[Federal Register Volume 59, Number 128 (Wednesday, July 6, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-16174]
[[Page Unknown]]
[Federal Register: July 6, 1994]
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UNITED STATES NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating LicensesInvolving
No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from June 13, 1994, through June 23, 1994. The
last biweekly notice was published on June 22, 1994 (59 FR 32226).
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, 11555 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By August 5, 1994, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to John N. Hannon: petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
room for the particular facility involved.
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 2, 1994
Description of amendment request: The proposed amendment would
merge Toledo Edison Company into Cleveland Electric Illuminating
Company. As described in the application, the company formed from the
merger is intended to be renamed. Therefore, the licensee uses the
nomenclature ``NEWCO'' as a temporary name of the combined operating
company, and will provide the permanent name by supplemental letter.
The amendment would (1) replace the Toledo Edison Company and Cleveland
Electric Illuminating Company with ``NEWCO'' as a licensee, (2)
designate ``NEWCO'' as the owner of the Perry Nuclear Power Plant, Unit
1, and (3) make other administrative changes to the license as
indicated in the amendment application. Centerior Service Company would
be unaffected by the amendment and would remain a licensee.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to the Operating License are administrative
and have no effect on any plant systems. All Limiting Conditions for
Operation, Limiting Safety Systems Settings and Safety Limits
specified in the Technical Specifications remain unchanged. This
change meets one of the examples of a change not likely to involve a
significant hazards consideration in that it is a purely
administrative changes (48 FR 14864).
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes to the Operating License are administrative
and have no effect on any plant systems. All Limiting Conditions for
Operation, Limiting Safety Systems Settings and Safety Limits
specified in the Technical Specifications remain unchanged. This
change meets one of the examples of a change not likely to involve a
significant hazards consideration in that it is a purely
administrative change (48 FR 14864).
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes to the Operating License are administrative
and have no effect on any plant systems. All Limiting Conditions for
Operation, Limiting Safety Systems Settings and Safety Limits
specified in the Technical Specifications remain unchanged. This
change meets one of the examples of a change not likely to involve a
significant hazards consideration in that it is purely an
administrative change (48 FR 14864).
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: John N. Hannon
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: May 20, 1994
Description of amendment request: The proposed amendment would
permit the licensee to use an alternate repair criteria (ARC),
designated as the F* criteria. Use of the F* criteria would
allow tubes with otherwise pluggable indications, to remain in service
as long as the indications are below the designated minimum distance of
the F* criteria. The F* criteria defines a length of 1.7
inches of undegraded expanded tube within the tubesheet as the minimum
distance acceptable for implementing this ARC. Below the F*
length, a circumferential tube defect can exist and the tube can remain
in service. The proposed amendment will change the plugging limit
definition and would exclude plugging steam generator tubes with
indications that satisfy the F* criteria. The F* criteria
maintains the structural integrity of the degraded tube as the primary
pressure boundary and allows the tube to remain in service for heat
transfer and core cooling.
This alternate repair criteria qualification is documented in
Babcock & Wilcox Nuclear Technologies (BWNT) Topical Report BAW-10196 P
Revision 1, ``W-D4 F* Qualification Report'', which is included as
part of the licensee's submittal.
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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The supporting qualification report for the subject criteria
demonstrates that the presence of the tubesheet will enhance the
tube integrity in the region of the tube-to-tubesheet roll
expansions by precluding tube deformation beyond its initial
expanded outside diameter. The resistance to a tube rupture is
strengthened by the presence of the tubesheet in that region. The
results of hardrolling of the tube into the tubesheet provides a
mechanical leak limiting seal between the tube and the tubesheet. A
tube rupture cannot occur because the contact between the tube and
the tubesheet does not permit sufficient movement of tube material.
The type of degradation for which the F* criteria has been
developed (cracking with a circumferential orientation) can
theoretically lead to a postulated tube rupture event provided that
the postulated through-wall circumferential crack exists near the
top of the tubesheet. An evaluation including analysis and testing
has been done to determine the resistive strength of the expanded
tubes within the tubesheet. This evaluation provides the basis for
the acceptance criteria for tube degradation subject to the F*
criteria.
The F* length of roll expansion is sufficient to preclude
tube pullout from tube degradation located below the F*
distance, regardless of the extent of the tube degradation. The
Technical Specification leakage rate requirements and accident
analysis assumptions remain unchanged in the unlikely event that
significant leakage from this region does occur. The tube rupture
and pullout is fully bounded by the existing steam generator tube
rupture analysis included in the UFSAR. The leakage testing of the
roll expanded tubes indicates that for tube expansion lengths
approximately equal to the F* distance, any postulated primary
to secondary leakage from F* tubes would be insignificant. The
proposed alternate repair criteria does not adversely impact any
other previously evaluated design basis accident.
The leakage from an F* tube would be limited by the tube-
to-tubesheet interface since this leak would occur below the
secondary face of the tubesheet. Qualification testing and previous
experience indicate that normal and faulted leakage is well below
Technical Specification and administrative limits creating no
increase in the consequences associated with tube rupture type
leakages. The UFSAR analyzed accident scenarios are still bounding
since the normal and faulted leak rates are well within the normal
operating limit of 150 gallons per day. This conclusion is
consistent with previous F* programs approved and used at other
operating plants.
All of the design and operating characteristics of the steam
generator and connected systems are preserved since the F*
criteria utilizes the ``as rolled'' tube configuration that exists
as part of the original steam generator design. The F* joint
has been analyzed and tested for design, operating, and faulted
condition loadings in accordance with Regulatory Guide 1.121 safety
factors. The potential for tube rupture is not increased from the
original submittal as demonstrated in the qualification analyses and
testing completed in the BWNT report.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
B. The proposed changes do not create the possibility of a new
or different type of accident from any accident previously
evaluated.
Implementation of the proposed F* criteria does not
introduce any significant changes to the plant design basis. Use of
the criteria does not provide a mechanism to initiate an accident
outside of the region of the expanded portion of the tube. In the
unlikely event the failed tube severed completely at a point below
the F* region, the remaining F* joint would retain
engagement in the tubesheet due to its length of expanded contact
within the tubesheet bore. This engagement length would prevent any
interaction of the severed tube with neighboring tubes. Any
hypothetical accident as a result of any tube degradation in the
expanded region of the tube would be bounded by the existing tube
rupture accident analysis. Tube bundle structural integrity will be
maintained. Tube bundle leak tightness will be maintained such that
any postulated accident leakage from F* tubes will be
negligible with regard to offsite doses.
Therefore, there is not a potential for creating the possibility
of a new or different type of accident from any accident previously
evaluated.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
The use of the F* criteria has been demonstrated to
maintain the integrity of the tube bundle commensurate with the
requirements of Reg Guide 1.121 and the primary to secondary
pressure boundary under normal and postulated accident conditions.
Acceptable tube degradation for the F* criteria is any
degradation indication in the tubesheet region, more than the
F* distance from the secondary face of the tubesheet or the top
of the last hardroll contact point which ever is further into the
tubesheet. The safety factors used in the verification of the
strength of the degraded tube are consistent with the safety factors
in the ASME Boiler and Pressure Vessel Code and Reg Guide 1.121 used
in steam generator design. The F* distance has been verified by
various testing to be greater than the length of the roll expanded
tube-to-tubesheet interface required to preclude both tube pullout
and significant leakage during normal and postulated accident
conditions. The protective boundaries of the steam generator
continue to be maintained with the use of the F* criteria. A
tube with the indication of degration previously requiring removal
from service can be kept in service through the F* criteria.
Since the joint is constrained within the tubesheet bore, there is
no additional risk associated with the previously analyzed tube
rupture event. The leak testing acceptance criteria are based on the
primary to secondary leakage limit in the Technical Specifications
and the leakage assumptions used in the UFSAR accident analyses.
Implementation of the alternate repair criteria will decrease
the number of tubes which must be taken out of service with tube
plugs or repaired by sleeves. Both plugs and sleeves reduce the RCS
flow margin; thus, implementation of the F* criteria will
maintain the margin of flow that would otherwise be reduced in the
event of increased plugging or sleeving.
Based on the above, it is concluded that the proposed change
does not result in a significant reduction in margin with respect to
plant safety as defined in the UFSAR or the Technical Specification
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: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: October 22, 1993
Description of amendment request: The proposed amendment would
modify the Reactor Trip System (RTS) and Engineered Safety Feature
(ESF) instrumentation surveillance requirements to incorporate the
applicable changes specified in NRC-approved WCAP-10271 and related
supplements. Four specific changes were approved by the Nuclear
Regulatory Commission for the RTS and Engineered Safety Feature
Actuation System analog channels. These changes are limited to the
specific Reactor Protection System (RPS) channels evaluated in the WCAP
(including all supplements) and are subject to the conditions specified
by the NRC.
1. The surveillance or test frequency may be changed from monthly
to quarterly.
2. The time allowed for a channel to be inoperable or out of
service in an untripped condition may be changed from 1 hour to 6
hours.
3. The time a channel in a functional group may be bypassed to
perform testing may be increased from 2 to 4 hours. This bypass time
applies to either an inoperable channel when testing is done in the
tripped mode or to the channel in test when testing is done in the
bypass mode. The Allowed Outage Time for maintenance of a channel is 12
hours.
4. Routine channel testing may be performed in the bypassed
condition instead of the tripped condition.
