordance with Generic Letter (GL) 93-07,
``Modification of the Technical Specification Administrative Control
Requirements for Emergency and Security Plans.''
Date of issuance: July 21, 1995
Effective date: July 21, 1995
Amendment No.: 169
Facility Operating License No. DPR-40: The amendment revised the
TechnicalSpecifications.
[[Page 39459]]
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18627) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 21, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay
Power Plant, Unit 3, Humboldt County, California
Date of application for amendment: April 10, 1995
Brief description of amendment: This amendment revised License No.
DPR-7, to permit the provisions of 10 CFR 50.59 to be applied with
respect to changes to the facility or procedures described in the
Decommissioning Plan or changes to the Decommissioning Plan, and the
conduct of tests or experiments not described in the Decommissioning
Plan.
Date of issuance: July 7, 1995
Effective date: This license amendment is effective as of the date
of its issuance and must be fully implemented no later than 30 days
from the date of issuance.
Amendment No.: 29Facility License No. DPR-7: This amendment revised
License No. DPR-7
Date of initial notice in Federal Register: June 6, 1995 (60 FR
29885)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 7, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Humboldt County Library, 636 F
Street, Eureka, California 95501.
PECO Energy Company, Public Service Electric and Gas Company,
Delmarva Power and Light Company, and Atlantic City Electric
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit
No. 3, York County, Pennsylvania
Date of application for amendment: November 21, 1994
Brief description of amendment: This amendment changes the
technical specifications (TS) by allowing the third Type A Containment
Integrated Leakage Rate Test in the second 10-year service period to be
conducted during refueling outage 11 scheduled for September 1997. This
TS change is consistent with a one-time exemption from Appendix J to 10
CFR Part 50 that extends the 10-year service period and allows the
three type A tests to be performed at intervals that are not
approximately equal.
Date of issuance: July 10, 1995Effective date: July 10, 1995
Amendment No.: 210
Facility Operating License No. DPR-56: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 23, 1995 (60 FR
27340)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 10, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
PECO Energy Company, Public Service Electric and Gas
Company,Delmarva Power and Light Company, and Atlantic City
Electric Company,Docket No. 50-278, Peach Bottom Atomic Power
Station,Unit No. 3, York County, Pennsylvania
Date of application for amendment: June 23, 1993, as supplemented
by letters dated April 5, May 2, June 6, June 8, July 6 (two letters),
July 7, July 20, July 28 (two letters), September 16, September 30, and
October 14, 1994 and June 22, 1995.
Brief description of amendment: The amendment raises the authorized
maximum power level from 3293 MWt to a new limit of 3458 MWt. The
amendment also approves changes to the Technical Specifications to
implement operation at the increased power limit.
Date of issuance: July 18, 1995
Effective date: As of date of issuance and is to be implemented
prior to startup in Cycle 11, currently scheduled for October 1995.
Amendment No.: 211
Facility Operating License No. DPR-56: Amendment revised the
License and the Technical Specifications.
Date of initial notice in Federal Register: August 29, 1994 (59 FR
44432)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 18, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: April 10, 1995
Brief description of amendments: Remove the response time limit
Tables 3.3.1-2, 3.3.2-3, and 3.3.3-3 from the Technical Specifications,
and add the information to the Final Safety Analysis Report in
accordance with Generic Letter 93-08.
Date of issuance: July 11, 1995
Effective date: July 11, 1995
Amendment Nos.: 148 and 118
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 6, 1995 (60FR
29887)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 11, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
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 21, 1994, as
supplemented April 6, and July 3, 1995
Brief description of amendments: These amendments make changes
affecting the Administrative Controls Section of the Technical
Specifications. The areas changed are Nuclear Effectiveness and
Efficiency Design Study (NEEDS) Organization Title Changes; Minimum
Shift Crew Composition; delete Independent Technical Review Section
from TS; delete Nuclear Review Board (NRB) Review Section from TS; and
delete NRB Audit Section from TS.
Date of issuance: July 18, 1995
Effective date: Units 1 and 2, as of the date of issuance and shall
be implemented within 30 days.
Amendment Nos.: 96 and 60
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 10, 1995 (60 FR
24914)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 18, 1995.No significant
hazards consideration comments received: No
[[Page 39460]]
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: August 31, 1994, as
supplemented July 3, 1995
Brief description of amendments: These amendments modify TS
Sections 3.4.9.1, 3.4.9.2, 3.9.11.1, 3.9.11.2, and the associated Bases
Sections 3/4.4.9 and 3/4.4.11, to permit the use of either an
``analytical approach'' (i.e., calculation) or ``demonstrations'' to
ensure the operability of an alternate decay heat removal method,
rather than the existing TS requirement which stipulates that
operability of the alternate decay removal method be demonstrated.
