[Federal Register Volume 59, Number 168 (Wednesday, August 31, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-21325]
[[Page Unknown]]
[Federal Register: August 31, 1994]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from August 8, 1994, through August 19, 1994. The
last biweekly notice was published on August 17, 1994 (59 FR 42332).
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By September 30, 1994, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.714 which
is available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and
to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
room for the particular facility involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: June 9, 1994
Description of amendment request: The proposed amendment would
increase the allowed out-of-service time to increase from 7 days to 14
days for the automatic depressurization system (ADS), the high pressure
coolant injection (HPCI) system and the reactor core isolation cooling
(RCIC)system. The proposed change includes a change to Section 4.5.H,
``Maintenance of Filled Discharge Pipe'' to reflect Amendment 149,
issued September 28, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The Operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Safety criteria used to determine the acceptability of extending
continued operation with one ADS valve, the HPCI or RCIC system out-
of-service (OOS) is consistent with Pilgrim's licensing basis. For
example, events with the expected frequency of occurrence greater
than once-per-reactor lifetime are required to meet the transient
MCPR [minimum critical power ratio] thermal limit: more than 99.9%
of the fuel rods are expected to avoid boiling transition. Very low
probability events, such as a LOCA [loss-of-coolant accident], are
required to satisfy the criteria of 10CFR50.46: the primary
criterion being that the Peak Cladding Temperatures (PCT) be
maintained less than 2200 deg.F.
For intermediate frequency events, e.g. safe shutdown in the
event of a fire, 10CFR50 Appendix R involves a ``no fuel damage''
criterion. To evaluate these types of events, the GE [General
Electric] SAFER/GESTR-LOCA licensing methodology was used to
calculate the system responses and PCTs.
Analyses performed by Pilgrim's NSSS [nuclear steam supply
system] vendor, General Electric, [***] for various limiting-case
scenarios involving ADS, HPCI, or RCIC out-of-service situations
demonstrated 10CFR50.46 limits (i.e. a PCT less than 2200 deg.F)
were met. (The most severe PCT was 1500 deg.F). The core damage
frequency analysis for Pilgrim is unchanged by operating Pilgrim in
accordance with this proposed amendment. The 14 day OOS for HPCI,
RCIC and ADS also conforms to the 00S time for these systems found
in BWR [boiling-water reactor] Standard Technical Specifications.
Hence, increasing the allowed 00S time from 7 to 14 days does not
result in a challenge to fuel cladding integrity or BWR Standard
Technical Specifications, and operating Pilgrim in accordance with
the proposed amendment will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The removal of the association between LPCI [low-pressure
coolant injection] and Core Spray system testing and surveilling
their filled discharge pipes is an administrative change because the
specified surveillance frequency is unchanged. This proposed change
reflects Amendment 149, issued by the NRC September 28,
1993, and is proposed to ensure consistency between Pilgrim's
Technical Specification sections. This administrative change will
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
As discussed above, a variety of limiting-case scenarios were
analyzed to demonstrate the effects of increasing the 00S time for
one ADS valve, the HPCI system, or the RCIC system. The conclusion
of the analyses is that this proposed change does not violate
Pilgrim's licensing basis or 10CFR50.46 requirements.