In addition, a number of editorial changes are made to improve
clarity, and two-loop operating requirements are proposed to be
deleted. Also, the surveillance test interval for RTS interlocks is
proposed to be changed from monthly to once-per-refueling (about 18
months). Although not part of WCAP-10271, this was previously approved
generically by NRC.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The determination that the results of the proposed changes are
within all acceptable criteria was established in the SER(s) [Safety
Evaluation Report(s)] prepared for WCAP-10271, WCAP-10271 Supplement
1, WCAP-10271 Supplement 2 and WCAP-10271 Supplement 2, Revision 1
issued by letters dated February 21, 1985, February 22, 1989 and
April 30, 1990. Implementation of the proposed changes is expected
to result in an acceptable increase in total Reactor Protection
System yearly unavailability. This increase, which is primarily due
to less frequent surveillance, results in an increase of similar
magnitude in the probability of an Anticipated Transient Without
Scram (ATWS) and in the probability of core melt resulting from an
ATWS and also results in a small increase in Core Damage Frequency
(CDF) due to Engineered Safety Features Actuation System
unavailability.
Implementation of the proposed changes is expected to result in
a significant reduction in the probability of core melt from
inadvertent reactor trips. This is a result of a reduction in the
number of inadvertent reactor trips (0.5 fewer inadvertent reactor
trips per unit per year) occurring during testing of RPS
instrumentation. This reduction is primarily attributable to testing
in bypass and less frequent surveillance.
The reduction in [***] core melt frequency is sufficiently large
to counter the increase in ATWS core melt probability resulting in
an overall reduction in total core melt probability.
The values determined by the WOG [Westinghouse Owners Group] and
presented in the WCAP for the increase in CDF were verified by
Brookhaven National Laboratory (BNL) as part of an audit and
sensitivity analyses for the NRC Staff. Based on the small value of
the increase compared to the range of uncertainty in the CDF, the
increase is considered acceptable.
Changes to Surveillance Test Frequencies for the Reactor Trip
System Interlocks do not represent a significant reduction in
testing. The currently specified test interval for interlock
channels allows the surveillance requirement to be satisfied by
verifying that the permissive logic is in its required state using
the annunciator status light. The surveillance, as currently
required, only verifies the status of the permissive logic and does
not address verification of channel setpoint or operability. The
setpoint verification and channel operability are verified after a
refueling shutdown. The definition of the channel check includes
comparison of the channel status with other channels for the same
parameter. The requirement to routinely verify permissive status is
a different consideration than the availability of trip or actuation
channels which are required to change state on the occurrence of an
event and for which the function availability is more dependent on
the surveillance interval. The change in surveillance requirement to
at least once every 18 months does not therefore represent a
significant change in channel surveillance and does not involve a
significant increase in unavailability of the Reactor Protection
System.
The proposed changes do not result in an increase in the
severity or consequences of an accident previously evaluated.
Implementation of the proposed changes affects the probability
of failure of the RPS but does not alter the manner in which
protection is afforded nor the manner in which limiting criteria are
established.
Therefore, the proposed changes do 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 do not result in a change in the manner in
which the Reactor Protection System provides plant protection. No
change is being made which alters the functioning of the Reactor
Protection System (other than in a test mode). Rather the likelihood
or probability of the Reactor Protection System functioning properly
is affected as described above. Therefore,the proposed changes do
not create the possibility of a new or different kind of accident
nor involve a reduction in a margin of safety as defined in the
Safety Analysis Report.
The proposed changes do not involve hardware changes except
those necessary to implement testing in bypass. Some existing
instrumentation is designed to be tested in bypass and current
technical specifications allow testing in bypass. Testing in bypass
is also recognized by IEEE [Institute of Electrical and Electronics
Engineers] Standards. Therefore, testing in bypass has been
previously approved and implementation of the proposed change for
testing in bypass does not create the possibility of a new or
different kind of accident from any previously evaluated.
Furthermore, since the other proposed changes do not alter the
functioning of the RPS, the possibility of a new or different kind
of accident from any previously evaluated has not been created.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes do not alter the manner in which safety
limits, limiting safety system setpoints or limiting conditions for
operation are determined. The impact of reduced testing other than
as addressed above is to allow a longer time interval over which
instrument uncertainties (e.g., drift) may act. Experience has shown
that the initial uncertainty assumptions are valid for reduced
testing.
Implementation of the proposed changes is expected to result in
an overall improvement in safety by:
a. Less frequent testing will result in fewer inadvertent
reactor trips and fewer actuations of Engineered Safety Feature
Actuation System components.
b. Improvements in the effectiveness of the operating staff in
monitoring and controlling plant operation. This is due to less
frequent distraction of the operator and shift supervisor to attend
to instrumentation testing.
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: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: Walter R. Butler
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: April 19, 1994
Description of amendment request: The proposed amendment would
modify Specifications 3.4.9.3 and 3.4.11 to incorporate power operated
relief valve (PORV) Technical Specification (TS) changes in accordance
with the guidance in Generic Letter 90-06 as implemented in NUREG-1431
Improved Standard Technical Specifications (ISTS), with some exceptions
and modifications to reflect plant specific design features. Certain
other TS sections would also be modified to address related TSs.
The proposed changes involve the details of (a) limiting conditions
of operation, and (b) surveillance testing for equipment needed to
protect the reactor vessel from overpressure conditions. This equipment
includes PORVs and their associated block valves, charging pumps,
reactor coolant system (RCS) vent, accumulators, and the overpressure
protection system. Numerous administrative changes are also proposed,
such as renumbering sections, spelling out mathematical symbols,
changes in nomenclature for consistency, and relocating sentences and
paragraphs.
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 changes consolidate the power operated relief valve
requirements into Specifications 3.4.9.3 and 3.4.11 which generally
adopt the new Improved Standard Technical Specifications of NUREG-
1431 to address the concerns identified in Generic Letter 90-06
except for those changes required to reflect plant specific design
features. These changes are proposed to enhance safety and improve
the reliability of the PORVs and block valves. Since the proposed
changes augment or preserve the requirements contained in the
current technical specifications, we have concluded that these
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated in the UFSAR
[Updated Final Safety Analysis Report].
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not involve any physical changes to the
PORVs or their setpoints. These changes do not delete any function
previously provided by the PORVs nor has the probability of
inadvertent opening been increased. Accordingly, no new failure
modes have been defined for any plant system or component important
to safety nor has any new limiting single failure been identified as
a result of these changes. Therefore, these changes will not create
the possibility of a new or different kind of accident from any
accident previously evaluated in the UFSAR.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes have been incorporated to enhance safety
and improve the reliability of the PORVs and block valves to ensure
their availability when called upon to perform their function. These
changes do not affect the manner by which the facility is operated
or involve a change to equipment or features which affect the
operational characteristics of the facility. Therefore, operation of
the facility in accordance with the proposed amendment would not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis. The staff notes
that a significant effort has been made by the NRC and by industry over
the last several years to improve and tighten the TS requirements for
overpressure protection systems. These efforts are documented in
Generic Letter 90-06 and NUREG-1431. The changes proposed by the
licensee appear to result in TSs that are significantly more
comprehensive and restrictive than those now existing for Beaver Valley
Units 1 and 2, and, therefore, should help to reduce the probability of
an accident. The staff disagrees with the licensee's claim that the
changes do not affect the manner by which the facility is operated
(consideration number 3 above). However, the staff believes that the
proposed TS changes require more restrictive operation (such as more
careful control of the number of charging pumps which can inject into
the RCS) and do not involve a significant reduction in a margin of
safety. Based on the NRC staff's 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: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: Walter R. Butler
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: June 9, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification section 4.8.1.1.2 to replace the current
qualitative examination of new diesel generator fuel oil for water/
sediment and particulate contamination with a quantitative examination
for the same properties.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated
since the diesel generator availability and reliability is not being
changed. The quantitative acceptance criteria for new fuel oil is
not being changed. Diesel generator performance will therefore not
be changed due to the proposed revision to SR [surveillance
requirement] 4.8.1[.1].2.d.1.d. The diesel generator will continue
to provide sufficient electrical power to ESF [engineered safety
feature] systems. The ESF systems will continue to function, as
assumed in the safety analyses, to ensure that the fuel, reactor
coolant system, and containment
design limits are not exceeded.
Therefore, this changes will not increase the probability or
consequences of an accident previously evaluated due to the
continued availability and reliability of the emergency diesel power
source.
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 alter the method of operating the
plant. This change will continue to ensure that the addition of new
fuel oil complies with accepted standards regarding fuel oil
quality. Since design requirements continue to be met and the
integrity of the reactor coolant system pressure boundary is not
challenged, no new failure mode has been created. As a result, an
accident which is different than any already evaluated in the
Updated Final Safety Analysis Report will not be created due to this
change.
Therefore, the proposed change does 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 margin of safety is not reduced because the emergency diesel
generators will continue to provide sufficient capacity, capability,
redundancy, and reliability to ensure availability of necessary
power to ESF systems. The ESF systems will continue to function, as
assumed in the safety analyses, to ensure that the fuel, reactor
coolant systems, and containment design limits are not exceeded. The
replacement of the clear and bright qualitative examination with the
proposed quantitative test to determine the actual water/sediment
and particulates will ensure that new fuel oil meets the required
limits for these properties prior to addition to the storage tank,
therefore assuring that the quality of the stored fuel is unaffected
by the addition of new fuel.
Therefore, this proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: Walter R. Butler
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley
Power Station, Unit No. 2, Shippingport, Pennsylvania
Date of amendment request: February 16, 1994
Description of amendment request: The proposed amendment would
delete the Appendix B Section 4.2.2 requirement to perform infrared
aerial photography every other year.
