[Federal Register Volume 59, Number 91 (Thursday, May 12, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-11226]
[[Page Unknown]]
[Federal Register: May 12, 1994]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 18, 1994, through April 29, 1994. The
last biweekly notice was published on April 28, 1994 (59 FR 22000).
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 June 10, 1994, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
room for the particular facility involved.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of amendment request: March 30, 1994.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.4.9, Pressure/Temperature
Limits, and its associated Bases, by changing the Unit 1 heatup and
cooldown curves to incorporate a newly determined reactor vessel
reference nil-ductility temperature, RTNDT, and by updating the
removal schedule of vessel surveillance capsules for both units in
accordance with ASTM E185-82. Changes would also be made to the Unit 1
Low Temperature Overpressure Protection System (LTOPS) setpoint curve
in TS 3.4.9.3 to reflect the new pressure/temperature limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The use of new pressure-temperature limit curves and low
temperature overpressure protection curves will not change any
postulated accident scenarios. The revised curves were developed using
industry standards and regulations which are recognized as being
inherently conservative. The pressure-temperature low temperature
overpressure curves provide reactor coolant system (RCS) limits to
protect the reactor pressure vessel from brittle fracture by clearly
separating the region of normal operations from the region where the
vessel is subject to brittle fracture. The heatup and cooldown limits
are designed to ensure that the 10 CFR 50 Appendix G Pressure
Temperature limits for the RCS are not exceeded during any condition of
normal operation including anticipated operational occurrences.
General Design Criterion 32 of 10 CFR 50 Appendix A requires that
the reactor coolant boundary shall be designed with sufficient margin
to assure that when stressed under operating, maintenance, testing, and
postulated accident condition[s], (1) the boundary behaves in a
nonbrittle manner and (2) the probability of rapidly propagating
fracture is minimized.
10 CFR 50 Appendix G, ``Fracture Toughness Requirements,'' requires
that the effects of changes in the fracture toughness of reactor vessel
materials caused by neutron radiation throughout the service life of
[a] nuclear reactor be considered in the pressure-temperature limits.
The change is used in conjunction with the material initial reference
temperature (RTNDT) to establish the limiting pressure-temperature
curves. Regulatory Guide 1.99, Rev. 2, contains procedures for
calculating the effects of neutron radiation embrittlement of the low-
alloy steels currently used for light-water-cooled reactor vessels.
Using the Regulatory Guide 1.99, Revision 2, Braidwood Unit 1
Surveillance Capsule U results, and Appendix G to 10 CFR 50, new
Pressure-Temperature curves [were] prepared for the projected reactor
vessel exposure at 32 EFPY of operation. These new curves, in
conjunction with the heatup and cooldown ranges and the revised Low-
Temperature Overpressure Protection System setpoints, provide the
required assurance that the reactor pressure vessel is protected from
brittle fracture up 32 EFPY of operation. No changes to the design of
the facility have been made and no new equipment has been added or
removed. The revised analysis and resultant adjustment of the operating
limitations provide assurance that the Reactor Coolant System is
protected from brittle fracture.
Revising the Reactor Vessel Material Surveillance Program
Withdrawal Schedule does not result in the addition or removal of any
equipment, or any design changes to the facility. Capsule lead times
are revised and, for Braidwood Unit 2, Capsule X will be removed next
vice Capsule W. The proposed removal schedules remain consistent with
ASTM 185-82.
Therefore, the proposed amendment to the pressure temperature
limitations does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
B. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The use of the new pressure-temperature operating limits and the
new low temperature overpressure protection curve does not change any
postulated accident scenarios. The new curves do not represent any
appreciable change in the current methodologies; they merely provide
assurance that the Reactor Coolant System is protected from brittle
fracture. No new accident or malfunction mechanism is introduced by the
amendment and no physical plant changes will result from this
amendment.
Revision of the Reactor Vessel Material Surveillance Program
Withdrawal Schedule does not introduce a new accident or malfunction
mechanism. Capsule lead times are revised, and, other than changing the
order of specimen removal, consistent with ASME 185-82, no physical
plant changes will result from this revised schedule.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
C. The proposed change does not involve a significant reduction in
a margin of safety.
The new pressure-temperature operating limits low temperature
overpressure protection curves were generated with the currently
accepted conservative methodology using capsule surveillance data. The
new pressure-temperature curves were developed using industry standards
and regulations (ASME Code Section III, and NRC Regulatory Guide 1.99,
Revision 2) which are recognized as being inherently conservative. The
use of the new pressure- temperature operating limits and low
temperature overpressure protection limits would not change postulated
accident scenarios.
The proposed revision to the Reactor Vessel Material Surveillance
Program Withdrawal Schedule would not change postulated accident
scenarios. Capsule lead times are revised, and, other than changing the
order of specimen removal, consistent with ASTM 185- 82, no physical
plant changes will result from this revised schedule. 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: 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: James E. Dyer.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: March 31, 1994.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) to provide allowable outage
times for automatic actuation channel surveillance testing and
restoration time for an inoperable engineered safety feature actuation
system automatic actuation channel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following evaluation is provided for the three categories of
the significant hazards consideration standards:
a. Proposed changes to allow 8 hours for master relay and logic
testing, 12 hours for slave relay testing and 6 hours to restore an
inoperable ESFAS Automatic Actuation Channel prior to entering the
shutdown action clock.
(1) The determination that these changes are within all acceptable
criteria was established in the NRC's SER prepared for WCAP-10271,
Supplement 2, Revision 1. The Technical Specification changes proposed
by this license amendment request conform to NRC guidance contained in
the SER. The NRC found that implementation of the proposed changes is
expected to result in a small and acceptable increase in ESFAS
unavailability. This increase in probability results in a small
increase in calculated core damage frequency and public risk. The
calculated increase in core damage frequency was judged to be
acceptable since the increase was small and well within the range of
uncertainty associated with the analysis. The values presented in WCAP-
10271 Supplement 2 Revision 1 for increase in core damage frequency
were verified by Brookhaven National Laboratory as part of an audit and
sensitivity analyses performed for the NRC Staff.
Based on the small increase in core damage frequency as compared
with the range of uncertainty in the analysis, the NRC agreed that the
calculated increase is acceptable. This conclusion was documented in
the NRC's SER dated February 22, 1989. The applicability of these
conclusions has been verified through a plant specific review of the
generic analysis in WCAP-10271, Supplement 2, Revision 1. The ESFAS
Automatic Actuation Channel allowed outage and restoration times
included in this license amendment request are consistent with the
generic analysis. In addition, the NRC stated that the majority of the
increase in unavailability was due to the decrease in frequency of
surveillance testing vice the changes in allowed outage and restoration
times. Therefore, considering the above information, the proposed
allowed outage and restoration time changes do not involve a
significant increase in the probability of occurrence or consequences
of an accident previously evaluated.
(2) The proposed changes do not involve the physical alteration of
any plant system and do not result in a change in the manner in which
the ESFAS system performs its function. The increases in allowed outage
and restoration times only affects the probability of the ESFAS
Automatic Actuation Channel functioning properly as described above.
