[Federal Register Volume 63, Number 57 (Wednesday, March 25, 1998)]
[Notices]
[Pages 14482-14497]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-7652]
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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
[[Page 14483]]
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 March 2, 1998, through March 13, 1998. The
last biweekly notice was published on March 11, 1998 (63 FR 11913).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By April 24, 1998, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the
[[Page 14484]]
Commission may issue the amendment and make it immediately effective,
notwithstanding the request for a hearing. Any hearing held would take
place after issuance of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick
Steam Electric Plant, Unit 1, Brunswick County, North Carolina
Date of amendment request: February 23, 1998.
Description of amendment request: The amendment request proposes
changes to the Brunswick Steam Electric Plant Unit 1 Technical
Specifications (TS) in support of Cycle 12 operation, including a
change to the Minimum Critical Power Ratio safety limit (safety limit
MCPR) to a value equivalent to the generic safety limit MCPR for
General Electric type GE-13 fuel. The request would additionally remove
a footnote limiting the stated value for the safety limit MCPR to a
specific fuel cycle and reference to an NRC safety evaluation
documenting acceptance of methods used for determining the current
cycle safety limit MCPR. The amendment request is provided both in the
format of the current TS as well as improved Standard Technical
Specifications (iSTS). The Brunswick licensee applied for conversion to
ISTS on November 1, 1996, as supplemented on October 13, 1997, and
February 26, 1998, and that application is currently undergoing NRC
staff review. For iSTS, the licensee has proposed two safety limits
MCPR, one pertaining to two-recirculation loop operation and the other
to single-recirculation loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed license amendment establishes a revised safety
limit MCPR value of 1.09 [two-recirculation loop and 1.10 for
single-recirculation loop operation] for use during Unit 1 Cycle 12
operation. General Electric (GE) has determined that both generic
and plant-specific evaluations [two-loop operation] yield the same
calculated safety limit MCPR value. Additionally, a document
referenced by the Technical Specification 6.9.3.2 of methodologies
used in determining core operating limits is being removed.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established, consistent with NRC[-]
approved methods, to ensure that fuel performance during normal,
transient, and accident conditions is acceptable.
The probability of an evaluated accident is not increased by
revising the safety limit MCPR value to 1.09 [two-loop/1.10 single-
loop]. The change does not require any physical plant modifications
or physically affect any plant components. Therefore, no individual
precursors of an accident are affected.
The proposed license amendment establishes a revised safety
limit MCPR that ensures the fuel is protected during normal
operation and during any plant transients or anticipated operational
occurrences. Specifically, the reload analysis demonstrates that a
safety limit MCPR value of 1.09 [two-loop/1.10 single-loop] ensures
that less than 0.1 percent of the fuel rods will experience boiling
transition during any plant operation if the limit is not violated.
The methods for calculating the safety limit MCPR have been
approved by the NRC and are described in GE's reload licensing
methodology topical report NEDE-24011, ``General Electric Standard
Application for Reactor Fuel (GESTAR II).'' Based on (1) the
determination of the new safety limit MCPR value using conservative
approved methods, and (2) the operability of plant systems designed
to mitigate the consequences of accidents not having been changed;
the consequences of an accident previously evaluated have not been
increased.
Additionally, removal of the footnote on the safety limit MCPR
value in Technical Specification 2.1.2 and removal of reference
``c'' from the document list in Technical Specification 6.9.3.2 will
not increase the probability or consequences of accidents previously
evaluated. The footnote on the safety limit MCPR value in Technical
Specification 2.1.2 and reference ``c'' in Technical Specification
6.9.3.2 were associated with the safety limit MCPR value of 1.10 for
Unit 1 Cycle 11 operation. Since the current safety limit MCPR value
of 1.10 applies only to Unit 1 Cycle 11 operation, the footnote on
the safety limit MCPR value in Technical Specification 2.1.2 and the
reference ``c'' in Technical Specification 6.9.3.2 are no longer
needed and should be deleted. Thus, removal of the footnote on the
safety limit MCPR value in Technical Specification 2.1.2 and removal
of reference ``c'' from Technical Specification 6.9.3.2 is an
administrative change that has no effect on the probability or
consequences of accidents previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
This proposed license amendment involves a revision of the
safety limit MCPR from 1.10 to 1.09 [two-loop/1.10 single-loop]
based on the results of both cycle-specific and generic analyses,
removal of the footnote on the safety limit MCPR value in Technical
Specification 2.1.2, and the removal of a document reference listed
in Technical Specification 6.9.3.2 describing the methods used only
during Unit 1 Cycle 11 to determine core operating limits. Creation
of the possibility of a new or different kind of accident would
require the creation of one or more new precursors of that accident.
New accident precursors may be created by modifications of the plant
configuration, including changes in allowable modes of operation.
This proposed license amendment does not involve any modifications
of the plant configuration or changes in the allowable modes of
operation. Therefore, no new precursors of an accident are created
and no new or different kinds of accidents are created.
3. Does this change involve a significant reduction in a margin
of safety?
As previously stated, the methods for calculating the safety
limit MCPR have been previously approved by the NRC and are
described in GE's reload licensing methodology topical report NEDE-
24011. Use of these methods ensures that the resulting safety limit
MCPR satisfies the fuel design safety criteria that less than 0.1
percent of the fuel rods experience boiling transition if the safety
limit is not violated. Based on the assurance that the fuel design
safety criteria will be met, the proposed license amendment 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
[[Page 14485]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Pao-Tsin Kuo (Acting).
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: May 16, 1997.
Description of amendment request: The proposed changes would
replace the existing Technical Specification (TS) 4.6.2.3 a.2 cooling
water flow rate of 1425 gpm with a new value of 1300 gpm.
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.
Cooling water flow to the Containment Fan Coolers is provided by
the Emergency Service Water (ESW) System, and Emergency Service
Water is not an initiating system in any FSAR [Final Safety Analysis
Report] Chapter 15 analyses. Revising the minimum cooling water flow
to the Containment Fan Coolers will not increase the probability of
initiating any previously evaluated accident, because Containment
Fan Cooler performance and integrity will not be adversely affected.