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 will delete from Facility Operating License
No. NPF-73 the Appendix B Section 4.2.2 requirement to perform
infrared aerial photography every other year. The acceptance limit
which forms the licensing basis for this technical specification is
related to environmental impact and has no impact on the margin of
safety, accident analysis, or other design basis impacting the
margin of safety. No increase in adverse environmental impact has
been identified over that previously identified in the Final
Environmental Statement - Operating License Stage, environmental
impact appraisals, or in any decisions of the Atomic Safety and
Licensing Board. The Final Environmental Statement concluded, based
on a model of combined drift from Units 1 and 2, that no adverse
impacts to sensitive species of natural vegetation or to sensitive
species of crops were expected. The staff also examined infrared
aerial photographs taken from 1975 through 1983 and found no injury
to vegetation from cooling tower drift in the vicinity of Unit 1.
Continued terrestrial monitoring was performed for Beaver Valley
Unit 2 by infrared aerial photography in 1986, 1988, 1990, and 1992.
The results as provided in the Annual Environmental Reports Non-
Radiological concluded, ``Based on interpretation of the infrared
photographs and field verification, there is no evidence to suggest
that the BVPS [Beaver Valley Power Station] cooling towers are
causing vegetation stress.''
Based on the compilation of the infrared aerial photography
performed for both BV-1 and BV-2, deletion of this terrestrial
monitoring requirement will have no impact on the environment or the
operation of the plant. Therefore, the proposed change will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The Appendix B Section 4.2.2 requirement to perform infrared
aerial photography reflects a commitment described in Final
Environmental Statement Section 5.14.1. Therein it is stated that
the preoperational monitoring studies for BV-2 are based primarily
on the BV-1 operational monitoring programs. ``Results of these
studies have shown that there were no BV-1 operational impacts on
flora, thus, the only terrestrial monitoring planned for BV-2 is
continued infrared aerial photography every other year. The
photographs will be compared with preoperational photographs of the
BV-2 area, nd any signs of injury as a result of salt drift and
other sources will be checked. The details of this terrestrial
monitoring program will be specified in the Environmental Protection
Plan that will be included in Appendix B of the operating license.''
The subject of this concern is the impact of salt and water drift on
area vegetation including sensitive agricultural crops. From the
standpoint of soil salinization (the effects of the accumulation of
salts in the soil), described in the Environmental Report-Operating
License Stage Section 5.3.3, no appreciable impact resulting from
operation of the natural draft cooling towers is anticipated. This
is based on the average rate of precipitation of 36.2 inches
annually which greatly reduces the potential for accumulation of
salt in the soil. The terrestrial monitoring program has been
performed in accordance with the Environmental Protection Plan and
has provided additional verification that operation of both cooling
towers has not produced any evidence of vegetation stress. 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?
Infrared aerial photography surveillance does not affect safety
systems and/or systems important to safety. Terrestrial monitoring
is not used in any accident analysis and does not provide a basis
for evaluating the radiological consequences of an accident.
Deleting the requirement to perform infrared aerial photography will
not result in any environmental impact from operation of the cooling
tower and will not affect the operation of the cooling tower. The
operating history of both the BV-1 and the BV-2 cooling towers has
demonstrated that there is no evidence of vegetation stress in
accordance with the results obtained from the infrared aerial
photography and other associated methods of environmental
monitoring. Therefore, based on the above, the proposed change will
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: Walter R. Butler
Florida Power and Light Company, et al., Docket No. 50-335, St.
Lucie Plant, Unit No. 1, St. Lucie County, Florida
Date of amendment request: May 23, 1994
Description of amendment request: The amendment will revise
Technical Specification (TS) 3.5.2 for Emergency Core Cooling Systems
(ECCS) by removing the option that allows High Pressure Safety
Injection (HPSI) Pump 1C to be used as an alternative to the preferred
pump for subsystem operability. HPSI pump 1C is an installed spare
which is not required to be maintained in an operable status, and this
change is being requested to upgrade the ECCS operability requirements
consistent with actual plant operating needs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed license amendment will remove the option of using
High Pressure Safety Injection Pump 1C (HPSI-1C) to satisfy, in
part, the Emergency Core Cooling System (ECCS)
operabilityrequirements specified in Limiting Condition for
Operation (LCO) 3.5.2. HPSI-1C is an installed spare pump that is
not required to be operable unless it is being used in place of the
preferred B-train ECCS high pressure pump. The required functional
response of the ECCS or the required availability of the minimum
equipment necessary to accomplish the ECCS safety function will not
be changed by removing the spare pump option from the Technical
Specifications.
The calculated cooling performance of the St. Lucie Unit 1 ECCS
during postulated accidents conforms to the criteria set forth in 10
CFR 50.46 and the ability to achieve this required performance,
including considerations of single-failure criteria, is independent
from optional use of HPSI-1C. Therefore, operation of the facility
in accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. Eliminating the
option for the licensee to utilize HPSI-1C in place of the preferred
B-train ECCS high pressure pump does not involve the addition of new
or different types of equipment to the previously analyzed ECCS.
Equipment important to safety will continue to perform their safety
functions as previously analyzed and will not be affected by this
proposed amendment. Therefore, operation of the facility in
accordance with the proposed amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
Removing the option to employ the installed spare HPSI pump 1C
in lieu of the preferred B-train pump to determine ECCS operability
only removes an operational flexibility that has rarely been used by
the licensee. St. Lucie Unit 1 accident analyses do not take credit
for an installed spare pump, the minimum complement of safety
injection equipment required for safe operation of the facility and
that is required by LCO 3.5.2 is not changed, and the results of
plant accident and transient analyses are not influenced by this
proposed amendment. The proposed change does not alter the bases for
any Technical Specification related to the establishment of, or
maintenance of, a nuclear safety margin. Therefore, operation of the
facility in accordance with the proposed amendment would not involve
a significant reduction in a margin of safety.
Based on the discussion presented above and on the supporting
Evaluation of Proposed TS Changes, FPL has concluded that this
proposed license amendment involves no significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: May 23, 1994
Description of amendment request: The proposed amendments will
relocate the seismic monitoring instrumentation Limiting Conditions for
Operation, Surveillance Requirements, and the associated tables
contained in Technical Specifications 3.3.3.3, 4.3.3.3.1 and 4.3.3.3.2
to the Updated Final Safety Analysis Report. The basis for this request
is consistent with NUREG-1432, ``Standard Technical Specifications,
Combustion Engineering Plants'' and with the ``Final Policy Statement
on Technical Specifications Improvements for Nuclear Power Reactors,
``published in the Federal Register (58 FR 39132) dated July 22, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes are administrative in nature in that the
specifications for operation and surveillance of the Seismic
Monitoring Instrumentation system will be relocated from Appendix A
of the facility operating license to the Updated Final Safety
Analysis Report for St. Lucie Unit 1 and Unit 2. Changes to the
system will be controlled by 10 CFR 50.59 and the safety analysis
report is required to be updated pursuant to 10 CFR 50.71(e).
Relocation of these requirements to the UFSAR is consistent with the
NRC ``Final Policy Statement on Technical Specifications
Improvements for Nuclear Power Reactors'' published in the Federal
Register (58 FR 39132) dated July 22, 1993.
Seismic monitoring instrumentation is not an accident initiator
nor a part of the success path(s) which function to mitigate
accidents evaluated in the plant safety analyses. The proposed
technical specification change does not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident, nor do the changes
alter any assumptions or conditions in any of the plant accident
analyses. Therefore, operation of the facility in accordance with
the proposed amendment would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment to relocate the existing Technical
Specification requirements for Seismic Monitoring Instrumentation to
the Updated Final Safety Analysis Report will not change the
physical plant or the modes of plant operation defined in the
Facility License. The change does not involve the addition or
modification of equipment nor does it alter the design or operation
of plant systems. Therefore, operation of the facility in accordance
with the proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed changes are administrative in nature in that
operating and surveillance requirements for the Seismic Monitoring
Instrumentation system will be relocated from Appendix A of the
facility license to the Updated Final Safety Analysis Report for St.
Lucie Unit 1 and Unit 2. Seismic monitoring instruments are not used
to actuate safety-related equipment, provide interlocks, or
otherwise perform plant control functions. The instruments are used
to record the magnitude of a seismic event, should it occur.
Conditions evaluated in plant accident and transient analyses do not
involve seismic instruments. The proposed changes do not alter the
basis for any technical specification that is related to the
establishment of, or the maintenance of, a nuclear safety margin.
Therefore, operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
Based on the above discussion and the supporting Evaluation of
Technical Specification changes, FPL has determined that the
proposed license amendment involves no significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: May 23, 1994
Description of amendment request: The proposed amendment revises
Technical Specifications Section 3/4.7.1.1, Turbine Cycle, Safety
Valves, to delete a specific reference to the 1974 edition of the ASME
Code and refer to testing in accordance with Technical Specification
4.0.5, the In-Service Inspection and In-Service Testing Specification.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1)Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the main steam code safety valves will continue to
be tested in accordance with current NRC requirements as implemented
through 10 CFR 50.55a. The NRC specifies the ASME code requirements
for a facility through revisions to 10 CFR 50.55a and through the
review and approval of the plant specific in-service testing plan
for pumps and valves at the beginning of each in-service inspection
interval.
The probability or consequences of an accident are not increased
because testing of the main steam safety valves is in accordance
with the appropriate NRC requirements.
(2) Use of the modified specification would not create the
possibility of a new or different kind of accident from any
previously evaluated.