Therefore, the allowed outage and restoration time changes proposed in
this license amendment request do not create a new or different type of
accident from any previously evaluated.
(3) The proposed allowed outage time and restoration time changes
do not alter the manner in which safety limits, limiting safety system
setpoints or limiting conditions for operation are determined. The
impact of the revised ESFAS Automatic Actuation Channel allowed outage
and restoration times is addressed above. Implementation of the
proposed changes is expected to result in an overall improvement in
safety by allowing adequate time for required ESFAS testing and quality
repairs leading to improved equipment reliability due to a more
appropriate restoration time. Therefore, it may be concluded that the
proposed allowed outage and restoration time changes do not involve a
significant reduction in margin of safety.
b. Proposed change to the minimum required degree of redundancy for
the High-High Containment Pressure channels in Table 3.4-1.
(1) Changing the minimum required degree of redundancy in Table
3.4-1 for the High-High Containment Pressure Channels (Table 3.4-1
items II.3, III.B.3, and IV.3) provides consistency with Technical
Specification 3.4.2.c which allows an inoperable High-High Containment
Pressure channel to be placed in bypass. Placement of an inoperable
High-High Containment Pressure Channel in bypass is preferred to reduce
the probability of an inadvertent containment spray event. Also, these
channels are designed with a two out of four logic so that the failed
channel may be bypassed rather than tripped. With the failed channel
bypassed, single failure criterion is still met because the logic is
now a two out of three. Furthermore, with the one channel bypassed, a
single channel failure will not inadvertently initiate a containment
spray. Therefore, this change can be considered an administrative
change to correct Table 3.4-1 to agree with the Action requirements of
Technical Specification 3.4.2.c. As such this proposed change does not
involve an increase in the probability of occurrence or consequences of
an accident previously evaluated.
(2) Correcting the minimum required degree of redundancy in Table
3.4-1 for the High-High Containment Pressure channels is an
administrative change which does not involve the physical alteration of
any plant system and does not result in a change in the manner in which
the ESFAS system performs its function. Therefore, the proposed
correction to Table 3.4-1 does not create the possibility of a new or
different kind of accident from any previously analyzed.
(3) Correcting the minimum required degree of redundancy in Table
3.4-1 to be consistent with the Actions of Technical Specification
3.4.2.c is an administrative change and as such does not involve any
reduction in a margin of safety.
c. Proposed change to the delete footnote +++ from Table 3.4-1.
(1) Deleting footnote +++ from Table 3.4-1 removes the
inconsistency between it and Technical Specification 3.4.2.c which
states that channels other than the High-High Containment Pressure
channels shall be placed in trip during testing. The change does not
affect the manner in which ESFAS provides plant protection. In addition
the change does not affect the functioning of ESFAS or the way Zion
Station conducts channel testing. Instrument channel testing will
continue to be conducted in the tripped mode with the exception of the
High-High Containment Pressure channels, which can be tested in bypass
because of the risk of a spurious Containment Spray event. Automatic
Actuation Channel testing will be performed in accordance with the
allowed outage times of new Specification 3.4.2.d. As such this
proposed change does not involve any significant increase in the
probability of occurrence or consequences of an accident previously
evaluated.
(2) Deleting footnote +++ from Table 3.4-1 does not involve the
physical alteration of any plant system and does not result in a change
in the manner in which ESFAS performs its function. Therefore this
change does not involve the physical alteration of any plant system and
does not result in a change in the manner in which the ESFAS system
performs its function. Therefore, the proposed correction to Table 3.4-
1 does not create the possibility of a new or different kind of
accident from any previously analyzed.
(3) Deleting footnote +++ from Table 3.4-1 does not alter the
manner in which safety limits, limiting safety system setpoints or
limiting conditions for operation are determined. Implementation of
this change will not alter ESFAS testing. Therefore implementation of
this change does not involve any reduction in a margin of safety.
d. Proposed editorial change to Technical Specification 3.4.2.c.
The editorial change to Technical Specification 3.4.2.c to change
``Containment Hi-Hi pressure channels'' to ``High-High Containment
Pressure channels'' is purely an administrative change which has no
affect on plant safety.
e. Summary.
The foregoing analyses demonstrate that the proposed License
Amendment to the Zion Station Technical Specifications does not involve
a significant increase in the probability of occurrence or consequences
of a previously evaluated accident, does not create the possibility of
a new or different kind of accident and 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: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690.
NRC Project Director: James E. Dyer.
Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of amendment request: April 22, 1994.
Description of amendment request: The proposed change revises the
reactor vessel pressure-temperature limits in the Technical
Specifications. The change insures that the vessel fracture toughness
requirements of Section V of 10 CFR Part 50, Appendix G, are satisfied
through the end of life.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The margin above Nil Ductility Transition Temperature (NDTT) is
governed by 10 CFR 50 Appendix G and remains unchanged. The proposed
change will not involve a significant increase in the probability or
consequences of a previously evaluated accident.
2. Will the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The predicted shifts in NDTT are based on a revised reference
temperature consistent with Regulatory Guide 1.99, Revision 2, dated
May 1988. This method of revising temperature-pressure limits is the
same as in the past (ASME Code Section III, Appendix G).
3. Will the proposed change involve a significant reduction in the
margin of safety?
The proposed curves were generated for an End of Licensed Life (May
31, 2000) Effective Full Power Year exposure and are conservative in
nature until that time. The margin of safety [is] unchanged.
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: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: L. B. Marsh.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: April 18, 1994.
Description of amendment request: License Amendment No. 81, issued
on July 15, 1993, changed the numbering of surveillance requirements
for Technical Specifications 3/4.3.1, ``Control Rod Operability,'' 3/
4.3.2, ``Control Rod Maximum Scram,'' and 3/4.10.2, ``Rod Pattern
Control System.'' However, Action Statements referencing these
surveillance requirements were overlooked and were not appropriately
renumbered. The purpose of the proposed technical specification change
would be to renumber the overlooked references.
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) These changes do not affect the intent or implementation of the
applicable Technical Specifications. The changes simply make the
affected Technical Specifications consistent. Since these are only
editorial changes which do not impact the plant design or operations,
they cannot increase the probability or the consequences of any
accident previously evaluated.
(2) The proposed changes are editorial only and do not affect the
plant design or operation. No new failure modes are introduced by such
changes and, therefore, the request will not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
(3) The proposed changes merely correct an editorial oversight.
These changes do not alter or delete any technical requirements and,
therefore, do not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606.
NRC Project Director: John N. Hannon.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket No.
50-461, Clinton Power Station, Unit No. 1, DeWitt County, Illinois
Date of amendment request: April 18, 1994.