Date of issuance: July 18, 1995
Effective date: Units 1 and 2, as of the date of issuance and shall
be implemented within 30 days.
Amendment Nos.: 97 and 61
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55884)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 18, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: August 22, 1994, as
supplemented July 3, 1995
Brief description of amendments: These amendments revise Technical
Specification Surveillance Requirement 4.1.3.1.4a to delete the
requirement that the Scram Discharge Volume (SDV) be determined
operable by testing the SDV vent and drain valves from a configuration
of less than or equal to 50% rod density.
Date of issuance: July 18, 1995
Effective date: Units 1 and 2, effective as of date of issuance and
shall be implemented within 30 days.
Amendment Nos.: 98 and 62
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55881)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 18, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: August 22, 1994, as
supplemented by letter dated July 3, 1995
Brief description of amendments: The amendments revise the
Technical Specifications surveillance requirements for scram insertion
times and revise the TS surveillance requirements for control rod block
and source range monitoring instrumentation.
Date of issuance: July 18, 1995
Effective date: Units 1 and 2, effective as of the date of
issuance and shall be implemented within 30 days.
Amendment Nos.: 99 and 63
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55881)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 18, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Public Service Company of Colorado, Docket No. 50-267, Fort St.
Vrain Nuclear Generating Station (FSV), Unit No. 1, Platteville,
Colorado
Date of application for amendment: Amendment No. 88, April 14,
1995.
Brief description of amendment: This amendment would revise the FSV
Decommissioning Technical Specifications (DTS) by: revising the FSV DTS
to reflect recent organizational changes resulting from corporate
restructuring to prepare for repowering the site with natural gas-power
turbines and to incorporate editorial changes. The staff has determined
that the proposed amendment does not require a significant hazard
consideration, pursuant to 10 CFR 50.92.Possession-Only License No.
DPR-34: Amendment revises the DTS.
Local Public Document Room location: Weld Library District -
Downtown Branch, 919 7th Street, Greeley, CO 80631.
Sacramento Municipal Utility District, Docket No. 312, Rancho Seco
Nuclear Generating Station, Sacramento County, California
Date of application for amendment: February 28, 1995
Brief description of amendment: This amendment relocates the
quality assurance audit frequencies from the technical specifications
to the Rancho Seco Quality Manual and changes the reporting frequency
of the Radioactive Effluent Release Report from semi-annual to annual.
Date of issuance: July 19, 1995
Effective date: July 19, 1995
Amendment No.: 122
Facility Operating License No. NPF-1: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 1995 (60 FR
16200)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 19, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Central Library, Government
Documents, 828 I Street, Sacramento, California 95814.
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: September 15, 1993, as
supplemented by letter dated September 6, 1994.
Brief description of amendments: The amendments revised Technical
Specification (TS) Table 2.2-1, ``Reactor Protective Instrumentation
Trip Setpoint Limits,'' Table 3.3-1, ``Reactor Protective
Instrumentation,'' Table 3.3-3, ``Engineered Safety Feature Actuation
System Instrumentation,'' and Table 3.3-4, ``Engineered Safety Feature
Actuation System Instrumentation Trip Values,'' and the associated
Bases. The revisions to the notes in these tables change the pressure
at which the low pressurizer pressure trip bypass shall be
automatically removed to a consistent value of ``before pressurizer
pressure exceeds 500 psia (the corresponding bistable allowable value
is less than or equal to 472 psia).'' In addition, the wording of the
notes is revised to make the notes more consistent with each other.
Date of issuance: July 14, 1995
[[Page 39461]]
Effective date: July 14, 1995, to be implemented within 30 days of
the date of issuance.
Amendment Nos.: Unit 2 - Amendment No. 120; Unit 3 - Amendment No.
109
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 29, 1993 (58
FR 50975). The September 6, 1994, supplemental letter provided
additional clarifying information and did not change the initial no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated July 14, 1995. No significant hazards consideration
comments received: No
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557,Irvine, California 92713.