The use of this modified specification can not create the
possibility of a new or different kind of accident from any
previously evaluated since there is no physical change to the
facility or the set points for the main steam safety valves. The
valves will be tested in accordance with current requirements. No
new failure mode is introduced due to the change because no plant
change is being made and main steam safety valve test methods are
consistent with the endorsed edition of the ASME Code.
(3) Use of the modified specification would not involve a
significant reduction in a margin of safety.
The existing technical specification references an outdated
version of the ASME Code. This change corrects the reference to
Specification 4.0.5 which ensures that in-service testing of ASME
Code Class 1, 2, and 3 pumps and valves will be performed in
accordance with a periodically updated version of Section XI of the
ASME Boiler and Pressure Vessel Code and Addenda as required by 10
CFR 50.55a.
Safety valve setpoints or tolerances are not changed by this
proposal. Therefore, the modified specification corrects the ASME
Code reference and does not involve a significant reduction in a
margin of safety.
Based on the above, we have determined that the proposed
amendment does not (1) involve a significant increase in the
probability or consequences of an accident previously evaluated, (2)
create the probability of a new or different kind of accident from
any previously evaluated, or (3) involve a significant reduction in
a margin of safety; and therefore does not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: March 27, 1992, as supplemented on
January 6, and May 27, 1994.
Description of amendment request: The proposed amendment would
revise the limiting conditions for operation and surveillance
requirements for primary containment integrity, secondary containment
integrity and other systems and equipment of Technical Specifications
Section 3.7.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated because the requested revisions do not affect
the FSAR safety analyses involving these system.
Definitions
The revisions to Definition 15, ``Primary Containment
Integrity'' and Definition 16, ``Secondary Containment Integrity''
agree with the corresponding definitions of the STS. These changes
are administrative in nature in that they only clarify the
requirements for containment integrity and the appropriate means of
isolating penetrations. These changes do not affect the operation or
function of the containment isolation systems and, therefore, do not
result in a significant increase in the probability or consequences
of an accident previously evaluated.
Primary Containment Integrity
The revision to TS section 3.7.A, ``Primary Containment
Integrity'', only adds a specific requirement to restore primary
containment integrity within 1 hour or commence a plant shutdown.
These actions are consistent with the actions specified in STS for
primary containment integrity. No changes to the primary containment
boundary or the requirements for primary containment integrity have
been proposed. Therefore, this change does not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
Primary Containment Power Operated Isolation Valves
The revisions to TS section 3.7.B, ``Primary Containment Power
Operated Isolation Valves'' are editorial in nature in that the
wording has only been changed to be consistent with the STS
requirements for primary containment isolation valves. These changes
do not affect the function of the valves, the requirements to
isolate a penetration with an inoperable containment isolation valve
or the actual methods of isolation. Penetrations are still required
to be isolated within 4 hours in a manner that cannot be adversely
affected by a single active failure. Therefore, these changes do not
result in a significant increase in the probability or consequences
of an accident previously evaluated.
Drywell Average Air Temperature
The addition of limits, actions, and surveillance requirements
for drywell average air temperature are intended to ensure that the
initial assumptions in the DAEC Primary Containment Response
Analysis to a DBA remain valid. The temperature limit (135 deg.F)
corresponds to the initial drywell average temperature assumed for
this analysis in the UFSAR. The specified limits, actions and
surveillance requirements are consistent with STS. The addition of
this limit to the TS will not affect the actual operation or
function of any equipment but will ensure that the containment
analysis remains valid. Therefore, the addition of this limit will
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
Pressure Suppression Chamber-Reactor Building Vacuum Breakers
The changes to TS section 3/4.7.D only provide additional detail
and operability requirements for the pressure suppression chamber-
reactor building vacuum breakers. These additional details are
consistent with the requirements of STS. Specifying separate
operability requirements for vacuum breakers inoperable for opening
(but known to be closed), or open better reflects the dual functions
of these valves (vacuum relief and containment isolation). The
additional surveillance requirement will better ensure that the
containment isolation function of these valves is maintained. The
rewording of existing surveillances only clarifies current
requirements. These changes do not affect the actual function,
setpoints, or number of valves required to be operable and therefore
do not result in a significant increase in the probability or
consequences of an accident previously evaluated.
Drywell-Pressure Suppression Chamber Vacuum Breakers
The changes to TS section 3/4.7.E, only provide additional
detail and operability requirements for the drywell-pressure
suppression chamber vacuum breakers. These additional details are
consistent with the requirements of STS. Specifying separate
operability requirements for vacuum breakers inoperable for opening
(but known to be closed) or open better reflects the dual functions
of these valves. The additional requirement to verify that each
vacuum breaker is closed at least once per week will better ensure
that the isolation boundary between the drywell and torus is
maintained. The elimination of the requirement to exercise all
operable drywell-pressure suppression chamber vacuum breakers upon
determination that a vacuum breaker is inoperable for opening will
not affect the reliability of these vacuum breakers. The only valid
reason to exercise the operable vacuum breakers is if a common mode
failure is suspected. We have reviewed the maintenance history of
these valves and have not identified any instance of common mode
failures. Conditional testing of these changes will not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
Main Steam Isolation Valve Leakage Control System (MSIV-LCS)
The change to TS Section 3/4.7.F, ``MSIV-LCS'', deletes the
unnecessary and potentially non-conservative conditional
surveillance testing of the redundant MSIV-LCS subsystems. Although
the proposed change will reduce the amount of testing of the MSIV-
LCS, reliability of these systems would not be decreased and the
necessary assurance that the alternate systems/subsystems/components
will operate when needed is provided by the ASME Section XI IST
Program.
The possibility of human error will decrease with reduced
testing. Human error such as a misalignment of valves after the
system is returned to its normal configuration following testing and
the misdirection of the operators attention from monitoring and
directing plant operations is less likely to occur if this testing
is eliminated. Additionally, reducing the scope and frequency of
surveillance testing will decrease the probability of equipment
failure (due to testing) which could require plant shutdown.
Therefore, this change will not increase the probability of
occurrence or consequences of an accident previously evaluated.
Suppression Pool Level and Temperature
The changes to TS section 3/4.7.F, ``Suppression Pool Level and
Temperature'', are intended to clarify these requirements and make
them more consistent with STS. The revision to the applicability
statement which deletes the requirement for suppression pool level
and temperature to be within the specified limits during work which
has the potential to drain the vessel is in accordance with STS.
Suppression pool level and temperature limits ensure that the
suppression pool has the capability of acting as a heat sink for
design basis events but are not appropriate or applicable during the
refueling or cold shutdown conditions. No changes have been made to
the actual suppression pool temperature or level limits and
therefore, the assumptions made in the accident and transient
analyses remain valid. These limits are consistent with the STS. The
revisions to the surveillance requirements are also intended to
improve clarity and consistency with STS. The deletion of the
requirement to monitor suppression pool water temperature every 5
minutes during relief valve operation is appropriate in that plant
operating and emergency operating procedures already specify what
actions are to be taken when suppression pool average water
temperature increases above 95 deg.F including initiation of
suppression pool cooling. Monitoring pool temperature every 5
minutes during these events is not necessary and is redundant to
other actions. Therefore, these changes will not significantly
increase the probability of occurrence of the consequences of an
accident previously evaluated.
Containment Atmosphere Dilution
The revisions to the applicability of TS section 3.7.H,
``Containment Atmosphere Dilution'', requiring the containment
atmosphere dilution system to be operable only when the reactor is
in power operation and the primary containment is required to be
inerted will not significantly increase the probability or
consequences of an accident previously evaluated because the CAD
system can only function when the containment is inerted. The
function of the CAD system is to inject nitrogen into the
containment after a LOCA and ensure the containment remains inerted.
Drywell inspections performed after plant startup and prior to plant
shutdown require that the primary containment be de-inerted for
personnel access. Therefore, CAD system operability is not required
during these inspections. No changes to the actual function or
purpose of the CAD system are proposed.
Oxygen Concentration
The changes to TS section 3/4.7.1, ``Oxygen Concentration'' are
administrative in that they only clarify the requirement that both
the suppression chamber and the drywell must have oxygen
concentrations less than 4 [percent] by volume. The revisions to the
surveillance requirements are consistent with STS. Decreasing the
frequency of verification of oxygen concentration from twice per
week to once per week is in accordance with STS and reflects the
fact that during power operation, the containment is inerted and
slightly pressurized such that air (oxygen) cannot leak into the
containment. Therefore, these changes will not significantly
increase the probability or consequences of an accident previously
evaluated.
Secondary Containment
The deletion of the requirement to operate the SGTS immediately
after a secondary containment violation is identified will not
affect the reliability of the secondary containment in that
containment integrity is normally fully restored immediately after a
violation is identified. The testing of the SGTS involves insertion
of a Group III containment isolation signal and is only appropriate
if the restoration of secondary containment involves a temporary or
new secondary containment boundary. These modifications to a
secondary containment boundary, however, would require that the SGTS
be operated as part of post modification testing. Deleting the
requirement for the SGTS to be operated after minor secondary
containment violations will reduce the possibility of human error
(such as misalignment of valves after the system is returned to its
normal configuration) due to reduced testing. Operation of the SGTS
after a secondary containment violation is not required by STS.
Revision of the definition of calm wind conditions will not
affect the reliability or availability of the secondary containment
or SGTS. An engineering evaluation on the effects of wind speed and
direction on the ability of the SGTS to maintain 1/4'' vacuum in
secondary containment has been performed. The results indicate that
while wind effects can be seen on individual instruments, there is
minimal effect on the average instrument readings with wind speeds
up to 15 mph. A discussion of this evaluation has been added to the
Bases of TS section 3.7. Therefore, these changes will not
significantly increase the probability or the consequences of an
accident previously evaluated.