Description of amendment request: Test methods for carbon adsorber
filters specified in Technical Specification Sections 3/4.6.6.3,
``Standby Gas Treatment System,'' and 3/4.7.2, ``Control Room
Ventilation System,'' specify the 1979 version of ASTM D3803. The
proposed change would delete the year of the standard so that more
recent versions of the standard could be used.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
(1) The proposed changes to the Technical Specification
surveillance requirements for determining the methyl iodide penetration
of carbon samples would not involve a significant increase in the
probability or the consequences of any accident previously evaluated
because the proposed change merely allows Illinois Power (IP) to
utilize a more up-to-date version of the same test method currently
specified. More recent versions of the test method are more effective
at detecting unsatisfactory charcoal performance because they include
equilibration periods to ensure that all samples have a common starting
point before being challenged with radioactive gas. The proposed change
would not affect the quality of the charcoal or the reliability of the
filter subsystems as it only relates to testing and involves no changes
to the design or operation of the ventilation subsystems themselves.
The updated standards provide more accurate and repeatable test results
and do not change the properties or acceptance criteria for these
properties. As a result, the performance capabilities of the associated
filter subsystems would not be adversely impacted by the proposed
change.
(2) The proposed change would not involve a change in the design or
operation of any plant system or component. In addition, the proposed
change would not reduce the level of filter train subsystem reliability
nor would it create an initiating event for any accident. Because the
performance, function, and redundancy of the original design remain
unchanged, the proposed change would not create the potential for a new
event. Furthermore, since no new types of equipment would be introduced
into the plant design and the proposed change would not adversely
impact existing equipment, no potential for a different type of
malfunction is created by the proposed change. Therefore, this proposed
change cannot create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) The margin of safety for the charcoal filter subsystems as
defined in the Bases to the Technical Specifications associated with
the proposed change refers to the ability of the filters to remove
radioiodines. The proposed change would allow IP to upgrade the
currently specified test for determining charcoal adsorber performance
with one which utilizes the same type of methodology, but provides
greater accuracy and repeatability. The newer versions of the test
method are more effective at detecting unsatisfactory charcoal
performance because they include equilibration periods to ensure that
all samples have a common starting point before being challenged with
radioactive gas. Thus, the proposed change would not involve a
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606.
NRC Project Director: John N. Hannon.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: March 28, 1994.
Description of amendment request: The proposed amendment would
revise technical specifications Tables 3.2.4 and 4.2.1, to change one
of the initiating parameters of the reactor building ventilation
isolation system and standby gas treatment system (SGTS) from Low
Reactor Water Level to Low Low Reactor Water Level. This revision is
being made in order to improve plant performance by reducing the
potential for unnecessary secondary containment isolation and SGTS
initiations.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed amendment will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The function of the Standby Gas Treatment System and secondary
containment is to mitigate the consequences of a loss of coolant
accident and fuel handling accidents. The proposed changes maintain
this capability. The revised Standby Gas Treatment System initiation
and secondary containment isolation parameter of low low reactor water
level provides the required detection of loss of coolant accidents and
is consistent [with] ECCS actuation to mitigate the consequences of
this accident. The low low reactor water level instrumentation is set
to trip when reactor water level is 6'6'' above the top of the active
fuel. This trip currently initiates closure of the Group 1 Primary
containment isolation valves, activates the Emergency Core Cooling
systems and starts the emergency diesel generator. This trip setting
level was chosen to be low enough to prevent spurious operation but
high enough to initiate Emergency Core Cooling system operation and
primary system isolation so that no melting of the fuel cladding will
occur, post accident cooling can be accomplished, and the guidelines of
the 10 CFR 100 will not be violated. Therefore, this amendment will not
cause a significant increase in the probability or consequences of an
accident previously evaluated for the Monticello plant.
The proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously analyzed. The
proposed changes to Technical Specifications for the standby gas
treatment system and secondary containment do not alter the function 8
of the systems or its interrelationships with other systems. An adverse
interaction which could be postulated to occur is the initiation of the
Standby Gas Treatment System without a coordinated trip of the
Mechanical Vacuum Pump. The Mechanical Vacuum Pump is operated during
plant startups to draw a vacuum on the main condenser prior to
admission of steam. The Mechanical vacuum discharges to the offgas
stack and thus can create a back pressure on the Standby Gas Treatment
System, reducing initiation of Standby Gas Treatment System flow below
required values. The proposed initiation of Standby Gas Treatment
System on low low reactor water level maintains the necessary
coordination by having the Standby Gas Treatment System initiate
subsequent to isolation or tripping of the Mechanical Vacuum Pump on a
low reactor water level signal from the primary containment isolation
logic. Therefore, this amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The proposed amendment will not involve a significant reduction in
the margin of safety.
The proposed amendment changes the initiation of the Standby Gas
Treatment System and secondary containment isolation from being
concurrent with the low reactor water signal (which is indicative that
the reactor core is in danger of being inadequately cooled) to being
concurrent with reactor low low water level (which is also an indicator
that the capability to cool the core is threatened and assures that no
melting of the fuel cladding will occur, post accident cooling can be
accomplished, and the guidelines of 10 CFR 100 will not be violated). A
review of the accident analyses provided in Section 14 of the USAR has
determined that these analyses did not specifically credit initiation
of the Standby Gas Treatment Systems and secondary containment
isolation at the accident precursor reactor water level of low level.
Furthermore, this review determined that the low low reactor water
level setpoint has no adverse impact on the ability of the Standby Gas
treatment System and secondary containment to perform its design basis
function as credited in the accident analyses.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Ledyard B. Marsh.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: March 24, 1994.
Description of amendment request: The proposed modification to
Technical Specification (TS) Sections 3.11.1.4, 6.9.1.8, and 6.14.1
would change the frequency for submitting the Semiannual Radioactive
Effluent Release Report to the NRC from semiannually to annually.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed TS changes are administrative in nature. The proposed
changes simply involve revising the frequency for submitting the
Semiannual Radioactive Effluent Release Report to the NRC from
semiannually to annually in order to implement the amended reporting
requirements of 10 CFR 50.36a as promulgated in Final Rule 57 FR 39353.
Since the information contained in this report is reviewed and
evaluated after the effluents are released, no accidents previously
evaluated are impacted by the proposed TS changes. Radiological
effluent releases from the station will continue to be controlled as
required by the TS, including those requirements specified in the
Offsite Dose Calculation Manual (ODCM) and Process Control Program
(PCP). The proposed TS changes do not involve any changes to the
operation or physical configuration of any plant systems or equipment.
The proposed changes do not impact any initial or final accident
conditions or assumptions previously evaluated. The radiological
consequences of these previously evaluated accidents are not affected
by the proposed changes.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed TS changes are administrative in nature. The proposed
changes simply involve revising the frequency for submitting the
Semiannual Radioactive Effluent Release Report to the NRC from
semiannually to annually in order to implement the amended reporting
requirements of 10 CFR 50.36a as promulgated in Final Rule 57 FR 39353.
Radiological effluent releases from the station will continue to be
controlled as required by the TS, including those requirements
specified in the ODCM and PCP. The proposed TS changes do not involve
any modifications to plant systems or equipment.
Therefore, the proposed TS changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed TS changes do not involve a significant reduction
in a margin of safety.
The proposed TS changes are administrative in nature, and will only
involve revising the frequency for submitting the Semiannual
Radioactive Effluent Release Report to the NRC from semiannually to
annually as currently stipulated in 10 CFR 50.36a. The specific
radiological effluent release information contained in this report will
continue to be provided as required. The station radiological effluent
releases will continue to be controlled as required by TS, including
those requirements specified in the ODCM and PCP.