The heat removal capacity of the Containment Fan Coolers will be
maintained consistent with the assumptions used in the existing HNP
[Harris Nuclear Plant] containment analyses, and, therefore,
containment integrity should not be challenged.
Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
(2) The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment will not create any new accident
scenarios, because the change does not introduce any new single
failures, adverse equipment or material interactions, or release
paths.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) The proposed amendment does not involve a significant
reduction in the margin of safety.
Although the proposed amendment replaces the TS 4.6.2.3 a.2
cooling water flow rate of 1425 gpm with a lower flow rate of 1300
gpm, a cooling water flow rate of greater than or equal to 1300 gpm
maintains adequate heat removal capacity as required by existing HNP
containment analyses. The Bases for TS 4.6.2.3 a.2 is to ensure that
adequate heat removal capacity is available, when the Containment
Fan Coolers are operated in conjunction with the Containment Spray
Systems, during post-LOCA [Loss-of-Coolant Accident] conditions to
prevent the pressure inside containment from exceeding its design
rating.
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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Pao-Tsin Kuo (Acting).
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power
Station, Unit No. 2, Shippingport, Pennsylvania
Date of amendment request: October 22, 1997
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) by reducing the reactor
coolant system (RCS) specific activity limits in accordance with
Generic Letter 95-05. The definition of DOSE EQUIVALENT I-131 would be
replaced with the Improved Standard TS definition wording in the first
sentence and an equation added based on dose conversion factors derived
from International Commission on Radiation Protection (ICRP) ICRP-30.
TS 3.4.8, Specific Activity, would be revised by reducing the DOSE
EQUIVALENT I-131 limit from 1.0 [micro] Ci[curies]/gram to 0.35
[micro]Ci[curies]/gram. Item 4.a in TS Table 4.4-12, Primary Coolant
Specific Activity Sample and Analysis Program, TS Figure 3.4-1, and the
Bases for TS 3/4.4.8 would be modified to reflect the reduced DOSE
EQUIVALENT I-131 limit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change reduces the reactor coolant system (RCS)
specific activity limits of Specification 3.4.8 from 1.0 [micro]Ci/
gram to 0.35 [micro]Ci/gram and lowers the graph in Figure 3.4-1 by
39 [micro]Ci/gram following the guidance provided in Generic Letter
(GL) 95-05. This reduces the RCS acvitity allowed to leak to the
secondary side when the plant is operating so that additional margin
is available to support a higher allowable accident-induced leakage
value as justified by analysis.
The proposed changes to Specification 3.4.8 and the definition
of DOSE EQUIVALENT I-131 ensure these requirements are consistent
with the latest analyses.
These changes implement the more restrictive RCS activity limits
in accordance with applicable analyses and GL 95-05 to ensure the
regulations are satisfied. Therefore, these changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not alter the configuration of the
plant or affect the operation with the reduced specific activity
limit. By reducing the specific activity limit, the limit would be
reached sooner to initiate evaluation of the out of limit condition.
The proposed changes will not result in any additional challenges to
the main steam system or the reactor coolant system pressure
boundary. Consequently, no new failure modes are introduced as a
result of the proposed changes. As a result, the main steam line
break, steam generator tube rupture and loss of coolant accident
analyses remain bounding. Therefore, the proposed change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change reduces the RCS specific activity limit to
0.35 [micro]Ci/gram along with lowering the Figure 3.4-1 limits by
39 [micro]Ci/gram. Reduction of the RCS specific activity limits
allows an increase in the limit for the projected SG [steam
generator] leakage following SG tube inspection and repair in
accordance with the voltage-based SG tube alternate repair criteria
(ARC). This follows the guidance provided in GL 95-05 and
effectively takes margin available in the specific activity limits
and applies it to the projected SG leakage for the ARC. This has
been determined to be an acceptable means for accepting higher
[[Page 14486]]
projected leakage rates while still meeting the applicable limits of
10 CFR [Part] 100 and GDC [General Design Criterion] 19 with respect
to offsite and control room doses.
The capability for monitoring the specific activity and
complying with the required actions remains unchanged. In addition,
there is no resultant change in dose consequences. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: March 3, 1998.
Description of amendment request: The licensee proposed to revise
Section 6.2.3.2 of the units' Technical Specifications. Currently, this
section prescribes that the Catawba Safety Review Group (SRG) be
composed of at least five individuals and at least three of these shall
have a bachelor's degree in engineering or related science and at least
2 years professional level experience in his/her field, at least 1 year
of which experience shall be in the nuclear field. The licensee
proposed to revise this section to provide the option of replacing one
of the three degreed individuals with one with at least 15 years of
professional level experience in his/her field, at least 10 years of
which experience shall be in the nuclear field, at least 3 years of
which nuclear experience shall be supervisory/managerial experience in
engineering, and shall hold or have held a Senior Reactor Operator
license. The licensee also proposed to editorially revise this section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below.
1. Would the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed amendment would only change administrative
requirements related to personnel qualifications for one of the five
SRG [Safety Review Group] positions. The SRG is an oversight group,
and the individual who meets the new qualification requirements
would be expected to perform at the same level of quality as an
individual who meets the current qualification requirements.
Changing qualification requirements for an individual who primarily
performs an oversight function will not have any direct effect on
the design or operation of any plant structures, systems, or
components. No previously analyzed accidents were initiated by the
functions of the SRG, and the SRG was not a factor in the
consequences of previously analyzed accidents. Therefore, the
proposed change would have no impact on the consequences or
probabilities of any previously evaluated accidents.
2. Would the change create the possibility of a new or
difference kind of accident from any accident previously evaluated?
No. The proposed change would not lead to any hardware or
operating procedure change. Hence, no new equipment failure modes or
accidents from those previously evaluated will be created.
3. Would the change involve a significant reduction in a margin
of safety?
No. Margin of safety is associated with confidence in the design
and operation of the plant. The proposed change to the Technical
Specifications does not involve any change to plant design or
operation. Thus, the margin of safety previously analyzed and
evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Project Director: Herbert N. Berkow.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: February 7, 1997.
Description of amendment request: The proposed amendment, if
approved, would revise Technical Specification (TS) as delineated
below:
1. 4160 Volt Tie From Unit 2.