Secondary Containment Automatic Isolation Dampers
The addition of operability requirements, actions and
surveillance requirements for secondary containment isolation
dampers better ensures the integrity and isolation capability of the
secondary containment. The new specifications are consistent with
the requirements of the STS. The actual function or operation of the
secondary containment isolation valves/dampers will not be affected.
The appropriate valves/dampers will be incorporated in plant
procedures that are subject to the change control provisions of TS.
Therefore, these changes will not increase the probability of
occurrence or consequences of an accident previously evaluated in
the TS.
Standby Gas Treatment System
The change to the output requirements of the inlet heaters for
each train of the SGTS from 11 kw to 22 kw better ensures that these
heaters (and the SGTS) can perform their design function. The 22 kw
output requirement ensures that the inlet air humidity does not
exceed the 70 [percent] humidity specified in the UFSAR. This change
does not affect the actual operation of the heaters or the SGTS.
The requirement to demonstrate the HEPA filter uniform air
distribution after HEPA filter replacement or after structural
maintenance on the filter system housing (rather than annually) will
not decrease the reliability of the SGTS. The air flow test will be
performed after work or modifications which have the ability to
disrupt the system geometry or result in potential flow blockage.
Revising the shutdown LCO requirement in the various
specifications from requiring the plant to be in Cold Shutdown in 24
hours to requiring Hot Shutdown in 12 hours and Cold Shutdown (or
other condition not requiring equipment operability) in the
following 24 hours is consistent with STS and the shutdown
requirements in TS section 3.5. This new requirement will allow the
reactor to be shutdown in a more controlled manner and will not
result in a significant increase in the probability or consequences
of an accident previously evaluated.
The revisions to the Bases are administrative in that they only
reflect the changes to the individual specifications described
previously in this section or correct minor discrepancies. All
changes are consistent with the applicable specifications.
(2) The proposed amendment will not increase the possibility of
a new or different kind of accident from any accident previously
evaluated for the following reasons.
As described in the above response to question 1, none
of the proposed changes alters the design of the plant or equipment
or the plant's transient response. The changes to the definitions
and limiting conditions for operation applicable to TS section 3.7
are consistent with STS and better ensure that equipment assumed to
be operable in our accident analysis will be operable upon demand.
The addition of limiting conditions for operation for drywell
average temperature and secondary containment isolation valves will
better ensure that the assumptions in our accident analysis remain
valid.
The changes to the surveillance requirements are consistent with
the STS. Those systems required to mitigate accidents evaluated in
the UFSAR will still be operable and available.
The reduction in conditional surveillance testing of certain
systems and equipment will reduce the probability of equipment
failure as a result of excessive testing or due to human error.
(3) The proposed amendment will not involve a significant
reduction in a margin of safety for the following reasons.
The revisions to the limiting conditions for operation in
Chapter 3.7 of the TS will not invalidate the original licensing
basis assumptions and will not invalidate any assumptions or input
parameters for any DAEC event analysis. These changes provide more
specific guidance only and are in accordance with the STS. Extending
the time period within which the DAEC must achieve Cold Shutdown
conditions will permit increased operator attention and minimal
distractions for operators during shutdown, thus minimizing the
risks of unexpected operational transients.
Additional surveillance testing for certain instrumentation and
systems will provide additional assurance that these systems will be
available when needed.
Elimination of unnecessary or conditional surveillance testing
will not reduce the minimum necessary equipment operability
requirements or equipment reliability. Elimination of the redundant
testing will reduce equipment failure due to excessive testing or
human error.
In summary, the proposed administrative changes do not change
the probability or consequences of an accident previously evaluated,
do not create the possibility of a new or different kind of accident
and do not involve a reduction in the margin of safety.
Therefore, the proposed license amendment is judged to involve
no significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036
NRC Project Director: John N. Hannon
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: May 27, 1974
Description of amendment request: The amendment would temporarily
allow the Operations Manager to not have a senior reactor operator
(SRO) license for Millstone 3, providing other conditions are met.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:...The proposed change does not
involve an SHC [significant hazards consideration] because the change
would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change affects only an administrative control,
which was based on the existing industry guidance in ANSI N18.1-
1971, that recommended the Operations Manager hold a SRO license.
The current guidance in Section 4.2.2 of ANSI/ANS-3.1-1987
recommends, as one option, that the Operations Manager have held a
license for a similar unit and the Operations Middle Manager hold a
SRO license. The Operations Middle Manager position does not exist
at NNECO [Northeast Nuclear Energy Company]. Therefore, the proposed
change requests an exception to ANSI N18.1-1971 to allow use of
ANSI/ANS-3.101987 in a limited circumstance. Specifically, the
proposed revision to Technical Specification 6.3.1 would temporarily
allow the Operations Manager to have held a SRO at a PWR
[pressurized water reactor] other than Millstone Unit No. 3. The
proposed revision would be in effect for the period ending three
years after the Staff's approval for this request.
The proposed exception to ANSI N18.1-1971 will allow at least
one of the Operations Assistants (instead of a Operations Middle
Manager) to hold, and continue to hold, a SRO license, if the
Operations Manager does not hold a license. The proposed change
includes the requirement if the Operations Manager does not hold a
SRO license at Millstone Unit No. 3, he shall have held a license
for a similar unit in accordance with Section 4.2.2 of ANSI/ANS-3.1-
1987. For those areas of knowledge that require a SRO license, at
least one of the Operations Assistants hold a SRO license and
provides technical guidance normally required by the Operations
Manager.
The proposed change does not alter the design of any system,
structure, or component. It does not change the way any plant
systems are operated. It does not reduce the knowledge,
qualifications, or skills of any operator on watch, and does not
affect the way the Operations Department is managed by the
Operations Manager in maintaining the effective performance of his
personnel and to ensure the plant is operated safely and in
accordance with the requirements of the Operating License.
The proposed change does not detract from the Operations
Manager's ability to perform his primary responsibilities. In this
case, by having previously held a SRO license for a similar unit, he
will have gained the necessary training, skills, and experience to
fully understand the operation of plant equipment and the watch
requirements for operators.
The proposed change does not weaken the supervisory chain that
presently exists in the Operations Department. All Control Room
operators will continue to be supervised by the licensed Shift
Supervisor.
In summary, the proposed change does not affect the ability of
the Operations Manager to provide the plant oversight required of
his position. In addition, it does not have any [e]ffect on the
probability or consequences of any previously evaluated accident.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed change to Technical Specification 6.3.1 does not
affect the design or function of any plant system, structure, or
component. It does not affect, in any way, the performance of NRC
licensed operators, nor does it change the way any plant equipment
is operated. Operation of the plant in conformance with technical
specifications and other license requirements will continue to be
supervised by personnel who hold an NRC SRO license. The proposed
change to Technical Specification 6.3.1 ensures that the Operations
Manager will be a knowledgeable and qualified individual. The
proposed change does not introduce any new failure modes.
3. Involve a significant reduction in a margin of safety.
The proposed change involves only an administrative control
which is not related to the margin of safety as defined in the
technical specifications. The proposed change does not reduce the
level of knowledge or experience required of an individual who fills
the Operations Manager position, nor does it affect the conservative
manner in which the plant is operated. All Control Room operators
will continue to be supervised by personnel who hold a SRO license.
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, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: May 31, 1994
Description of amendment request: This amendment will change the
frequency for monitoring the Spray Pond ground water level from once
per month to once every six months in Technnical Specification
Requirement 4.7.1.3.c for each Susquehanna unit.
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:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed Technical Specification change to extend the
monthly surveillance interval for Spray Pond ground water level
measurement to biannual does not affect the the probability or
consequences of an accident previously evaluated. The safety
analysis performed for this change concludes that the ground water
level is stable, predictable, and significantly below the acceptance
criteria established in the Technical Specifications. Thus, less
frequent monitoring of the ground water level does not increase the
probability that the Spray Pond will become inoperable due to rising
ground water levels or the probability of any accident scenarios
associated with Spray Pond inoperability. The Technical
Specification change will not impact the function or the method of
operation of plant systems, structures, or components. Thus, the
consequences of a malfunction of equipment important to safety
previously evaluated in the FSAR are not increased by the change.
II. This proposal does not create the possibility of a new or
different kind of accident or from any accident previously
evaluated.
The proposed Technical Specification change to extend the
monthly surveillance interval for Spray Pond ground water level
measurement to biannual does not create the possibility of a new or
different kind of accident or from any accident previously
evaluated. The proposed change dies not affect systems, structures,
or components (SSCs) or the operation of the SSCs; and therefore
does not create the possibility of a new or different kind of
accident.
III. This change does not involve a significant reduction in a
margin of safety.
The proposed Technical Specification change to extend the
monthly surveillance frequency to biannual does not reduce the
margin of safety. The ground water level in the vicinity of the
Spray Pond has been proven to be stable and predictable through
twelve years of monthly data collection. This data has shown the
highest ground water levels (still considerably lower than the
Technical Specification limit) to occur in the months of April and
October. Therefore, surveillance of the ground water level at the
observation sites during April and October will adequately monitor
this aspect of the Spray Pond operability.
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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: Charles L. Miller
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of amendment request: May 25, 1994
Description of amendment request: Salem is in the process of
upgrading the Radiation Monitoring System. This upgrade will involve a
replacement on many of the existing radiation monitors. The proposed
change modifies Tables associated with Technical Specifications 3/
4.3.3.1 Radiation Monitoring Instrumentation, 3/4.3.3.8 Radioactive
Liquid Effluent Monitoring Instrumentation, and 3/4.3.3.9 Radioactive
Gaseous Effluent Monitoring Instrumentation. The proposed change
relocates the Salem specific radiation monitor numbers from the table
to a cross reference in the Bases. No required radiation monitoring
functions are being changed.