Therefore, the proposed TS changes do not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101.
NRC Project Director: Charles L. Miller.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of amendment request: February 25, 1994.
Description of amendment request: The proposed Technical
Specification (TS) changes would permit the submittal of the
Radioactive Effluents Release Report on an annual rather than a
semiannual basis; allow changes to the Offsite Dose Calculation Manual
(ODCM) to be submitted in the Radioactive Effluent Release Report
rather than in the monthly operating report; remove the title of
Executive Vice President--Operations from the TS; remove the list of
audit frequencies from the TS and place them under Quality Systems
management; change the title of Associate Manager, Health Physics to
Radiation Protection Manager; remove references to specific letters;
remove TS 6.4 on training; and correct various typographical errors.
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 does not involve a significant increase
in the probability or consequences of an accident previously evaluated
because the administrative change does not affect plant operations in
any manner.
2. The proposed amendment does not create the possibility of a new
or different kind of accident than previously evaluated because the
proposed change is administrative in nature and no physical alterations
of plant configuration or changes to setpoints or operating parameters
are proposed.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety. The change is only administrative.
The NRC staff has reviewed the licensee's analysis and based on
this review it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Fairfield County Library,
Garden and Washington Streets, Winnsboro, South Carolina 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Project Director: William H. Bateman.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of amendment request: March 11, 1994.
Description of amendment request: The proposed Technical
Specification (TS) changes would delete surveillance requirement
4.8.1.4.a.3, which requires periodic testing of penetration protection
fuses, and its associated Basis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The probability or consequences of an accident previously
evaluated is not significantly increased.
Fuses are simple protection devices and can only degrade by being
more resistive which is in the conservative direction. The proper type
and size fuse is assured as part of design, procurement, and initial
installation. The testing provides no additional assurance of
operability. Therefore, the deletion of periodic retesting of these
fuses will not increase the probability or consequences of an accident
previously evaluated.
2. [The proposed amendment will not] [c]reate the possibility of a
new or different kind of accident from any previously analyzed.
The design of the penetration protection and the installation of
the fuses has not changed in any way. Any undetected failure of a fuse
would fall under single failure criteria. A current limiting fuse must
have high electrical current in order to perform its intended function.
Any fuse which has opened the circuit through the penetration would be
detected. (This is not a concern of the Technical Specifications.)
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety. Deletion
of this surveillance requirement will not minimize the intent of this
Technical Specification. This TS is to assure continued operability of
the containment penetration conductor overcurrent protection which
helps to ensure containment integrity. Testing, however, may introduce
the potential for damage to the fuses and fuse clips. Therefore, the
deletion of this TS requirement 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 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: Fairfield County Library,
Garden and Washington Streets, Winnsboro, South Carolina 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Project Director: William H. Bateman.
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: November 22, 1993.
Description of amendment request: The proposed amendment would
result in the replacement of most of the analog Riley temperature
instrumentation associated with leak detection with digital equipment
from the General Electric Company NUMAC product line. Technical
Specification changes would be made to instrumentation surveillance
requirements for temperature instruments associated with main steam
line isolation, reactor water cleanup system isolation, reactor core
isolation cooling system isolation, and residual heat removal system
isolation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The Technical specifications are proposed to be revised to perform
a Channel Functional Test on a semiannual frequency versus the current
monthly frequency for both ambient and differential temperature and for
the MSL tunnel temperature timer functions, for the above listed piping
lines. Additionally, this evaluation addressees the potential for, and
implications of, common mode failures due to software, hardware and/or
electromagnetic and radio-frequency interference (EMI/RFI).
The NUMAC instrumentation has certain design features which
contribute to its reliability. The replacement NUMAC LDMs are digital
instruments that use a microcomputer to monitor the ambient and
differential temperatures (also the MSL tunnel temperature timer) and
provide outputs and automatic self-testing and calibration. A
description of the major design features include: a) isolation of the
essential microcomputer by a serial data link from the front panel
display (and display microcomputer), b) a self-test system feature that
provides automatic testing of internal circuits and reports failures,
c) thermocouple failure detection, d) provisions to test the output
relays without the use of jumpers (reducing the threat of spurious
isolation), and e) two independent built-in instrument power supplies
(that automatically switchover to the other supply in the event of
failure). Also, several features have been included, among these, a) a
hardware ``watchdog'' timer to monitor against software cycling in
continuous loops, and b) software structured with the safety-related
essential tasks running at the highest priority in the system. These
capabilities increase the reliability of the collected data, reduce the
possibility of inadvertent isolation and plant shutdowns, reduce the
need for frequent calibrations, and reduce the likelihood of common
mode failures.
The NUMAC Leak Detection Monitors will maintain the same
environmental and electrical physical independence criteria
(qualifications) as the existing Leak Detection System components. The
LDMs, the associated thermocouple input units (TCIUs), and relay output
units (ROUs) will be mounted seismically such that qualification of
these components and the Control Room panels will be maintained. The
LDMs are qualified for the PNPP Control Room environment. The LDMs (one
per division) will be physically and electrically independent of each
other and do not share power supplies, thermocouple inputs, output
relays, microcomputer logic units, display units or enclosures and
mounting locations. A postulated gross failure of any one NUMAC LDM,
such as gross malfunction of the input unit, microcomputer logic unit
or the relay output unit, will not propagate to the other NUMAC LDM,
such as gross malfunction of the input unit, microcomputer logic unit
or the relay output unit, will not propagate to the other NUMAC LDM.
Thus, a failure within one NUMAC LDM will not prevent or disable the
function of the other NUMAC LDM. A failure within one NUMAC LDM may
cause the loss of one division of the isolation trip logic. However,
since the other redundant division (the MSLs have three other
divisions) will not be affected by this failure, the Leak Detection
System will still be able to perform its designed safety-related
function and provide the necessary system isolation. This is the same
as the current Leak Detection System design basis.
The possibility of a common mode failure of both NUMAC LDM
divisions is minimized by the design of the NUMAC hardware and
software, the verification and validation (V&V) of the software to
reduce the likelihood of errors, the testing of the software (to
discover and eliminate errors), and the design of and testing of the
hardware to demonstrate its resistance to EMI/RFI. The NUMAC instrument
design features, by effectively eliminating the potential for common
mode failures, maintain the Leak Detection System within its current
licensing basis. (A discussion of common mode failure protection is
presented in more detail in the answer to question two.) Therefore, the
design, isolation and separation criteria remain the same.
Additionally, as described within Chapter 7 of the Updated Safety
Analysis Report (USAR), diversity is provided to the ambient and
differential temperature monitoring trip functions for the various
systems by alternative leak detection methods (such as measuring steam
line flow or pressure) that provide backup in the event of the loss of
both divisions of the NUMAC Leak Detection Monitors. These alternative
leak detection methods are physically separate from those being
performed by the NUMAC LDM and constitute a diverse, redundant, safety
related backup capable of responding to a design basis line break for
the various systems. Therefore, a common mode failure of both LDM
divisions would not prevent any of the necessary system isolation from
occurring.