TS sections 3.7.2.b & d to delete reference to the optional use of
the 4160 volt tie from the unit 2 transformer.
2. Emergency Load Sequence and Power Transfer.
a. The testing required by Section 4.5.1.1.b of the TS would be
considered satisfactory if the pumps have started and valves have
completed travel. The need to evidence the successful starting of pumps
and fans and the complete travel of valves by observation of control
board component operating lights will be deleted. Neither would a
second means of verification, such as: the station computer or control
board indicating lights initiated by separate limit switch contacts be
required.
b. Section 4.5.1.2.b would be revised in the same manner as
4.5.1.1.b above.
3. Reactor Building Cooling and Isolation System.
a. Section 4.5.3.1.a.1 of the TS would be revised to delete the
need to simultaneously test start a spray pump using a Reactor Building
30-psi high pressure test signal while testing the emergency loading
sequence.
The proposed change also eliminates the need to evidence the
successful starting of the spray pumps by observation of the control
board indicating lights or the use of the station computer for Sections
4.5.3.1.a.1 and 4.5.3.1.b.2.
4. Instrument Surveillance Requirements.
Table 4.1-1 of the TS would be revised to delete the strong motion
accelerometer and its quarterly battery check surveillance requirement.
5. Air Intake Tunnel (AIT) Fire Protection Systems.
Section 5.5 of the TS would be deleted. The description of the
equipment contained in Section 5.5 would be transferred to the Final
Safety Analysis Report (FSAR).
6. Hydrogen Recombiner System.
The Bases for Section 4.4.4 TS would be changed to reflect a
reduction in the time interval for operation of the hydrogen recombiner
following a loss of cooling accident (LOCA) from 9.8 to 9 days.
7. Various editorial and typographical errors would be corrected.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The revised TS eliminate overly prescriptive
requirements for evidencing component performance, the requirement
for redundant diesel block loading tests,
[[Page 14487]]
instrumentation from SR [surveillance requirement] tables having no
associated LCO [limiting condition for operation], AIT fire
protection systems descriptive text, and correct previous
typographical errors. Several of the proposed revisions involve
changes which are consistent with NUREG-1430, the Revised Standard
Technical Specifications (RSTS) for B&W plants. The reliability of
systems and components depended upon to prevent or mitigate the
consequences of accidents previously evaluated is not degraded by
the proposed changes because assurance of system and equipment
availability is maintained by surveillance testing program
requirements.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The revised
surveillance requirements create no new failure modes. Verification
of equipment operation continues to be required by plant procedures.
Elimination of the AIT fire protection system descriptive text from
the TSs would not create a new or different kind of accident since
the change has no effect on surveillance methodology and frequency
requirements. They are maintained in the Fire Protection Program.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety because no operating limits are affected.
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: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas, Director.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: March 3, 1998.
Description of amendment request: The proposed revision to the
Millstone Unit 3 licensing basis would eliminate the requirement to
have the recirculation spray system directly inject into the reactor
coolant system following a design basis accident.
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:
Northeast Nuclear Energy Company (NNECO) has reviewed the
proposed revision in accordance with 10CFR50.92 and has concluded
that the revision does not involve a significant hazards
consideration (SHC). The basis for this conclusion is that the three
criteria of 10CFR50.92(c) are not satisfied. The proposed revision
does not involve an SHC because the revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The change to the Emergency Operating Procedures (EOP) to
eliminate the use of Recirculation Spray System (RSS) direct
injection during cold and hot leg recirculation does not effect the
probability of any accident. The elimination of the requirement to
have RSS directly [inject] into the reactor coolant system did not
increase the consequences of the previously evaluated accidents.
These consequences were evaluated based on very conservative
assumptions concerning the containment pressure after the design
basis Loss of Coolant Accident (LOCA), containment integrated
leakage rates, and the fraction of the sprayed volume. None of these
assumptions were affected by the elimination of the direct cold-leg
injection.
Therefore, the proposed revision does not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The modification to the RSS did not create the possibility of a
new or different accident from those previously analyzed. The change
involved elimination of the direct injection flow path from the
design basis of the system but did not involve physical
modifications to the system itself. The operability of the affected
valves within the direct injection alignments remained unchanged and
these paths were still available to the operators for contingencies
beyond the design basis. The EOPs provided clear and explicit
guidance to that effect.
Therefore, the proposed revision does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
In considering the impact on the margin of safety as defined in
the bases of the Technical Specifications, the impact of the change
on the design basis analysis of the fission product barriers must be
evaluated.
The minimum Emergency Core Cooling System flow requirement for
long-term core cooling is that the modified alignment deliver
sufficient flow to satisfy the inventory lost to the boil off in the
vessel due to the decay heat and the extended boiling from hot metal
in the downcomer and the lower plenum. The analysis determined that
these requirements were being met.
The elimination of the direct injection resulted in a flow
reduction through the RSS heat exchanger, from approximately 4000
gpm [gallons per minute] to 1200 gpm, thus reducing the rate of the
heat transfer from the containment to the service water system. The
design basis of the containment heat removal systems (circa 1986) is
that the containment pressure will decrease to subatmospheric within
one hour after the Design Basis Accident to compensate for the
reduction in heat removal from the containment, a smaller allowable
RSS pump degradation was assumed in the revised containment
analysis. The original RSS pump performance curve was based on a 10
percent reduction in developed head from the design curve. For the
modification, a 5 percent reduction was used. The results of the
analysis show that with these changes the design basis of
maintaining subatmospheric containment pressure was met.
Based on the above, elimination of the direct injection did not
reduce the margin of safety because there was no violation of the
acceptance limits and no weakening of the protective boundaries.
Therefore, the proposed revision does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: Phillip F. McKee.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of amendment requests: November 2, 1995, as supplemented by
letter dated January 9, 1998. The January 9, 1998, submittal supersedes
the staff's proposed no significant hazards consideration determination
evaluation for the requested changes that was published on April 10,
1996 (61 FR 15995).
Description of amendment requests: In the November 2, 1995, letter,
the
[[Page 14488]]
licensee proposed to revise Technical Specification (TS) 3.8.1, ``AC
Sources--Operating,'' to extend the offsite circuit completion time and
to extend the allowed outage time for an emergency diesel generator.