Some scenarios result in elevated PCTs, but they are still
significantly below the 10CFR50.46 limit of 2200 deg.F. Therefore,
since the licensing-basis and code required PCT continues to be met
and because the proposed change comports the requirements of
BWR Standard Technical Specifications, operating Pilgrim in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
As discussed in above question 1, the proposed change to section
4.5.H.1 is administrative and does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
Certain scenarios analyzed for system unavailability result in
evaluated PCTs. However, these elevated PCTs are significantly below
the 10CFR50.46 limit of 2200 deg.F. Therefore, there is no reduction
in the safety margin for PCT resulting from the change from 7 to 14
days. The proposed change also corresponds to the requirements of
BWR Standard Technical Specifications concerning 00S for HPCI, RCIC
and ADS. Therefore, operating Pilgrim Station in accordance with
this proposed 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 standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Walter R. Butler
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County,North Carolina
Date of amendment request: July 22, 1994
Description of amendments request: The proposed amendment would
implement a performance based assessment program, including
corresponding organizational and functional changes. Specifically, the
changes affect the Independent Review (IR) function, the independent
assessment of plant activity and the Independent Safety Engineering
Group. These functions will be performed by the proposed Nuclear
Assessment Section (NAS). The NAS would perform internal evaluations
and assessment activities and serve as plant management's staff for the
objective oversight of plant performance relating to nuclear safety,
reliability, and quality. The NAS's fundamental role will be to: (1)
assist plant management in the early identification of issues which may
prevent the plant from achieving quality performance on a sustained
basis; and (2) ensure effective correction of deficiencies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated because it is a programmatic and administrative
change which does not physically alter any safety-related systems,
nor does it affect the way in which any safety-related systems
perform their functions. Since the design of the facility and system
operating parameters are not changing, the proposed amendment does
not involve an increase in the probability or consequences of any
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. As stated in Item 1, the proposed amendment is a
programmatic and administrative change which does not physically
alter any safety-related systems; nor does it affect the way in
which any safety-related systems perform their functions. Since the
design of the facility and system operating parameters are not
changing, the proposed amendment 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 because it is a programmatic and
administrative change which provides assurance that plant operations
continue to be conducted in a safe manner through the performance
based assessment programs. As stated in Item 1, the proposed
amendment does not physically alter any safety-related systems; nor
does it affect the way in which any safety-related systems perform
their functions. Since the design of the facility and system
operating parameters are not changing, the proposed amendment does
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: July 22, 1994
Description of amendment request: The proposed amendment would
implement a performance based assessment program, including
corresponding organizational and functional changes. Specifically, the
changes affect the Independent Review (IR) function, the independent
assessment of plant activity and the Independent Safety Engineering
Group. These functions will be performed by the proposed Nuclear
Assessment Section (NAS). The NAS would perform internal evaluations
and assessment activities and serve as plant management's staff for the
objective oversight of plant performance relating to nuclear safety,
reliability, and quality. The NAS's fundamental role will be to: (1)
assist plant management in the early identification of issues which may
prevent the plant from achieving quality performance on a sustained
basis; and (2) ensure effective correction of deficiencies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated because it is a programmatic and administrative
change which does not physically alter any safety-related systems,
nor does it affect the way in which any safety-related systems
perform their functions. Since the design of the facility and system
operating parameters are not changing, the proposed amendment does
not involve an increase in the probability or consequences of any
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. As stated in Item 1, the proposed amendment is a
programmatic and administrative change which does not physically
alter any safety-related systems; nor does it affect the way in
which any safety-related systems perform their functions. Since the
design of the facility and system operating parameters are not
changing, the proposed amendment 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 because it is a programmatic and
administrative change which provides assurance that plant operations
continue to be conducted in a safe manner through the performance
based assessment programs. As stated in Item 1, the proposed
amendment does not physically alter any safety-related systems; nor
does it affect the way in which any safety-related systems perform
their functions. Since the design of the facility and system
operating parameters are not changing, the proposed amendment does
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: July 28, 1994
Description of amendment request: The proposed amendment would
change TS Sections 5.3.1.3, 5.4.2.1, 5.4.22, and the Section 5
references to allow the use of fuel enriched to 4.95 plus 0.05 weight
percent (w/o) U235.
The proposed license change is required to support delivery of
reload batch enrichments anticipated for Cycle 17 and beyond. These
reloads will require the use of fuel enrichments exceeding the current
TS limit of 4.20 plus 0.05 weight percent (w/o) U235 (nominal
4.20).
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. Increasing the allowable U235 enrichment
will have no influence on the probability of an accident previously
evaluated. No changes will be made to any safety related equipment,
systems, or setpoints used in determining the probability of an
evaluated accident. Neither will the proposed amendment allow
operation of the facility or safety equipment outside applicable
limitations or restrictions. Plant design bases will not be altered.
With respect to the Fuel Handling Accident, the manner in which the
fuel is handled will not be altered. The heat load on the Spent Fuel
Pool will not be increased and the cooling and circulation systems
and equipment will be unaltered. Therefore, there will be no
significant increase in the probability of an accident previously
evaluated.