No changes are being made to the isolation logic of the Leak
Detection System. No accident initiators or precursors are affected by
the proposed changes to the Channel Functional Test surveillance
intervals for the various trip functions. One purpose of a Channel
Functional Test is to check the instrument setpoints. The NUMAC
instrument setpoints are set digitally and do not drift. An engineering
evaluation has established that the Channel Functional Test
surveillance interval can be extended from one to six months. The
potential for common mode failures has been accounted for in the design
and measures have been taken to lower the probability of this to an
acceptable level (see the answer to question two). Also, alternative
leak detection methods exist for this eventuality. Since the NUMAC Leak
Detection Monitoring equipment meets or exceeds the design and
licensing criteria specified for the Leak Detection System, the
proposed upgrade cannot increase the probability of occurrence of any
accident previously evaluated.
A portion of the Leak Detection System logic causes a closure of
the Main Steam Isolation Valves on a steam leak signal. This transient,
described in Chapter 15 of the USAR, may also occur due to a LDS
equipment malfunction. Since this modification replaces some of the
existing Riley temperature monitoring instrumentation with more
reliable instrumentation the probability of this transient is reduced
(no radiological consequences are associated with this event). The LDS
is also used to mitigate the consequences of a pipe break outside
primary containment by isolating the affected system connected to the
Reactor Coolant Pressure Boundary (RCPB). The replacement of the Riley
instrumentation with NUMAC LDMs will not change, degrade, or prevent
the Leak Detection System response to mitigate the radiological
consequences of an accident. Therefore, replacement of the Riley
temperature modules with NUMAC Leak Detection Monitors will not
significantly increase the consequences of any accident previously
evaluated.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The single failure criterion requires that any single failure
within a safety-related system not prevent proper protective action of
the overall system when the system is required to function. The Leak
Detection System design is such that a failure of one division will not
prevent the system from performing its safety function. Common mode
failure protection provisions have been addressed in the NUMAC LDM
design.
The comprehensive General Electric software V&V and configuration
management control programs minimize, although they cannot entirely
eliminate, the likelihood of a common mode NUMAC instrument failure due
to software problems. The hardware (firmware) and software for the PNPP
NUMAC Leak Detection Monitors will undergo a formal software
verification and validation (V&V) process by General Electric, that is
to be completed by the end of the year, equivalent to the one reviewed
and approved by the NRC for the safety-related Wide Range Neutron
Monitor.
The NUMAC instruments are designed to minimize both their
susceptibility to, and generation of, electromagnetic and radio-
frequency interference (EMI/RFI) to prevent spurious operations and
allow their use in safety-related systems. As part of a broader plan by
GE to improve the testing has been performed by GE on the Leak
Detection Monitor configuration in order to both expand the overall
qualification region, and to obtain test data specific to this
application. This testing ensures the qualification of the Thermocouple
Input Unit (TCIU), a NUMAC circuit board which is unique to the LDM
application, and also extends the NUMAC EMI/RFI qualification region to
include both higher and lower frequencies than previously tested.
The NUMAC instrument design concept has undergone review by the
NRC, and the initial instruments of the NUMAC product line (the
Logarithmic Radiation Monitor and Wide Range Neutron Monitoring System)
have received NRC approval via Safety Evaluation of the associated GE
Licensing Topical Reports. The various types of NUMAC equipment in
operation at other nuclear power plants have components and software
modules which are similar to and in some instance identical to the
NUMAC LDMs. Therefore, based on the NRC reviewed and approved NUMAC
software and hardware control programs instituted by GE, the design
features to minimize software/hardware (or their interface) problems,
design features to minimize susceptibility to EMI/RFI, and testing to
demonstrate resistance to EMI/RFI, installation of NUMAC Leak Detection
Monitors at the PNPP does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed changes do not involve a significant reduction in a
margin of safety.
The replacement of the analog Riley temperature modules with the
microcomputer based NUMAC Leak Detection Monitors will not affect any
design conditions or impact the margins of safety for the various Leak
Detection System monitored parameters in the Technical Specification
Table 3.3.2-2 will not be changed or affected by this modification.
Only the CHANNEL FUNCTIONAL test interval is being extended.
The NUMAC Leak Detection Monitor design, with the attention paid
towards minimizing the potential for, and the effect of, software/
hardware and/or EMI/RFI related problem or common mode failures and
resulting operational experience has demonstrated that replacement of
the existing Riley temperature modules with NUMAC Leak Detection
Monitors would not result in a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: February 8, 1994, as supplemented March
25, 1994.
Description of amendment request: The amendment would revise the
WNP-2 Technical Specifications. Specifically, the amendment would
increase 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 the note (j) reference would be deleted from RCIC-V-8 and
RCIC-V-63. The note (j) indicates that the stroke time specified in the
Table reflects the requirement for containment isolation only.
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 proposed action does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
RCIC-V-8 and V-63 are containment isolation valves and are normally
open. Failure of the valves to open or close cannot cause an accident.
The mitigating capability of RCIC-V-8 and V-63 is not changed in that
the valves will continue to be closed within the established time
limits. This ensures protection of the safety related equipment
necessary for continued compliance with the requirements of General
Design Criterion 4. In those accidents which involve a source term and
potential adverse dose release consequences, no credit is taken for the
closing of the valves; therefore the increase in the allowable time for
closing does not increase the consequences of those accidents.
2. This proposed action does not create the possibility of a new or
different kind of accident from any accident previously evaluated. The
requested Technical Specifications change does not represent a change
in modes of operation. It does not, in itself, require physical
modification to the plant, although it will be used to allow a gear
change in RCIC-MO-8. The new gears represent a standard configuration
for Limitorque motor operators and will require a routine design
change. The required Technical Specification change maintains the
licensing basis for the plant as discussed in response to question 1.
Hence, no new or different kind of accident is possible as a result of
implementing this change.
3. This proposed action does not involve a significant reduction in
a margin of safety. The increase in stroke time will increase the peak
temperature in the HELB profiles and thereby decrease the margin
available from the equipment qualification limits. However, sufficient
margin remains to assure the equipment operability is maintained and
there is no reduction in the margin of safety. Additionally, there is
no reduction in the margin of safety because increasing the stroke time
will not change the postulated radiological releases.
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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352.
Attorney for licensee: M.H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502.
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.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: April 14, 1994.
Description of amendments request: The proposed change would
relocate the instrument response time tables 3.3.1-2, Reactor
Protection System (RPS) Response Times; 3.3.2-3, Isolation System
Instrumentation (ISI) Response Time; and 3.3.3-3, Emergency Core
Cooling System (ECCS) Response Times, from the Technical Specifications
to the Updated Final Safety Analysis Report. The RPS, ISI, and ECCS
instrument limiting conditions for operation (LCO) will be revised to
read that the instruments ``shall be operable'' without a reference to
a specific response time table in these LCOs. The references to the
response time tables will also be deleted from the Surveillance
Requirements.
Date of publication of individual notice in Federal Register: April
26, 1994 (59 FR 21785).