The January 9, 1998, letter modifies the original request to (1)
further extend the offsite completion time and allowed outage time for
an emergency diesel generator, and (2) add a new TS 5.5.2.14,
``Configuration Risk Management Program,'' that ensures a
proceduralized probabilistic risk assessment-informed process is in
place that assesses the overall impact of plant maintenance on plant
risk.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Emergency Diesel Generators (EDGs) are backup alternating
current power sources design to power essential safety systems in
the event of a loss of offsite power. EDGs are not accident
initiators in any accident previously evaluated. Therefore, this
change does not involve an increase in the probability of an
accident previously evaluated.
The EDGs provide backup power to components that mitigate the
consequences of accidents. The proposed changes to the Completion
Times do not affect any of the assumptions used in the deterministic
safety analysis.
To fully evaluate the effect of the EDG Completion Time
extension, Probabilistic Safety Analysis (PSA) methods were
utilized. The results of these analyses show no significant increase
in the core damage frequency. As a result, there would be no
significant increase in the consequences of accidents previously
evaluated.
The Configuration Risk Management Program is an Administrative
Program that assesses risk based on plant status. Adding the
requirement to implement this program for Technical Specification
3.8.1 does not affect the probability or the consequences of an
accident.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This proposed change does not alter the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the Limiting Conditions for
Operation or their Bases that are used in the deterministic analyses
to establish the margin of safety. PSA evaluations were used to
evaluate these changes and these evaluations determined that the
changes are either risk neutral or risk beneficial.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of amendment requests: December 19, 1997.
Description of amendment requests: The licensee proposed to revise
Technical Specification (TS) 3.4.9, ``Pressurizer,'' to reduce the
allowable pressurizer water volume for pressurizer operability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The limiting events impacted by this Technical Specification
change have been reanalyzed. These events are the Chemical and
Volume Control System (CVCS) Malfunction and CVCS Malfunction With a
Concurrent Single Failure of an Active Component, Inadvertent
Operation of the Emergency Core Cooling System (ECCS) During Power
Operation (Including Single Failure of an Active Component), and
Feedwater System Pipe Breaks. The probability of these events is not
changed by the restriction of the pressurizer level to 57%. An
operator action time of 15 minutes has been identified for the CVCS
malfunction and inadvertent ECCS operation events. Based on the
availability of operator alarms and indications and operator
Simulator training, 15 minute operator action is sufficient to
recognize and mitigate the inadvertent CVCS or ECCS operation.
Therefore, this change will not involve an increase in the
probability or consequences of any previously evaluated accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This amendment request does not involve any change to plant
equipment or operation. All the events identified in Chapter 15 of
the Updated Final Safety Analysis Report (UFSAR) were evaluated to
determine the impact of the change in pressurizer level. In addition
to the normally analyzed Inadvertent Operation of the ECCS During
Power Operation event a concurrent single failure of an active
component was considered in this evaluation. The analysis of this
event with single failure of an active component produced
consequences that are bounded by the CVCS malfunction with single
failure of an active component. No new or different kind of accident
will be created as a result of this Technical Specification change.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This amendment request does not change the manner in which
safety limits, limiting safety settings, or limiting conditions for
operation are determined. There are no changes to the acceptance
criteria for these events as a result of the proposed reduction in
the maximum pressurizer water level. This change does not reduce a
margin of safety since it lowers allowed pressurizer operational
level to 57%. An operator action time of 15 minutes has been
identified for the CVCS malfunction and inadvertent ECCS operation
events. Based on the availability of operator alarms and
indications, and demonstrated operator response in Simulator
training, 15 minute operator action has been demonstrated to be
adequate to recognize and mitigate the inadvertent CVCS or ECCS
operation. Therefore, this proposed change does 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
[[Page 14489]]
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of amendment requests: January 2, 1998.
Description of amendment requests: The licensee proposed to revise
Technical Specification (TS) 3.7.5, ``Auxiliary Feedwater (AFW)
System,'' to indicate the turbine driven AFW pump is operable when
running in the manual mode to support plant startups, shutdowns, and
testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Probabilistic analyses have been performed in support of
declaring P140 operable when the pump is manually actuated and
operating.
The results show that, considering P-140 to be in test for an
entire year, the core damage risk of a Main Steam Line Break/
Feedwater Line Break (MSLB/FWLB) slightly increases (4.3E-8/yr)
while the risk due to other initiating events decreases (3E-7/yr).
The net core damage impact of P-140 in test for an entire year is a
Core Damage Frequency (CDF) decrease of 2E-7/yr. Having P140
operating instead of being in standby increases its reliability.
This increased reliability reduces the risk due to other initiating
events, such as loss of main feedwater, medium and small Loss of
Coolant Accidents (LOCAs), Steam Generator Tube Rupture (SGTR), and
Loss of Offsite Power (LOP), which require Auxiliary Feedwater (AFW)
and which occur with much greater frequency than MSLB/FWLB. With the
overall CDF reduction a result of considering P140 being in a test
configuration for an entire year, the actual cumulative risk
incurred is the weighted fraction that P140 is in the test
configuration over a year period. Based on past experience, the pump
is running in manual approximately 500 minutes/year, which results
in an annual net cumulative CDF reduction on the order of 2E-10/yr
due to running P140 in the manual mode.
Therefore, the operation of the facility in accordance with this
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This change does not involve a plant hardware modification or
allow the operation of any plant equipment in any way other than
originally designed. This change only affects the administrative
tracking of the turbine-driven AFW pump when the steam driven AFW
pump is operating in the manual mode.
Therefore, the operation of the facility in accordance with this
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Pump history shows the pump is run approximately 500 minutes per
year. In all cases except for the one postulated scenario of the
Main Steam Isolation Signal followed by an Emergency Feedwater
Actuation Signal the turbine-driven AFW pump is not susceptible to
being tripped. Also, this postulated scenario does not affect the
capability of the motor-driven AFW pumps.
Even though there is a small increase in the CDF from the AFW
steam driven pump operating in manual mode based on the possibility
of a MSLB/FWLB, also considering other initiating events results in
an annual net cumulative CDF reduction on the order of 2E-10/yr due
to P140 running in the manual mode.