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. [This proposal does not involve] a significant increase in
the probability or consequences of an accident previously analyzed.
The proposed change to relocate the Salem specific radiation
monitor numbers to the Bases is administrative. There are no
modifications or changes in operating conditions associated with the
proposed changes. Providing a more accurate description of a
referenced radiation monitor in the note is editorial. The proposed
changes do not affect the probability of occurrence or the
consequences of accidents identified in the UFSAR. No accident
precursors are being generated by the proposed changes. Therefore,
the proposed changes do not involve a significant increase in the
probability or consequences of a previously analyzed accident.
2. [This proposal does not create] the possibility of a new or
different kind of accident.
The proposed changes to relocate the Salem specific radiation
monitor numbers to the Bases are administrative. The change to the
note on Table 3.3-6 is editorial to provide a more accurate
description of the required radiation monitor. There are no
modifications or changes in operating conditions associated with the
proposed changes. Therefore, the proposed changes will not increase
the possibility of a new or different kind of accident from any
accident previously identified.
3. [These changes do not involve] a significant reduction in a
margin of safety.
The Technical Specification operability requirements for the
radiation monitors are not being changed. Relocating the Salem
specific radiation monitor numbers to the Bases will not change any
requirements for the radiation monitors. The change to the note on
Table 3.3-6 is editorial to provide a more accurate description of
the required radiation monitor. Therefore, the changes to the
surveillance frequencies do not involve a significant reduction in
any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Charles L. Miller
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of amendment request: May 23, 1994, as supplemented June 15,
1994
Description of amendment request: The proposed amendment would
revise the Ginna Station Technical Specification (TS) 3.1.4 regarding
allowable primary coolant levels of specific activity. The limit for I-
131 dose equivalent of iodine activity in the reactor coolant would be
increased from 0.2 to 1.0 micro Ci/gm. The limit for total specific
activity of the reactor coolant would be increased from 84 to 100/E
micro Ci/gm, where E is the average beta and gamma energies per
disintegration in Mev. Both increased allowable levels are consistent
with NUREG-1431 ``Standard Technical Specifications, Westinghouse
Plants, September 1992,'' and NUREG-0800 ``Standard Review Plan for the
Review of Safety Analysis Reports for Nuclear Power Plants'' (Section
15.6.3).
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:
1. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed changes do not affect any accident initiators and
therefore the probability of any accident is not increased.
Consequences of the changes are analyzed and shown acceptable in the
[Westinghouse LOFTTR2 Analysis of Potential Radiological
Consequences Following a Steam Generator Tube Rupture at the R.E.
Ginna Nuclear Power Plant] analysis, WCAP-11668, Section III.
2. Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed changes involve no physical modifications to the
plant; therefore, no new accident can be postulated.
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety, as no margin of safety is reduced by the proposed changes,
as shown in WCAP-11668.
Based upon the above information, it has been determined that
the proposed changes to the Ginna Station Technical Specifications
do not involve a significant increase in the probability or
consequences of an accident previously evaluated, does not create
the possibility of a new or different kind of accident previously
evaluated, and does not involve a significant reduction in a margin
of safety. Therefore, it is concluded that the proposed changes meet
the requirements of 10 CFR 50.92(c) and do not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005
NRC Project Director: Walter R. Butler
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 6, 1994
Description of amendment request: The proposed amendment would
merge Toledo Edison Company into Cleveland Electric Illuminating
Company. As described in the application, the company formed from the
merger is intended to be renamed. Therefore, the licensee uses the
nomenclature ``NEWCO'' as a temporary name of the combined operating
company, and will provide the permanent name by supplemental letter.
The amendment would (1) replace the Toledo Edison Company and Cleveland
Electric Illuminating Company with ``NEWCO'' as a licensee, (2)
designate ``NEWCO'' as the owner of the Davis-Besse Nuclear Power
Station, Unit 1, and (3) make other associated changes to the license
as indicated in the amendment application. Centerior Service Company
would be unaffected by the amendment and would remain a licensee.
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, indicating that the proposed
changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators or
assumptions are affected. The proposed changes are administrative
and have no direct affect on any plant systems. All Limiting
Conditions for Operation, Limiting Safety System Settings and Safety
Limits specified in the Technical Specifications will remain
unchanged.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no accident conditions or
assumptions are affected. The proposed changes do not alter the
source term, containment isolation, or allowable radiological
consequences. The proposed changes are administrative and have no
direct effect on any plant systems.
2a. Not create the possibility of a new kind of accident from
any accident previously evaluated because no new accident initiators
are created. The proposed changes are administrative and have no
direct effect on any plant systems. The changes do not affect the
reactor coolant system pressure boundary and do not affect any
system functional requirements, plant maintenance, or operability
requirements.
2b. Not create the possibility of a different kind of accident
from any accident previously evaluated because no different accident
initiators are created. The proposed changes are administrative and
have no direct effect on any plant systems. The changes do not
affect the reactor coolant system pressure boundary and do not
affect any system functional requirements, plant maintenance, or
operability requirements.
3. Not involve a significant reduction in the margin of safety
because the proposed changes do not involve new or significant
changes to the initial conditions contributing to accident severity
or consequences. The proposed changes are administrative and have no
direct affect on any plant systems.
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: John N. Hannon
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: May 20, 1994
Description of amendment request: The proposed amendment would
remove Core Spray (CS) High Sparger Pressure Instrumentation from the
Vermont Yankee Technical Specifications for Emergency Core Cooling
System (ECCS) Actuation Instrumentation. In addition, an unrelated
administrative change is also proposed.
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 to remove the Core Spray High Sparger
Pressure Instrumentation from the Technical Specifications for ECCS
Actuation Instrumentation is consistent with NRC requirements
concerning this instrumentation. This instrumentation is considered
NNS [nonnuclear safety] and performs a local monitoring and alarm
function only. In addition, the NRC has recently approved the
removal of Core Spray Sparger Break Detection Instrumentation from
the Technical Specifications of another BWR [boiling water reactor]
with a similar situation.
The CS Sparger Piping is inspected every refueling outage to
verify its integrity. No cracks in the CS Sparger piping have been
identified since the first inspection in 1980. CS Sparger Piping
integrity is still assured. The instrumentation systems to be
removed from the ECCS Actuation Instrumentation Technical
Specifications do not perform any automatic control or trip
function. In addition, this instrumentation does not provide
information that is required to permit the control room operator to
take manual actions that are required for safety systems to
accomplish their safety functions for design basis accident events.
The proposed change does not result in any system hardware
modification, function change or new plant configuration. The
requested change to ECCS Actuation Instrumentation does not impact
any FSAR [Final Safety Analysis Report] safety analysis involving
the ECCS or Protection Systems. These monitoring functions are not
contributors to the initiation of accidents.
The administrative changes to correct typographical errors on
Tables 3.2.1 and 4.2.1 will have no affect on plant hardware, plant
design, safety limit setting or plant system operation and
therefore, do not modify or add any initiating parameters that would
significantly increase the probability or consequences of any
previously analyzed accident.
Therefore, it is concluded that there is not a significant
increase in the probability or consequence of an accident previously
evaluated.
2. The function of the Core Spray High Sparger Pressure
Instrumentation to be removed from the Technical Specifications is
for local indication and alarm only. These functions are not
necessary for operators to accomplish any safety functions.
The proposed change does not involve any change in hardware,
function, Technical Specification trip setpoints, plant operation,
redundancy, protective function or design basis of the plant. There
is no impact on any existing safety analysis or safety design
limits. Core Spray High Sparger Pressure Instrumentation functions
do not initiate nuclear system parameter variations which are
considered potential initiating causes of threats to the fuel and
the nuclear system process barrier.
As discussed above, the proposed administrative change only
corrects typographical errors concerning equipment identification
numbers. This change doe not affect any equipment and it does not
involve any potential initiating events that would create any new or
different kind of accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change to remove the Core Spray High Sparger
Pressure Instrumentation from the Technical Specifications for ECCS
Actuation Instrumentation does not affect any existing safety
margins. This equipment is NNS and performs a local indication and
alarming function only. The original intent of this detection system
was because the first BWR plants had only the CS System for long-
term core cooling. Later, plants like VY [Vermont Yankee] were
provided with Low Pressure Coolant Injection (LPCI) Systems in
addition to CS.
Existing Technical Specification requirements for automatic trip
functions are unaffected. Failure of the Core Spray High Sparger
Pressure Instrumentation does not preclude the ability of the CS
System to perform its safety function to mitigate the consequences
of accidents or of any other safety system to accomplish its safety
functions. Proper ECCS functioning post-accident is not relied upon
by NNS alarming functions but by such systems as safety related
reactor level indication.
The CS Sparger Piping is inspected every refueling outage to
verify its integrity. No cracks in the CS Sparger piping have been
identified since the first inspection in 1980. The removal from the
Technical Specifications has no affect on the bases of Protective
Instrumentation which is to operate to initiate required system
protective actions. The Core Spray High Sparger Pressure
Instrumentation does not perform any safety function.
As discussed above, the proposed administrative change which
corrects typographical errors does not affect any equipment involved
in potential initiating events or safety limits. [***].