Expiration date of individual notice: May 26, 1994.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Gulf States Utilities Company, Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of application for amendment: January 14, 1994.
Brief description of amendment request: The proposed amendment
would revise various instrumentation technical specifications by
extending the allowable outage times (AOTs) of the instruments, and by
increasing their channel functional surveillance test intervals (STIs)
to quarterly. The amendment also revises certain technical
specification actions to address loss-of-function concerns associated
with the AOT and STI changes.
Date of individual notice in Federal Register: April 26, 1994(59 FR
21787).
Expiration date of individual notice: May 26, 1994.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Gulf States Utilities Company, Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of application for amendment: February 22, 1994.
Brief description of amendment request: The proposed amendment
would revise the technical specifications (TS) for the main steam-
positive leakage control system (MS-PLCS) and the penetration valve
leakage control system (PVLCS) to be consistent with the requirements
contained in NUREG-1434, ``Standard Technical Specifications, General
Electric Plants (BWR/6).''
Date of individual notice in Federal Register: March 10, 1994 (59
FR 11331).
Expiration date of individual notice: April 11, 1994.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Gulf States Utilities Company, Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit
1, West Feliciana Parish, Louisiana
Date of application for amendment: March 3, 1994.
Brief description of amendment request: The amendment would revise
the technical specifications in accordance with the guidance provided
by Generic Letter 93-08, ``Relocation of Technical Specification Tables
of Instrument Response Time Limits.'' Generic Letter 93-08 recommends
the removal and subsequent relocation of various technical
specification tables which denote instrument and system response time
limits.
Date of individual notice in Federal Register: March 16, 1994 (59
FR 12380).
Expiration date of individual notice: April 15, 1994.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of application for amendments: October 26, 1993, as
supplemented March 28, 1994.
Brief description of amendments: The licensee is requesting a
revision to TS 5.3.1 for Palo Verde Nuclear Generating Station Units 1,
2, and 3 that will increase the maximum allowable fuel enrichment from
4.05 weight percent U-235 to 4.30 weight percent U-235. There was no
change requested to the current 52,000 MWD/MTU burnup. The licensee
provided a supplemental letter dated March 28, 1994, at the request of
the NRC to bring TS 5.3.1 into conformance with Generic Letter 90-02,
Supplement 1 and to clarify assumptions used in the Fuel Handling
Accident Analysis.
Date of issuance: April 19, 1994.
Effective date: April 19, 1994, to be implemented within 45 days of
issuance.
Amendment Nos.: 74, 60 and 46.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2860) The additional information contained in the supplemental letter
dated March 28, 1994, was clarifying in nature and thus 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 amendments is contained in a Safety
Evaluation dated April 19, 1994. No significant hazards consideration
comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004.
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 1, 1992.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3/4.4.3, ``Relief Valves,'' to improve the
reliability of the reactor coolant system's power-operated relief
valves (PORVs) and their associated block valves for overpressure
protection during normal operation and anticipated transients. The
amendments also revise TS 3/4.4.9, ``Pressure/Temperature Limits,'' to
improve the availability of the PORVs for low temperature overpressure
protection. Accompanying changes are also made to the associated TS
Bases. These revisions were made in response to Generic Letter 90-06,
``Resolution of Generic Issue 70, `Power-Operated Relief Valve and
Block Valve Reliability,' and Generic Issue 94, `Additional Low-
Temperature Overpressure Protection for Light-Water Reactors,' pursuant
to 10 CFR 50.54 (f).''
Date of issuance: April 20, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 188 and 165.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 30, 1992 (57
FR 45076).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 20, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: February 4, 1994.
Brief description of amendment: The amendment revises TS
Surveillance Requirement 4.6.4.1 to delete the 12-hour channel check,
thereby eliminating the need for continuous operation of the hydrogen
monitors.
Date of issuance: April 26, 1994.
Effective date: April 26, 1994.
Amendment No. 47.
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10001) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 26, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. STN
50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: March 21, 1994, as supplemented
March 24, 1994.
Brief description of amendments: The amendments add a one-time
revision to Technical Specification (TS) 3/4.7.1.1 to permit continued
activities at all four units with main steam Code safety valve lift
setpoint tolerances of 3%. The duration of this amendment
is until May 9, 1994, at which time the tolerances will be reset to
1%. A statement has also been added to TS 4.7.1.1 for
Braidwood stating that the provisions of TS 4.0.4 are not applicable to
Braidwood, Unit 1, Cycle 5 until initial entry into Mode 2 from its
refueling outage.
Date of issuance: April 18, 1994.
Effective date: April 18, 1994.
Amendment Nos.: 61, 61, 49, and 49.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
Amendment revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (59 FR 14685 dated March 29, 1994). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination within 15 days. No
comments have been received. The notice also provided an opportunity to
request a hearing by April 29, 1994, but indicated that if the
Commission makes a final no significant hazards consideration
determination any such hearing would take place after issuance of the
amendment. The Commission's related evaluation of the amendment and
final no significant hazards consideration determination is contained
in a Safety Evaluation dated April 18, 1994.
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.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: September 29, 1993, as
supplemented by letter dated April 8, 1994.
Brief description of amendment: The amendment revises Technical
Specification Section 3.9.A.5 and Tables 3.9-1 and 4.10-2 to delete
controls for the 21, 22, and 23 Boron Monitor Tanks, which are no
longer in service.
Date of issuance: April 28, 1994.
Effective date: As of the date of issuance to be implemented after
the inlet and outlet lines of the 21, 22, and 23 Boron Monitor Tanks
have been cut and capped.
Amendment No.: 169.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 24, 1993 (58
FR 62154)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of application for amendment: May 23, 1993.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.4.3.2.d and related Table 3.4.3.2-1 by changing
the allowable leakage for certain low pressure coolant injection (LPCI)
line pressure isolation valves and revises Table 3.6.3-1 to remove the
designation as containment isolation valves from the LPCI injection
reverse flow check and bypass valves. The related Bases are also
changed. Concurrently, the Commission granted an exemption from the
requirements of 10 CFR Part 50, Appendix J, III.C. for performing Type
C containment integrated leak rate tests of the containment isolation
valves in the low pressure coolant injection lines of the residual heat
removal system and to perform alternative testing.
Date of issuance: April 22, 1994.
Effective date: April 22, 1994, with full implementation within 45
days.
Amendment No.: 98.
Facility Operating License No. NPF-43. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46227) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 22, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: February 24, 1993.
Brief description of amendment: The amendment corrected
typographical errors in the plant technical specifications (TSs). These
errors were introduced in the original ANO-1 TS, and in subsequent
amendments. These changes are administrative in nature and are intended
to improve the readability of the plant technical specifications
without changing the meaning or intent of any specifications.
Date of issuance: April 26, 1994.
Effective date: April 26, 1994.
Amendment No.: 171.
Facility Operating License No. DPR-51. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67843)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 26, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 16, 1993, as supplemented by
letter dated April 5, 1994.
Brief description of amendment: The amendment revised the Technical
Specifications to provide acceptable conditions for operation when the
core operating limit supervisory system (COLSS) is out of service and
either or both control element assembly calculators (CEACs) are
operable.