Therefore, the operation of the facility in accordance with this
proposed change does not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear
Date of amendment request: August 20, 1997, as supplemented by
letters dated September 18, 1997 and October 31, 1997.
Description of amendment request: The proposed change would revise
the Vermont Yankee Technical Specifications Section 6.0,
``Administrative Controls,'' to add and revise reference to NRC-
approved methodologies which will be used to generate the cycle-
specific thermal operating limits in the Vermont Yankee Core Operating
Limits Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change will not involve any significant increase
in the probability or consequences of an accident previously
evaluated.
The change updates the Technical Specifications to include an
NRC approved method reference to allow calculation of thermal limits
with a revised method. It does not affect plant operation and will
not weaken or degrade the facility.
2. The proposed change will not create the possibility of a new
or different kind of accident since the change is administrative. No
physical alterations of the plant, setpoint changes, or operating
conditions are proposed.
3. The proposed change will not involve a significant reduction
in a margin of safety. The change involves an update to the
Administrative Controls in Section 6.0 of the Technical
Specifications by adding a reference to NRC approved methods. This
administrative change does not alter plant safety margins.
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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, N.W., Washington, DC 20037-1128.
NRC Project Director: Cecil O. Thomas, Director.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 4, 1998.
Description of amendment request: The amendment would revise
Technical Specification 3.2.4, quadrant power tilt ratio (QPTR), and
associated Bases, to clarify the required actions for the limiting
condition for operation (LCO) and other changes consistent with the
technical specification conversion application submitted by letter
dated May 15, 1997.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
[[Page 14490]]
1. Requirements for Determining QPTR
The Action to calculate QPTR once per hour until THERMAL POWER
was reduced to less than 50% RATED THERMAL POWER (RTP) when QPTR
exceeds the LCO requirements would be deleted and replaced by a new
requirement to determine QPTR at least once per 12 hours.
The proposed change involves only the compensatory measures to
be taken should the QPTR be outside its limit. The frequency with
which QPTR is calculated is not assumed in the initiating events for
any accident previously evaluated. In addition, the change does not
involve any new operating activities or hardware change. Therefore,
the proposed change would not significantly increase the probability
of an accident previously evaluated.
Once THERMAL POWER has been reduced appropriately in proportion
to the amount that QPTR exceeds 1.00, any additional change would be
sufficiently slow that a 12-hour interval for recalculating QPTR
will provide an adequate level of protection. Therefore, the
proposed change will not significantly increase the consequences of
any accident previously evaluated.
2. Completion Time for Resetting the Power Range Neutron Flux-High Trip
Setpoints
The proposed change to allow 72 hours for resetting the Power
Range Neutron Flux-High trip setpoints involves only the
compensatory measures to be taken should the QPTR be outside its
limit. These compensatory measures are not assumed in the initiating
events for any accident previously evaluated. The proposed actions
recognize that the required reduction in power (3% for each 1% of
indicated QPTR in excess of 1.00) provide adequate margin for fuel
design limits so that consequences of assumed accidents would not be
significantly affected. Therefore, the proposed change will not
adversely affect the probability or consequences of any accident
previously evaluated. Further, by permitting more time to perform
resetting the trip setpoints, the chances of a transient may be
reduced.
3. Delete(tion) of the Actions (a.3., a.4.) for verifying QPTR to be
restored within 24 hours and for identifying and correcting the cause
of the out-of-limit condition prior to increasing THERMAL POWER
The proposed changes would delete current Actions a.3. and a.4.
and add new Actions for QPTR out of limit including requirements for
measuring FQ(Z) and F N delta H prior to and
following a return to power and performing safety analyses to verify
safety requirements are met prior to increasing power above the
limits of Action a.1. The proposed changes involve only the
compensatory measures to be taken should the QPTR be outside its
limit. These compensatory measures are not assumed in the initiating
events for any accident previously evaluated. Therefore, the
proposed change will not affect the probability or consequences of
any accident previously evaluated.
4. Deletion of the Actions for QPTR in excess of 1.09
The proposed change would delete the required Actions for QPTR
in excess of 1.09 and Actions for QPTR in excess of 1.02 are
followed for all instances where QPTR exceeds 1.02. The proposed
change involves only the compensatory measures to be taken should
the QPTR be outside its limit. These compensatory measures are not
assumed in the initiating events for any accident previously
evaluated. The proposed actions recognize that the required
reduction in power (3% for each 1% of indicated QPTR in excess of
1.00) provide adequate margin for fuel design limits so that
consequences of assumed accidents would not be significantly
affected. Therefore, the proposed change will not affect the
probability or consequences of any accident previously analyzed.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
1. Requirements for Determining QPTR
The proposed change for calculating QPTR once every 12 hours
does not involve a physical alteration to the plant or change the
method by which any safety-related system performs its function. The
manner in which the plant would be operated would not be altered.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any previously evaluated.
2. Completion Time for Resetting the Power Range Neutron Flux-High Trip
Setpoints
The proposed change to allow 72 hours for resetting the Power
Range Neutron Flux-High trip setpoints does not involve a permanent
physical alteration to the plant; no new or different kinds of
equipment will be installed. The change would not alter the manner
in which the plant would be operated only the timing of actions that
provide potential mitigation of accidents. Thus, the change would
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Delete the Actions (a.3., a.4.) for verifying QPTR to be restored
within 24 hours and for identifying and correcting the cause of the
out-of-limit condition prior to increasing THERMAL POWER
The proposed changes would delete current Actions a.3, and a.4.