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: September 3, 1992
Brief description of amendments: These amendments revise TS 3/4.4.8
``Pressure/Temperature Limits - Reactor Coolant System,'' and their
associated Bases, following NRC guidance provided in Generic Letter 91-
01, ``Removal of the Schedule for Withdrawal of Reactor Vessel Material
Specimens from Technical Specifications.'' This generic letter allows
licensees to remove the reactor vessel material surveillance capsule
withdrawal schedules from the TS because they are a duplication of the
requirements of 10 CFR Part 50 Appendix H.
Date of issuance: July 17, 1995
Effective date: July 17, 1995, to be implemented within 30 days of
issuance
Amendment Nos.: Unit 2 - Amendment No. 121; Unit 3 - Amendment No.
110
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 17, 1993 (58
FR 8781)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 17, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: April 30, 1993, as supplemented
by letters dated July 6, 1994 (separate letters for each unit), and
letter dated January 27, 1995.
Brief description of amendments: These amendments revise TS 3/
4.4.8.1, ``Pressure-Temperature Limits,'' TS 3.4.8.3.1, ``Overpressure
Protection Systems-RCS Temperature less than or equal to deg.F [for
Unit 2, less than or equal to 246 deg.F for Unit 3],'' TS 3.4.8.3.2,
``Overpressure Protection Systems-RCS Temperature 256 deg.F
[for Unit 2, 246 deg.F for Unit 3],'' and the associated TS
Bases. The proposed change (1) revises the reactor coolant system (RCS)
pressure-temperature (P-T) limits and the low temperature overpressure
protection (LTOP) enable temperatures to be effective until 20
effective full power years (EFPY) of operation and (2) makes minor
editorial changes.
Date of issuance: July 18, 1995
Effective date: July 18, 1995, to be implemented within 30 days of
issuance
Amendment Nos.: Unit 2 - Amendment No. 122; Unit 3 - Amendment No.
111
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: Unit 2 - July 7, 1993
(58 FR 36445); Unit 3 - June 23, 1993 (58 FR 34094). The two
supplemental letters dated July 6, 1994, and the January 27, 1995,
supplemental letter provided clarifying information and did not change
the initial no significant hazards consideration determination.The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated July 18, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: March 30, 1994
Brief description of amendment: The amendments implement an analog
transmitter/trip system on BFN Unit 3, revise the reactor vessel water
level safety limit and limiting safety system setting for BFN Units 1
and 3, add instrument identifiers and revise calibration frequencies
and functional test requirements for BFN Unit 2, revise the calibration
frequency for instrumentation actuating the suppression chamber-reactor
building vacuum breakers, and provide editorial changes.
Date of issuance: July 17, 1995
Effective date: July 17, 1995
Amendment Nos.: 222, 237, 196
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994(59
FR 49435) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 17, 1995.No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 356114.
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: January 30, 1995
Brief description of amendment: The proposed amendment revises
reactor coolant system pressure-temperature curves, changes bases for
Technical Specification 3/4.4.9, Pressure Temperature Limits, and
revises License Condition 2.C(3)(d) to reflect a change from 10
effective full power years (EFPY) to 21 EFPY.
Date of issuance: July 20, 1995
Effective date: July 20, 1995
Amendment No.: 199
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 15, 1995 (60 FR
14029)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 20, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: September 8, 1994
Brief description of amendment: The amendment revises Technical
Specifications (TS) 4.2.2.2, 4.2.2.4, and 6.9.19. The changes address
[[Page 39462]]
incorporating a penalty in the Core Operating Limits Report (COLR) to
account for heat flux (FQ) increases greater than 2 percent
between measurements.
Date of issuance: July 20, 1995
Effective date: July 20, 1995
Amendment No.: 101
Facility Operating License No. NPF-30. Amendment revises the
Technical Specification Surveillance Requirements and Administrative
Controls.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65823). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 20, 1995. No significant
hazards consideration comments received: No
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of application for amendments: November 22, 1994
Brief description of amendments: The amendments revised the
Technical Specifications to delete unnecessary descriptive phrases
regarding the number of cells in the station and emergency diesel
generator batteries.
Date of issuance: July 11, 1995
Effective date: July 11, 1995
Amendment Nos.: 201 and 201
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 12, 1995 (60 FR
18630)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 11, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Dated at Rockville, Maryland, this 2nd day of August, 1995.
For the Nuclear Regulatory Commission
Jack W. Roe, 4Director, Division of Reactor Projects - III/IV, Office
of Nuclear Reactor Regulation
[Doc. 95-18810 Filed 8-1-95; 8:45 am]
BILLING CODE 7590-01-F