Based upon the above, it is concluded that the proposed changes
do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301
Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,
One International Place, Boston, Massachusetts 02110-2624
NRC Project Director: Walter R. Butler
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: March 18, 1992, modifications submitted
June 25 and July 28, 1992, and December 6, 1993
Description of amendment request: On December 6, 1993, the licensee
submitted a significant modification to its original request of March
18, 1992 (April 19, 1992 (57 FR 12349)). The modified request would
amend the Technical Specifications (TS) to provide two surveillance
tests to determine the operability of the catalyst beds in the
containment atmospheric control (CAC) system. One test compares
hydrogen content in the influent to the hydrogen content in effluent
process streams to assure the catalyst is operating. The second test
measures the temperature profile in the catalyst bed to ensure that
sufficient catalyst remains available for the recombination process
during postulated accident conditions. Since the original proposal has
been significantly changed, the staff is issuing a new notice and
proposed no significant hazards consideration determination.
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 staff's review is
presented below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not involve any changes in the design
or operation of the hydrogen recombiners, which are accident
mitigation systems. The proposed change revises surveillance
requirements to ensure that existing equipment will perform as
designed in response to postulated events. The proposed change does
not, therefore, involve an increase in the probability or
consequences of an accident previously evaluated.
2. Does the amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not involve any changes in the design
or operation of existing equipment, and does not, therefore, create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the amendment involve a significant reduction in a
margin of safety?
The proposed change increases surveillance requirements for
existing equipment to ensure that the equipment will perform as
designed. With equipment performing as designed in response to
postulated accidents, the proposed change does not affect any
existing margins of safety.
Based on the licensee's analysis and the staff's analysis, 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: Nicholas S. Reynolds, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Theodore R. Quay
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: March 31, 1994
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant Technical Specifications (TS)
by incorporating operability and surveillance requirements for the
recently installed Auxiliary Feedwater Pump Low Discharge Pressure Trip
instrumentation. Proposed surveillance requirements would be added to
Table TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations and Test
of Instrument Channels.'' TS 3.4, ``Steam and Power Conversions
System,'' would be revised to explicitly link operability of the
associated Auxiliary Feedwater Pump Low Discharge Pressure Trip channel
to operability of the associated auxiliary feedwater pump. In addition,
minor format inconsistencies in TS 3.4.b.1.A and 3.4.b.1.B would be
corrected.
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) Table TS 4.1-1
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change defines the necessary surveillance
requirements for the recently installed Auxiliary Feedwater Pump Low
Discharge Pressure Trip channels. The intent of adding surveillance
requirements to the TS's is to ensure the availability and
reliability of the components. The proposed change is an additional
restriction not presently included in the TS's. Therefore, it will
not increase the probability or consequences of an accident
previously evaluated in the USAR.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change adds surveillance requirements to the TS for
the Auxiliary Feedwater Low Discharge Pressure Trip channels. It
does not alter the plant configuration or overall plant performance.
Therefore, it does not create the possibility of a new or different
kind of accident.
3. Involve a significant reduction in the margin of safety.
This proposed revision is an additional requirement in the TS's
to ensure the availability and reliability of the Auxiliary
Feedwater Pump Low Discharge Pressure Trip channels. It does not
alter the input or assumptions of the safety analysis, and is an
enhancement from an overall safety standpoint. Therefore, it will
not involve a reduction in the margin of safety.
(b) TS 3.4
The proposed changes were reviewed in accordance with the
provision of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change defines the necessary operability
requirements for the recently installed Auxiliary Feedwater Pump Low
Discharge Pressure Trip channels. Installation of this protection
was recommended and approved by the NRC prior to their installation.
The proposed change requires that the reactor not be heated
350 deg.F unless both motor driven Auxiliary Feedwater
Pumps and their associated low discharge pressure trip channels are
operable. Also, the reactor shall not be heated
350 deg.F unless the turbine driven auxiliary feedwater pump and its
associated low discharge pressure trip channel are operable, or if
not demonstrated operable prior to 350 deg.F, they
shall be declared inoperable when 350 deg.F is exceeded.
Furthermore, when the reactor is 350 deg.F, an auxiliary
feedwater pump low discharge pressure trip channel may be inoperable
for a period not to exceed 4 hours. If this time is exceeded, the
associated auxiliary feedwater pump shall be declared inoperable and
the appropriate limiting condition for operation of TS 3.4.b.2
entered. The intent of adding these operability requirements to the
TS's is to ensure the availability of the components. The proposed
change is an additional restriction not presently included in the
TS's. Therefore, it will not increase the probability or
consequences of an accident previously evaluated in the USAR.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change adds operability requirements to the TS for
the Auxiliary Feedwater Pump Low Discharge Pressure Trip channels.
It does not alter the plant configuration or overall plant
performance. Therefore, it does not create the possibility of a new
or different kind of accident.
3. Involve a significant reduction in the margin of safety.
This proposed revision is an additional requirement in the TS's
to ensure the operability of the Auxiliary Feedwater Pump Low
Discharge Pressure Trip channels. It does not alter the input or
assumptions of the safety analysis, and is an enhancement from an
overall safety standpoint. Therefore it will not involve a reduction
in the margin of safety
(c) Administrative changes to TS 3.4.b.1.A and TS 3.4.b.1.B
The proposed changes were reviewed in accordance with the
provision of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will 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 an accident previously evaluated, or
3. Involve a significant reduction in the margin of safety.
The proposed changes are administrative in nature and do not
alter the intent of interpretation of the TS. Therefore, no
significant hazards exist.
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 Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: John N. Hannon
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: May 24, 1994
Description of amendment request: The amendment request proposes to
revise the Technical Specifications (TS) to implement the NRC's Final
Policy Statement on Technical Specification Improvements for Nuclear
Power Reactors (58 FR 39132). These improvements involve focusing the
Technical Specifications on those requirements that are of controlling
importance to operational safety by screening each TS in Section 3/4.1
through 3/4.11 using the criteria provided in the policy statement. The
purpose of the proposed amendment request is to relocate the
specifications that do not meet any of the four policy statement
criteria. The relocated specifications will be moved to Updated Final
Safety Analysis (USAR) Chapter 16. Based on the screening, all or part
of 38 technical specifications were identified as not meeting any of
the criteria and, therefore, as candidates for relocation. The licensee
has categorized the TS changes as (1) specifications relocated intact
to USAR Chapter 16, (2) specifications relocated with portions retained
in TS, (3) specifications relocated with programmatic requirements
referenced in Section 6 of TS, (4) modifications to retained
specifications to accomodate relocation of other specifications, and
(5) new specification requirements incorporated into the TS. The last
category is used to effect the retention of portions of relocated
specifications and accomodate the policy statement recommendation to
incorporate industry experience in the determination of TS content.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed Technical Specification changes involve relocating
requirements that are not conditions or limitations on reactor
operation necessary to obviate the possibility of an abnormal
situation or event giving rise to an immediate threat to the public
health and safety. The proposed changes were identified through the
application of criteria designed to cull those requirements that are
not important to operational safety from the Technical
Specifications. In this process, selected provisions of the
Technical Specifications identified for relocation were retained if
necessary to support a Technical Specification that was to be
retained. Thus, only specification requirements that have little or
no operational safety significance are proposed for relocation. In
addition, those requirements that would be relocated will be
included in the Updated Final Safety Analysis Report (USAR) and,
therefore, will be controlled and implemented as NRC commitments. In
this manner, those requirements that have no operational safety
significance but involve maintaining the plant in its as-designed
state (for example, through surveillance programs) would be
controlled.
In addition, the criteria for identifying requirements to be
retained in the Technical Specifications specifically call out, for
retention, those structures, systems, or components that are
required to mitigate accidents previously evaluated.
Based on the above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed changes involve relocating Technical Specification
requirements to another licensee-controlled document. No changes or
physical alterations of the plant are involved. Also, no changes to
the operation of the plant or equipment are involved. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety.
The proposed changes involve relocating Technical Specification
requirements to the USAR. The requirements to be relocated were
identified by applying the criteria endorsed in the Commission's
Policy Statement. Thus, those specifications that would be relocated
do not impose constraints on design and operation of the plant that
are derived form the plant safety analysis report or from
probabilistic safety assessment (PSA) information and do not belong
in the Technical Specifications in accordance with 10 CFR 50.36 and
the purpose of the Technical Specifications stated in the Policy
Statement. Therefore, relocation of these requirements does not
involve a significant reduction in the margin of safety.
In addition, revisions to the USAR will be evaluated in
accordance with the 10 CFR 50.59 process which considers the
reduction in safety margin. Therefore, any future revisions to the
provisions in the USAR will consider reductions in the margin of
safety using the criteria for identifying an unreviewed safety
question.
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, DC 20037
NRC Project Director: Theodore T. Quay
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: June 7, 1994
Description of amendment request: The proposed amendment revises
Technical Specification Table 2.2-1, Reactor Trip System
Instrumentation Setpoints, to change the over-temperature-delta-
temperature (OTDT) axial flux difference (AFD) limits to reflect
results of the Cycle 8 core maneuvering analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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
change. Operation at 3565 MWt does not affect any of the mechanisms
postulated in the USAR to cause LOCA or non-LOCA design basis
events. Analyses, evaluations and minimum DNBR [departure from
nucleate boiling ratio] calculations confirm that the USAR
conclusions remain valid for the proposed changes. On these bases 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 possiblity of a new
or different kind of accident from any previously evaluated.
There is no new type of accident or malfunction being created.