and add new Actions for QPTR out-of-limit including requirements for
measuring FQ(Z) and F N delta H prior to and
following a return to power and performing safety analyses to verify
safety requirements are met prior to increasing power above the
limits of Action a.1. The proposed changes do not involve a physical
alteration to the plant; no new or different kinds of equipment
would be installed. The changes would not alter the manner in which
the plant would be operated only the timing of actions that provide
potential mitigation of accidents. Thus, the changes would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
4. Deletion of the Actions for QPTR in excess of 1.09
The proposed change would delete the required Actions for QPTR
in excess of 1.09 and Actions for QPTR in excess of 1.02 are
followed for all instances where QPTR exceeds 1.02. The proposed
change does not involve a physical alteration to the plant or
changes in the way in which the plant is operated. The proposed
change involves only the compensatory measures to be taken should
QPTR be outside its limit. The assumptions of the accident analyses
are unaffected by the proposed change. No new permutations or event
initiators are introduced by the proposed alternate methods of
dealing with QPTRs in excess of 1.09. Therefore, there is no
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
1. Requirements for Determining QPTR
The proposed change for calculating QPTR once every 12 hours
does not change any accident analysis assumptions, initial
conditions or results. The proposed change will continue to ensure
that the plant is maintained in a safe condition while QPTR is in
excess of its limit. Additionally, calculating QPTR once per 12
hours as opposed to every hour while QPTR is in excess of its limit
would avoid the diversion of personnel resources from corrective
actions with regard to meeting the LCO. Therefore, the proposed
change will not involve a significant reduction in any margin of
safety.
2. Completion Time for Resetting the Power Range Neutron Flux-High Trip
Setpoints
The proposed change to allow 72 hours for resetting the Power
Range Neutron Flux-High trip setpoints will continue to ensure that
the plant is maintained in a safe condition within the envelope of
the safety analyses while QPTR is in excess of its limit. The
proposed actions recognize that the required reduction in power (3%
for each 1% of indicated QPTR in excess of 1.00) provide adequate
margin for fuel design limits so that consequences of assumed
accidents would not be significantly affected. Therefore, the
proposed change will not involve a significant reduction in any
margin of safety.
3. Delete the Actions (a.3., a.4.) for verifying QPTR to be restored
within 24 hours and for identifying and correcting the cause of the
out-of-limit condition prior to increasing THERMAL POWER
The proposed changes would delete current Actions a.3. and a.4
and add new Actions for QPTR out-of-limit including requirements for
measuring FQ(Z) and F N delta H prior to and
following a return to power and performing safety analyses to verify
safety requirements are met prior to increasing power above the
limits of Action a.1. The proposed changes will continue to ensure
that the plant is maintained in a safe condition within the envelope
of the safety analysis while QPTR is in excess of its limit.
Therefore, the proposed changes will not involve a significant
reduction in any margin of safety.
4. Deletion of the Actions for QPTR in excess of 1.09
The proposed change would delete the required Actions for QPTR
in excess of 1.09
[[Page 14491]]
and Action for QPTR in excess of 1.02 are followed for all instances
where QPTR exceeds 1.02. The proposed change will continue to ensure
that the plant is maintained in a safe condition within the envelope
of the safety analyses while QPTR is in excess of its limit. While
different actions are taken in response to a QPTR in excess of 1.09,
the proposed change will assure that accident analyses assumptions
continue to be met. Therefore, the proposed changes will not involve
a significant reduction in any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 4, 1998.
Description of amendment request: The amendment would revise the
technical specifications to (1) create separate functional units for
the analog and digital portions of the engineered safety features
actuation system (ESFAS) function associated with starting the turbine-
driven auxiliary feedwater pump on a loss of offsite power, and (2) add
a table notation to clarify that the testing of the time delay relays
for the 4 kV undervoltage, loss of voltage and grid degraded voltage
portion of the ESFAS is performed as part of the channel calibration.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since no
hardware changes are proposed. The recognition that different
OPERABILITY and surveillance requirements apply to analog vs.
digital circuitry does not impact any previously analyzed accidents.
The clarification that testing of the time delay relays is performed
as part of the CHANNEL CALIBRATION does not impact any previously
analyzed events. The proposed change will not affect any of the
analysis assumptions for any of the accidents previously evaluated.
The proposed change does not alter the current method or procedures
for meeting the surveillance requirements in Table 4.3-2. The
proposed change will not affect the probability of any event
initiators nor will the proposed change affect the ability of any
safety-related equipment to perform its intended function. There
will be no degradation in the performance of nor an increase in the
number of challenges imposed on safety-related equipment assumed to
function during an accident situation. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. The separation of analog and digital portions of
Functional Unit 6.f or the clarification of testing of the time
delay relays will not impact the normal method of plant operation.
The OPERABILITY requirements, ACTION Statement, and surveillance
requirements for the analog portion, new Functional Unit 6.f.1), are
identical to those of Functional Unit 8.a, while the requirements
for the digital portion, new Functional Unit 6.f.2), are consistent
with the current technical specifications, other than the new ACTION
Statement 30 provisions that defer to the TDAFW pump Specification
3.7.1.2 requirements and the performance of a TADOT during
appropriate plant conditions. These changes do not change any ESFAS
design standard and are appropriate for digital functions such as
this.
Testing of the time delay relays has been performed as part of
the 18 month CHANNEL CALIBRATION. The tolerancesfor the time delay
relays are sufficient to account for relay drift encountered during
the 18 month surveillance testing. The calculated tolerances for the
time delay setpoints have been evaluated to insure that safety-
related systems, subsystems and components would not be adversely
affect[ed] by the drift within the permissible tolerance band.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this change. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not affect the acceptance criteria for
any analyzed event. There will be no effect on the manner in which
safety limits or limiting safety system settings are determined nor
will there be any effect on those plant systems necessary to assure
the accomplishment of protection functions. There will be no impact
on any margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Previously Published Notices of Consideration of Issuance of
Amendments toFacility 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.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: January 26, 1998.
Brief description of amendment request: The proposed amendment
would change the SSES Technical Specifications facility staff
requirements to allow an individual who does not hold a current senior
reactor operator (SRO) license to hold the position of Manager-Nuclear
Operations (MNO) and require an individual serving in the capacity of
the Operations Supervisor-Nuclear to hold a current SRO license
[[Page 14492]]
and report directly to the MNO and be responsible for directing the
licensed activities of licensed operators.
Date of publication of individual notice in Federal Register:
February 24, 1998 (63 FR 9270).
Expiration date of individual notice: March 26, 1998.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: February 25, 1998, TXX-98050.