Date of issuance: April 22, 1994.
Effective date: April 22, 1994.
Amendment No.: 93.
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 5, 1994 (59 FR
620) The additional information contained in the supplemental letter
dated April 5, 1994, withdrew a portion of the original application and
thus, was within the scope of the initial notice and did not affect the
staff's proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated April 22, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 14, 1994.
Brief description of amendment: The amendment revised the Technical
Specifications in response to Generic Letter 93-08 issued by the NRC
and dated December 29, 1993, by relocating the reactor trip system and
engineered safety features actuation system response time limits to the
updated final safety analysis report.
Date of issuance: April 22, 1994.
Effective date: April 22, 1994.
Amendment No.: 94.
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12360) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 22, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: February 25, 1992.
Brief description of amendments: The amendments relate to your
application dated February 25, 1992, which requested a 40-year
operating license commencing from the date of issuance of the operating
license and, accordingly, would extend the operating license expiration
date for Turkey Point Units 3 and 4 to July 19, 2012 and April 10,
2013, respectively.
Date of issuance: April 20, 1994.
Effective date: April 20, 1994.
Amendment Nos. 162 and 156.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 1992 (57 FR
13130) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 20, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: December 28, 1993.
Brief description of amendments: These amendments include steam
generator overfill protection in the Technical Specifications in
response to Generic Letter 89-19, Request for Action Related to
Resolution of Unresolved Safety Issue A-47 ``Safety Implications of
Control Systems in LWR Nuclear Power Plants.''
Date of issuance: April 28, 1994.
Effective date: April 28, 1994.
Amendment Nos. 163 and 157.
Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10007) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 28, 1994. No significant
hazards consideration comments received: No.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: March 14, 1994.
Brief description of amendment: The amendment changes the Millstone
Unit 2 Technical Specifications (TS) to provide a one-time extension of
the surveillance frequency from the required 18-month to the next
refueling outage but no later than September 30, 1994, of the power
operated valves in the service water system (TS 4.7.4.1.b) and in the
boron injection flowpath (TS 4.1.2.2.c). This extends the surveillance
for these valves approximately 5 months.
Date of issuance: April 22, 1994.
Effective date: April 22, 1994.
Amendment No.: 173.
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 23, 1994 (59 FR
13751). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 22, 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.
Northeast Nuclear Energy Company, Docket Nos. 50-245, 50-336, and 50-
423, Millstone Nuclear Power Station, Units 1, 2 and 3, New London
County, Connecticut
Date of application for amendment: December 22, 1993.
Brief description of amendment: The amendments revise the Technical
Specifications (TS) as follows:
1. Change the title of the Nuclear Station Director to Senior Vice
President--Millstone Station.
2. Remove the requirement to provide a copy of Plant Operations
Review Committee (PORC) and Site Operations Review Committee (SORC)
meeting minutes to the Executive Vice President--Nuclear. The Senior
Vice President--Millstone Station replaces the Executive Vice
President--Nuclear for receipt of PORC and SORC meeting minutes.
3. Make editorial changes to the Millstone Unit No. 1 TS Index.
4. Correct a typographical error in Section 6.2.1.d of the
Millstone Unit No. 1 TS.
5. Correct a typographical error in Section 6.5.3.1.a of the
Millstone Unit No. 3 TS.
Date of issuance: April 26, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 74, 174, and 90.
Facility Operating License No. DPR-21, DPR-65, AND NPF-49.
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: February 16, 1994 (59
FR 7693) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 26, 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.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: April 19, 1993, as supplemented
by letter dated April 18, 1994.
Brief description of amendments: These amendments extend
surveillance test interval and allowed outage times for the containment
isolation actuation instrumentation.
Date of issuance: April 26, 1994.
Effective date: April 26, 1994.
Amendment Nos. 69 and 32.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34086) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 26, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: May 6, 1993, as supplemented by
letter dated April 18, 1994.
Brief description of amendments: These amendments extend
surveillance test interval and allowed outage times for selected
actuation instrumentation and makes editorial changes.
Date of issuance: April 26, 1994.
Effective date: April 26, 1994.
Amendment Nos. 70 and 33.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34087) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 26, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Public Service Electric and Gas Company
Delmarva Power and Light Company, and Atlantic City Electric Company,
Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit
Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: November 19, 1993.
Brief description of amendments: These amendments eliminate the
listing of specific position titles for the Plant Operations Review
Committee (PORC) composition in favor of allowing the Plant Manager to
appoint PORC members. This revision eliminates the need to change the
TS in the future whenever a position title is changed.
Date of issuance: April 26, 1994.
Effective date: April 26, 1994.
Amendments Nos.: 190 and 195.
Facility Operating License Nos. DPR-44 and DPR-56: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 5, 1994 (59 FR
628) The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 26, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
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: January 31, 1994.
Brief description of amendment: This amendment to the Appendix B
Technical Specifications (TSs), the Radiological Effluent TSs, revised
Section 3.5, and the associated Bases, to establish a threshold level
below which there will be no requirement to perform grab samples and
isotopic analyses of steam jet-air ejector (SJAE) effluent and revised
TS Table 3.10-1 to change the actions required when entering an SJAE
limiting condition for operation. Additionally, the amendment revised
the TSs to clarify instructions and make editorial corrections which
are administrative in nature.
Date of issuance: April 25, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 211.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10014).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 25, 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. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: March 12, 1993.
Brief description of amendment: The amendment revised the Technical
Specifications (TS) to incorporate the changes listed below:
(1) The frequency of high pressure water fire protection system
testing (specified in TS Section 4.12.A.1) was changed to accommodate
operation on a 24-month cycle.
(2) The frequency of fire pump diesel engine testing (specified in
TS Section 4.12.A.2) was changed to accommodate operation on a 24-month
cycle.
(3) The frequency of electrical tunnel, diesel generator building,
and containment fan cooler fire protection spray and/or sprinkler
system testing (specified in TS Section 4.12.B.1) was changed to
accommodate operation on a 24-month cycle.
(4) The frequency of fire barrier penetration seal inspection
(specified in TS Section 4.12.C.1) was changed to accommodate operation
on a 24- month cycle.
(5) The frequency of fire detection system testing (specified in TS
Section 4.12.D.1) was changed to accommodate operation on a 24-month
cycle.
(6) The frequency of fire hose station testing (specified in TS
Section 4.12.E.1) was changed to accommodate operation on a 24-month
cycle.
(7) The frequency of CO2 fire protection system testing
(specified in TS Section 4.12.G.1) was changed to accommodate operation
on a 24-month cycle. A new requirement was also added to exercise the
fire dampers on an annual basis.
These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to
Accommodate a 24-Month Fuel Cycle,'' as applicable.
In addition, TS Section 4.12 was reformatted, in its entirety, and
several administrative changes were made to improve clarity.
Date of issuance: April 20, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 146.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1993 (58 FR
25862) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 20, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: March 4, 1994.