The proposed change provides revised operating limits necessary to
support Cycle 8, and does not change the method and manner of plant
operation. The safety design bases in the USAR have not been
altered. Thus, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in the 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
demonstrate that all applicable safety analysis acceptance criteria
continue to be met for the proposed operating conditions. Items not
specifically cited in this safety evaluation have been reviewed and
have been found to be bounded by the evaluations performed for
Reference 1 [Wolf Creek Generating Station Technical
Specifications]. 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, DC 20037
NRC Project Director: Theodore. R. Quay
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenberg County, North Carolina
Date of amendment request: May 5, 1994, as supplemented June 16,
1994.
Description of amendment request: The proposed amendments would
change the Technical Specifications to increase Main Steam and
Pressurizer Code Safety Valve Setpoint Tolerances.
Date of publication of individual notice in Federal Register: June
21, 1994 (59 FR 32029).
Expiration date of individual notice: July 21, 1994
Local Public Document Room location:
Atkins Library, University of North Carolina, Charlotte (UNCC
Station), North Carolina 28223.
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
rooms for the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: September 29, 1992, as
supplemented on October 22, 1993, and November 11, 1993.
Brief description of amendments: The amendments revise the Site
Boundary Map and the Low Population Zone Map.
Date of issuance: June 22, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 190 and 167
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 28, 1992 (57 FR
48813) The Commission prepared an Environmental Assessment and Finding
of No Significant Impact which was published in the Federal Register on
May 13, 1994 (59 FR 25129). The Commission's related evaluation of
these amendments is contained in a Safety Evaluation dated June 22,
1994.No significant hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station,Plymouth County, Massachusetts
Date of application for amendment: February 11, 1993, as
supplemented December 2, 1993, January 5, February 22, March 1, April
15, and May 16, 1994.
Brief description of amendment: This amendment increases the
allowed fuel assembly storage cells from 2320 to 3859, changes the
maximum loads allowed to travel over the spent fuel assemblies from
1050 to 2000 lbs., and changes the limiting characteristics of
assemblies to be stored in the spent fuel from a maximum KINIFITY
less than or equal to 1.35 to a Maximum KINIFITY less than or
equal to 1.32 and a maximum lattice average uranium enrichment of less
than or equal to 4.6% by weight.
Date of issuance: June 22, 1994
Effective date: June 22, 1994
Amendment No.: 155
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 30, 1993 (58 FR
26171) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 22, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, IllinoisDocket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1
and 2, Rock Island County, IllinoisDate of application for
amendments: March 26, 1993
Brief description of amendments: The amendments revise Technical
Specification 3/4.6 for Dresden and Quad Cities Stations to allow
Single Loop Operation (SLO) with the recirculation loop suction and
discharge valves open. The amendments also delete outdated and
unnecessary portions of Technical Specification 3.6.H for Dresden Units
2 and 3 and provide more consistency to the BWR Standard Technical
Specifications (NUREG-0123, Revision 4).
Date of issuance: June 16, 1994
Effective date: June 16, 1994
Amendment Nos.: 127, 121, 147, and 143
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17594) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 16, 1994. No significant
hazards consideration comments received: No
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.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South CarolinaDate of
application for amendments: March 24, 1994, as supplemented April
11 and May 31, 1994
Brief description of amendments: The amendments revise the
Technical Specification (TS) to increase boron concentration for the
spent fuel storage pool during Modes 1-3 operation and for the
refueling canal during Mode 6 operation; include two reload related
topical reports in TS 6.9.1.9; and correct errors in nomenclature and
remove obsolete footnotes.
Date of issuance: June 13, 1994
Effective date: June 13, 1994
Amendment Nos.: 120 and 114
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22006) The April 11 and May 31, 1994, letters provided clarifying and
additional information that did not change the scope of the March 24,
1994, application and the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated June 13, 1994.No
significant hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
ConnecticutDate of application for amendment: December 17, 1993, as
supplemented April 12, 1994.
Brief description of amendment: The amendment changes the action
statements for the limiting conditions for operation associated with
the electrical power sources (Technical Specification 3.8.1.1).
Date of issuance: June 14, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 177
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994(59 FR
4943). The April 12, 1994, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated June 14, 1994.No
significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New YorkDate of
application for amendment: December 28, 1993
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 6.9(A)1.a. to permit startup reports for
cycles subsequent to the initial fuel cycle to address only those
startup tests that are actually performed. The amendment also revises
TS Section 6.9(A) to clarify requirements for the submission of routine
reports. These changes are consistent with the guidance provided in
NUREG-1433, ``Standard Technical Specifications - General Electric
Plants, BWR/4.''
Date of issuance: June 16, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 212
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4945) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 16, 1994.No significant hazards
consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: December 29, 1993
Brief description of amendment: The amendment revises Appendix B of
the Technical Specifications (TSs), the Radiological Effluent TSs.
Specifically, the amendment revises Appendix B Surveillance Requirement
3.1.a. and Table 3.10-2 to provide surveillance requirements for data
recorders associated with the gaseous effluent monitoring system. The
amendment also makes an editorial change to Appendix B Limiting
Condition for Operation 3.1.a. to improve consistency and clarity.
Date of issuance: June 16, 1994Effective date: As of the date of
issuance to be implemented within 30 days.
Amendment No.: 213
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 2, 1994 (59 FR
4946) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 16, 1994.No significant hazards
consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama.
Date of amendments request: May 13, 1991, as supplemented October
13, 1992.
Brief description of amendments: The amendments modify the TS for
the overpressure protection systems. The allowable outage time (AOT)
for one inoperable residual heat removal (RHR) relief valve with one or
more of the reactor coolant system cold leg temperatures less than or
equal to 310 degrees Fahrenheit is being decreased from 7 days to 24
hours for water-solid conditions. The required AOT for low temperature
conditions, other than water-solid, will remain at 7 days with one RHR
relief valve inoperable, provided the pressurizer level is less than or
equal to 30 percent and a dedicated operator is assigned to monitor and
control the reactor coolant system pressure.
Date of issuance: June 16, 1994
Effective date: June 16, 1994
Amendment Nos.: 108 and 100
Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: July 22, 1992 (57 FR
32577) and February 17, 1993 (58 FR 8787)The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
June 16, 1994.No significant hazards consideration comments received:
No
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns
Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama
Date of application for amendment: April 1, 1992 (TS 302)
Brief description of amendments: The amendments add requirements to
the Browns Ferry Units 1 and 3 Technical Specifications to provide
administrative controls for a post-accident sampling system, which were
requested by Generic Letter 83-36, ``NUREG-0737 Technical
Specifications.''
Date of issuance: June 21, 1994
Effective date: June 21, 1994
Amendment Nos.:207 and 180
Facility Operating License Nos. DPR-33 and DPR-68:
Date of initial notice in Federal Register: May 27, 1992 (57 FR
22269) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 21, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
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: December 23, 1992, as
supplemented on March 18, 1994.
Brief description of amendment: This amendment revises TS 3/4 3.3.5
for transfer switches used to meet 10 CFR Part 50, Appendix R (Fire
Protection) requirements, and specifies a new special report
requirement for TS 6.9.2.
Date of issuance: June 14, 1994
Effective date: June 14, 1994
Amendment No. 187
Facility Operating License No. NPF-3. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10016) The supplemental information submitted on March 18, 1994, did
not change the initial proposed finding of no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated June 14, 1994.No
significant hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: December 20, 1993, as amended
March 25 and April 25, 1994
Brief description of amendment: The amendment modifies the
technical specifications (TS) to address new containment purge and vent
valves to be installed in the 1994 refueling outage. The amendment
changes the containment purge and vent valve TS as follows: (1) removes
the requirement ensuring that valve position remains at less than or
equal to 70 degrees, (2) changes the containment leak testing
requirements for the metal-to-metal seated valves from 6 months to 2
years since they have improved seat designs, and (3) makes
administrative changes to delete an out-of-date note, to relocate an
action statement requirement from the TS surveillance section to the TS
action statement section, and to change a related TS reference to this
surveillance section. Valve opening position does not need to be
limited to less than or equal to 70 degrees. The resiliently-seated
valves have a permanently installed mechanical stop to limit the open
position to ensure adequate closure times. The metal-to-metal seated
valves are designed to close from the 90-degree open position.
Date of issuance: June 15, 1994
Effective date: 15 days from the date of issuance
Amendment No.: 124
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14901) The additional information contained in the March 25 and April
25, 1994, letters was clarifying in nature, is within the scope of the
initial notice, and did not affect the NRC staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated June 15, 1994.Public comments on proposed no significant hazards
consideration comments received: No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: February 8, 1994, as
supplemented March 25, 1994
Brief description of amendment: The amendment revises the WNP-2
Technical Specifications. Specifically, the amendment increases the
stroke time, as specified in Table 3.6.3-1, for reactor core isolation
cooling (RCIC) valve RCIC-V-8 from 13 seconds to 26 seconds and deletes
the Note (j) reference from RCIC-V-8 and RCIC-V-63. Note (j) indicates
that the stroke time specified in the table reflects the requirement
for containment isolation only.
Date of issuance: June 17, 1994
Effective date: June 17, 1994
Amendment No.: 125
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1994 (59 FR
24754) The additional information contained in the March 25, 1994,
letter was clarifying in nature, was within the scope of the initial
notice, and did not affect the NRC staff's proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
June 17, 1994.No significant hazards consideration comments received:
No.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352]
Dated at Rockville, Maryland, this 28th day of June 1994.
FOR THE NUCLEAR REGULATORY COMMISSION
Jack W. Roe,
Director, Division of Reactor Projects - III/IVOffice of Nuclear
Reactor Regulation
[Doc 94-16174 Filed 7-5-94 8:45 am]
BILLING CODE 7590-01-F