Description of amendment request: The proposed amendment would be a
temporary change to the Technical Specifications to remove the
requirement to demonstrate the load shedding feature of MCC XEB4-3 as
part of Surveillance Requirements (SRs) 4.8.1.1.2f.4)a) and
4.8.1.1.2f.6)a) until the plant startup subsequent to the next
refueling outage or until an outage of greater than 24 hours in
duration for each respective unit. This temporary change is requested
as a result of the failure to confirm the load shedding feature of MCC
XEB4-3 during the last performance of these SRs for the Unit 1 and Unit
2 train B diesel generators (DGs).
Date of individual notice in the Federal Register: March 9, 1998,
(63 FR 11458).
Expiration date of individual notice: April 8, 1998.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: March 18, 1997, as supplemented
by letters dated July 28, 1997, and September 9, 1997.
Brief description of amendments: The amendments revise the
operating licenses to reflect approval of Amendment 42 to the Palo
Verde Nuclear Generating Station Physical Security Plan. The amendments
revise the methods used to search materials, packages, and personnel
prior to their entry into the protected area, as described in the
security plan.
Date of issuance: March 4, 1998.
Effective date: March 4, 1998.
Amendment No.: Unit 1-115; Unit 2-108; Unit 3-87.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the operating licenses.
Date of initial notice in Federal Register: October 8, 1997 (62 FR
52580).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 4, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: December 17, 1997, as
supplemented by letters dated February 6, 1998 and March 12, 1998.
Brief description of amendment: The proposed change would revise
Technical Specifications Section 5.6.5, ``Core Operating Limits
Report.'' The revisions add reference to an additional approved
methodology for correlating departure from nucleate boiling (DNB)
ratios. The added methodology is the Siemens Power Corporation Topical
Report, EMF-92-153(P)(A), ``HTP: Departure from Nucleate Boiling
Correlation for High Thermal Performance Fuel.''
Date of issuance: March 16, 1998.
Effective date: March 16, 1998.
Amendment No. 178.
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4309). The February 6 and March 12, 1998 submittals provided clarifying
information that did not affect the initial determination of no
significant hazards considerations. The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated March 16,
1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois, Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendments: October 27, 1997.
Brief description of amendments: The amendments would change the
Dresden and Quad Cities Technical Specifications (TS) to clarify the
applicability, action and surveillance requirements for the Standby
Liquid Control System (SLCS). The changes would make the current TS
requirements for the SLCS consistent with the Improved Standard
Technical Specifications (ISTS) contained in NUREG-1433, ``Standard
Technical Specifications General Electric Plants, BWR/4.''
Date of issuance: March 6, 1998.
[[Page 14493]]
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 167, 162, and 180, 178.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 14, 1998 (63 FR
2277).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 6, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: for Dresden, Morris Area
Public Library District, 604 Liberty Street, Morris, Illinois 60450;
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,
Illinois 61021.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
PointNuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: October 2, 1996, as supplemented
July 31, 1997.
Brief description of amendment: The amendment revises Figures
3.1.A-1, 3.1.A-2 and 3.1.A-3, Section 3.1.B and its Bases, Figures
3.1.B-1 and 3.1.B-2, and the Bases of Section 4.3 and Figure 4.3-1 of
the Technical Specifications to incorporate the revised Indian Point
Unit 2 Heatup and Cooldown Limit Curves for Normal Operation.
Date of issuance: February 27, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 195.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 19, 1996 (61
FR 58901).
The July 31, 1997, letter provided clarifying information that did
not change the initial proposed no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 27, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: December 17, 1997Brief
description of amendments: The amendments revise Section 6.9.1.9 of the
Technical Specifications to reference updated or recently approved
topical reports, which contain methodologies used to calculate cycle-
specific limits contained in the Core Operating Limits Report.
Date of issuance: March 2, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1-163; Unit 2-155.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4310).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: July 21, 1997, as supplemented
February 18, 1998.
Brief description of amendment: Technical Specification Change
Request concerning Emergency Feedwater Surveillance Testing. This
request is to make several changes to the ANO-2 Technical
Specifications including extension of the emergency feedwater (EFW)
pump surveillance testing frequency, a reduction in the minimum steam
generator pressure required to perform the surveillance testing on the
turbine-driven EFW pump, and a modification to the EFW pump testing
requirements.
Date of issuance: March 12, 1998.
Effective date: March 12, 1998.
Amendment No.: 188.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: August 13, 1997 (62 FR
43367).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: September 23, 1997, as
supplemented by letters dated February 27 and March 4, 1998.
Brief description of amendment: The amendment changes the Reactor
Protective System (RPS) and Engineering Safety Actuation System (ESFAS)
trip set point and allowable values for steam generator low pressure.
The amendment also relocates the RPS and ESFAS response time tables
from the Technical Specifications to the Safety Analysis Report as
described in NRC Generic Letter 93-08, ``Relocation of Technical
Specification Tables of Instrument Response Time Limits,'' dated
December 29, 1993.
Date of issuance: March 12, 1998.
Effective date: March 12, 1998.
Amendment No.: 189.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: January 28, 1998, (63
FR 4311).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: September 23, 1997, as
supplemented by letters dated February 27 and March 4, 1998.
Brief description of amendment: The amendment reduces the minimum
required reactor coolant system flow rate in TS 3.2.5 until the ANO-2
steam generators are replaced. The reduced reactor coolant system flow
requirement will account for plugging of up to approximately 30 percent
of the tubes in the existing steam generators at ANO-2.
Date of issuance: March 12, 1998.
Effective date: March 12, 1998.
Amendment No.: 190.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications/license.
Date of initial notice in Federal Register: January 28, 1998, (63
FR 4312).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 1998.
[[Page 14494]]
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: December 5, 1997, as
supplemented December 11, 1997, January 9, February 12 and 19, 1998.
Brief description of amendment: To revise the Final Safety Analysis
Report (FSAR) and the Improved Technical Specification Bases to reflect
the modified reactor building fan recirculation system fan cooler
starting logic.
Date of issuance: March 9, 1998.
Effective date: March 9, 1998.
Amendment No.: 165.
Facility Operating License No. DPR-31: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 15, 1998 (63 FR
2423). The supplemental letters dated December 11, 1997, January 9,
February 12 and 19, 1998, did not change the original no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida.