Brief description of amendment: This amendment adds a new TS 3/
4.10.8, ``Inservice Leak and Hydrostatic Testing,'' to the Hope Creek
Generating Station TSs. The amendment also includes corresponding
changes to the TS Index, Table 1.2, ``OPERATIONAL CONDITIONS,'' and
provides Bases for TS 3/4.10.8. The added TS 3/4.10.8 permits the unit
to remain in OPERATIONAL CONDITION 4 with the average reactor coolant
temperature being increased above 200 deg.F, but not to exceed
212 deg.F, and certain OPERATIONAL CONDITION 3 Limiting Conditions for
Operation for secondary containment isolation, secondary containment
integrity and filtration, recirculation and ventilation system (FRVS)
operability being met.
Date of issuance: April 18, 1994.
Effective date: April 18, 1994.
Amendment No.: 69.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12384).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 18, 1994.
No significant hazards consideration comments received: None.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: December 23, 1992, as
supplemented August 12, 1993 and January 21, 1994 (TS 328).
Brief description of amendments: The amendments modify the
operability requirements for the low pressure coolant injection (LPCI)
mode of the residual heat removal (RHR) system while the reactor is
shut down. The amendments permit the RHR system to be considered
operable for LPCI when aligned for shutdown cooling if it can be
manually realigned and is not otherwise inoperable.
Date of issuance: April 19, 1994.
Effective date: April 19, 1994.
Amendment Nos.: 204, 223, and 177.
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Date of initial notice in Federal Register: March 31, 1993 (58 FR
16873).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 1994.
No significant hazards consideration comments received: None.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: February 7, 1994 (TS 93-11).
Brief description of amendments: The amendments replace the wording
in Surveillance Requirement 4.7.9.i, ``Snubber Service Life Program,''
with that from the Westinghouse Electric Corporation Standard TS,
Revision 4a. In addition, the amendments delete the wording in SR
4.7.9.c, ``Snubber Visual Inspection Performance and Evaluation,'' that
is inconsistent with Generic Letter 90-09.
Date of issuance: April 18, 1994.
Effective date: April 18, 1994.
Amendment Nos.: 179--Unit 1 171--Unit 2.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: February 7, 1994.
The Commission's related evaluation of the amendments are contained
in a Safety Evaluation dated April 18, 1994.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: February 7, 1994 (TS 93-19).
Brief description of amendments: The amendments revise Technical
Specification 5.3.1 to allow the substitution of filler rods for fuel
rods in fuel assemblies by incorporating the guidance in Generic Letter
90-02, Supplement 1.
Date of issuance: April 18, 1994.
Effective date: April 18, 1994.
Amendment Nos.: 180--Unit 1 172--Unit 2.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12367) The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated April 18, 1994.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.
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 application for amendment: September 28, 1992.
Brief description of amendment: The amendment revised Technical
Specification (TS) 2.2, Limiting Safety System Settings, TS 3.3.1,
Reactor Protection System Instrumentation, and TS 3.3.2, Isolation
Actuation Instrumentation by removing the functions associated with the
main steam line radiation monitors.
Date of issuance: April 22, 1994.
Effective date: April 22, 1994.
Amendment No. 58.
Facility Operating License No. NPF-58. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 6, 1993 (58 FR
598) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 22, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
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: July 28, 1992, as supplemented
on February 17, 1993.
Brief description of amendment: The proposed amendment would delete
Technical Specification (TS) 3/4.9.9, ``Refueling Operations--
Containment Purge and Exhaust Isolation System,'' and its bases,
because of its redundancy to other TSs that address the operability
requirements of the containment purge and exhaust isolation system.
Also, the proposed amendment would revise TS 3/4.3.2, ``Safety System
Instrumentation--Safety Features Actuation System Instrumentation,''
and TS 3.4.9.4, ``Refueling Operations--Containment Penetrations,'' and
its bases. The effect of this proposed change would be to allow the
bypass of the safety features actuation system in Mode 6,
``Refueling,'' by the use of the containment purge and exhaust system
noble gas monitor in conjunction with manual closure of the containment
purge and exhaust isolation valves instead of automatic closure.
Date of issuance: April 15, 1994.
Effective date: April 15, 1994.
Amendment No. 186.
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 6, 1993 (58 FR
599) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 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.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: October 8, 1993.
Brief description of amendments: The amendments revise the
technical specifications (TS) by deleting tables listing certain
components from the TS and relocating the lists to plant procedures in
accordance with the guidance provided in NRC Generic Letter 91-08,
``Removal of Component Lists from Technical Specifications.''
Date of issuance: April 22, 1994.
Effective date: April 22, 1994.
Amendment Nos.: 181 and 162.
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57860) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 22, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of application for amendment: July 29, 1993, as supplemented
by letters dated March 11 and 17, 1994.
Brief description of amendment: The amendment modifies the
Technical Specifications (TS) to reflect a new refueling mast.
Specifically, the amendment adds new values for protective features in
the TS to reflect the new refueling mast. Values for the old refueling
mast are retained in the TS.
Date of issuance: April 29, 1994.
Effective date: April 29, 1994.
Amendment No.: 121.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14900) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 29, 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.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301 Point
Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks Manitowoc
County, Wisconsin
Date of application for amendments: February 26, 1993.
Brief description of amendments: The amendments adding operating
conditions and limiting conditions for operation for the atmospheric
steam dump valves, the crossover steam dump system, the turbine stop
and governor valves, and the various turbine overspeed protection
features installed at the Point Beach Nuclear Plant. Additionally, the
amendments revised the surveillance requirements for the auxiliary
feedwater system, and added explanatory text to the bases for Sections
15.3.4 and 15.4.8.
Date of issuance: April 20, 1994.
Effective date: April 20, 1994.
Amendment Nos.: 147 and 151.
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43939) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 20, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 21, 1993, as supplemented by
letters dated March 14, 1994, and April 18, 1994.
Brief description of amendment: The amendment revises Technical
Specification Sections 6.5.1, Plant Safety Review Committee (PSRC) and
6.8, Procedures and Programs, in order to allow implementation of a
Qualified Reviewer Program for the review and approval of new
procedures and procedure changes. Technical Specification 6.5.1.6, PSRC
Responsibilities, has also been revised in accordance with Generic
Letter 93-07, ``Modification of Technical Specification Administrative
Control Requirements for Emergency and Security Plans,'' to delete
requirements for PSRC review of the Emergency Plan and Security Plan
and related implementing procedures.
Date of Issuance: April 28, 1994.
Effective date: April 28, 1994, to be implemented within 120 days
of issuance.
Amendment No.: Amendment No. 73.
Facility Operating License No. NPF-42. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 24, 1993 (58
FR 62159) The March 14, 1994, and April 18, 1994, supplemental letters
provided additional clarifying information and revised the
implementation period and did not change the initial no significant
hazards consideration. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated April 28, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room Locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Dated at Rockville, Maryland, this 4th day of May 1994.
For the Nuclear Regulatory Commission.
Jack W. Roe,
Director Division of Reactor Projects--III/IV Office of Nuclear Reactor
Regulation.
[FR Doc. 94-11226 Filed 5-11-94; 8:45 am]
BILLING CODE 7590-01-P