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Project Director: Frederick J. Hebdon.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: May 14, 1997, as supplemented
by letter dated October 9, 1997 (published in Federal Register as May
15, 1997).
Brief description of amendments: The amendments revised the
combined Technical Specifications (TS) for the Diablo Canyon Power
Plant (DCPP) Unit Nos. 1 and 2 to revise the surveillance frequencies
from at least once every 18 months to at least once per refueling
interval (nominally 24 months) including (1) reactor coolant system
total flow rate, (2) instrumentation for radiation monitoring, (3)
instrumentation and controls for remote shutdown, (4) instrumentation
for accident monitoring, and (5) several miscellaneous TS.
Date of issuance: February 27, 1998.
Effective date: February 27, 1998, to be implemented within 90 days
of the date of issuance.
Amendment Nos.: Unit 1-123; Unit 2-121.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 30, 1997 (62 FR
40855).
The October 9, 1997, supplemental letter provided additional
clarifying information and did not change the staff's initial no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 27, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
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: January 2, 1997, as supplemented
November 13, 1997.
Brief description of amendment: The amendment changes the Technical
Specifications by extending the surveillance interval for the
functional testing of certain Inservice Inspection American Society of
Mechanical Engineers Code Class 1, 2, and 3 pumps and valves from once
a month to once a quarter.
Date of issuance: March 2, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 178.
Facility Operating License No. DPR-64: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 26, 1997 (62 FR
14468).
The November 13, 1997, submittal contained clarifying information
that did not change the staff's proposed finding of no significant
hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 2, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: December 14, 1995, as
supplemented September 26, 1997.
Brief description of amendment: The amendment changes the James A.
FitzPatrick Technical Specifications (TSs) to incorporate the inservice
testing requirements of Section XI of the American Society of
Mechanical Engineers Boiler and Pressure Vessel Code. The amendment
supplements Amendment No. 241, dated December 2, 1997, by issuing seven
TS pages inadvertently omitted from Amendment No. 241.
Date of issuance: February 27, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 242.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1635).
The September 26, 1997, letter provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 27, 1998.
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.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: December 15, 1997.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) to adopt Option B, of 10 CFR Part 50,
Appendix J, ``Primary Reactor Containment Leakage Testing for Water-
Cooled Power Reactors,'' to implement a performance-based approach for
Type B and C testing. Additionally, the wording in the TSs would be
modified for the previous adoption of Option B on Type A testing and a
section added on the primary
[[Page 14495]]
containment leakage rate testing program.
Date of issuance: February 27, 1998.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos: 207 and 188.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 14, 1998 (63 FR
2281).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 27, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: December 17, 1997.
Brief description of amendments: The amendments extended the
surveillance interval of the containment spray nozzle air flow test to
ten years from five years.
Date of issuance: March 11, 1998.
Effective date: March 11, 1998.
Amendment Nos.: Unit 1--Amendment No. 94; Unit 2--Amendment No. 81.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4325).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 11, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, OES Nuclear, Inc.,
Pennsylvania Power Company, Toledo Edison Company, Docket No. 50-440
Perry Nuclear Power Plant, Unit 1, Lake County, Ohio.
Date of application for amendment: December 23, 1997.
Brief description of amendment: This amendment revised Technical
Specification 3.8.1, ``A.C. Sources--Operating,'' consistent with the
recommendations in NRC Generic Letter 94-01, ``Removal of Accelerated
Testing and Special Reporting Requirements for Emergency Diesel
Generators.''
Date of issuance: March 12, 1998.
Effective date: March 12, 1998.
Amendment No.: 92.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4326).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 12, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: July 11, 1997, as supplemented
November 21, December 22, 1997, and February 6, 1998.
Brief description of amendment: The amendment revised Technical
Specifications 3.7/4.7 and their associated Bases to incorporate Option
B of Appendix J to 10 CFR 50, and editorial changes to TS Table 4.7.2
Date of Issuance: February 26, 1998.
Effective date: February 26, 1998, with full implementation within
30 days.
Amendment No.: 152.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: (62 FR 45465). The
November 21, December 22, 1997, and February 6, 1998, letters did not
change the initial proposed no significant hazards determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 26, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: November 20, 1997.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.10 and its associated Bases to eliminate the use
of battery charger AB for meeting the requirement of the TS.
Date of issuance: March 5, 1998.
Effective Date: This license amendment is effective as of its date
of issuance, to be implemented within 30 days.
Amendment No.: 153
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 31, 1997 (62
FR 68319).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 5, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: August 22, 1997, as supplemented
by letter dated September 18 and October 31, 1997.
Brief description of amendment: The amendment revises the Technical
Specifications to address the new low pressure CO2 suppression system
for the East and West Switchgear Rooms and more clearly describes the
separation of the two rooms.
Date of Issuance: March 6, 1998.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment No.: 154.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 8, 1997 (62 FR
52590). Information provided by letter dated October 31, 1997, did not
affect the original no significant hazards consideration.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated March 6, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following
[[Page 14496]]
amendments. The Commission has determined for each of these amendments
that the application for the amendment complies with the standards and
requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations. The Commission has made
appropriate findings as required by the Act and the Commission's rules
and regulations in 10 CFR Chapter I, which are set forth in the license
amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By April 24, 1998, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
[[Page 14497]]
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388,
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: February 5, 1998, as
supplemented February 12, March 3 and 5, 1998.
Brief description of amendments: The amendments revised the
surveillance requirements in Technical Specification (TS) 4.6.1.2
(Requirement a). The change to the referenced TS adds a footnote
stating that the requirement for Type A testing will not apply to
certain instrument line penetrations.
Date of issuance: March 10, 1998.
Effective date: Both units, as of the date of issuance.
Amendment Nos.: 173 and 146.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No. On February 5, 1998, the staff issued a Notice of
Enforcement Discretion, which was immediately effective and remained in
effect until this amendment was issued.
The Commission's related evaluation of the amendments, finding of
emergency circumstances, consultation with the State of Pennsylvania,
and final no significant hazards consideration determination are
contained in a Safety Evaluation dated March 10, 1998.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Dated at Rockville, Maryland, this 18th day of March 1998.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-7652 Filed 3-24-98; 8:45 am]
BILLING CODE 7590-01-P