The proposed change does not increase maximum allowable burnup
or fission product inventory. Since fission product inventory is an
inconsequential function of enrichment, radiological consequences
evaluated in the Updated Final Safety Analysis Report (UFSAR) will
not increase. The proposed change will not alter the function of
safety related equipment designed to mitigate the consequences of an
accident previously evaluated or allow operation of the facility
outside applicable limitations or restrictions. Accordingly the
proposed change will not involve a significant increase in the
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 increase in allowable enrichment will not
result in any design, operation, or function changes to any safety
related equipment designed to prevent and/or mitigate accidents, to
any setpoints or systems, or to any portion of the plant design
basis. Operation of the facility will remain within all required
limitations and restrictions. With respect to the Fuel Handling
Accident, the manner in which the fuel is handled will not be
altered. The heat load on the Spent Fuel Pool will not be increased
and the cooling and circulation systems and equipment will be
unaltered. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety. NRC acceptance criteria and thus
the acceptable margin of safety to criticality for the Spent Fuel
Pool and New Fuel Storage Vault criticality are defined in Section
5.0 of the Technical Specifications. For the Spent Fuel Pool the
criteria specify that Keff must be maintained less than 0.95 when
the pit is flooded with unborated water. For the New Fuel Storage
Vault, the Keff must remain less than 0.95 if the vault is flooded
with unborated water, and must remain below 0.98 in an optimum
moderation event. Analyses performed in support of the proposed
change demonstrate that these acceptance criteria will continue to
be met. With respect to radiological consequences, the margin of
safety is defined by 10 CFR [Part] 100 limits which will not be
challenged. The analyses conclude that fission product inventory and
thus radiological consequences reported in Chapter 15 of the UFSAR
will not change. Accordingly the proposed license amendment will not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: July 29, 1994
Description of amendment request: The amendment would allow
operation of the plant with one Emergency Diesel Generator (EDG)
inoperable without entering a condition prohibited by Section 3.0 of
the Technical Specifications (TS). This TS request includes provisions
to avoid testing the operable EDG altogether under certain conditions
to ensure that one EDG is available to provide emergency power, if
needed, and to preserve the EDG overall life and reliability.
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. The proposed change involves changes in the
testing frequency of the EDGs when one EDG is inoperable, as well as
provision of additional measures to ensure that a source of off-site
power is available. The proposed change will also avoid testing of
an EDG when one EDG is inoperable if the EDG became inoperable for
reasons other than a common cause. Since the changes involve the
EDGs which perform an accident mitigation function and are not
involved in any accident initiation sequence, there is no
significant increase in the probability of a previously analyzed
accident. Since the changes involve the EDGs which perform an
accident mitigation function, and the changes provide additional
assurance that emergency power will be available for accident
mitigation, [there] is no significant increase in the consequences
of a previously analyzed accident. 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 change involves changes in the testing
frequency of the EDGs when one EDG is inoperable, as well as
provision of additional measures to ensure that a source of off-site
power is available. The proposed change will also avoid testing of
an EDG when one EDG is inoperable if the EDG became inoperable for
reasons other than a common cause. Since these changes do not
involve changes in the operation of the plant, or physical or
equipment changes and involve controls for accident mitigation
equipment, the proposed amendment will not created the possibility
of new or different kind of accident from any accident previously
evaluated. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety. The proposed change involves
changes in the testing frequency of the EDGs when one EDG is
inoperable, as well as provision of additional measures to ensure
that a source of off-site power is available. The proposed change
will also avoid testing of an EDG when one EDG is inoperable if the
EDG became inoperable for reasons other than a common cause. The
change reduces the required testing frequency of an operable EDG,
hence reducing time that no EDG will be available for automatic
starting and loading. These changes will provide assurance that
emergency power will be available to mitigate the effects of any
accident and will prevent excessive wear on the EDGs. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
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: July 22, 1993
Description of amendment request: The proposed amendment would
allow implementation of a performance based assessment program and the
corresponding functional and organizational changes in the Nuclear
Assessment Department. The changes affect the independent review
function, the independent assessment of plant activity, and the
independent Safety Engineering Group. These functions will be performed
by the proposed Nuclear Assessment Section.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The change does not involve a significant hazards consideration
for the following reasons:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated because it is a programmatic and administrative
change which does not physically alter any safety-related systems,
nor does it affect the way in which any safety-related systems
perform their functions. Since the design of the facility and system
operating parameters are not changing, the proposed amendment does
not involve an increase in the probability or consequences of any
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. As stated in Item 1, the proposed amendment is a
programmatic and administrative change which does not physically
alter any safety-related systems; nor does it affect the way in
which any safety-related systems perform their functions. Since the
design of the facility and system operating parameters are not
changing, the proposed amendment 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 because it is a programmatic and
administrative change which provides assurance that plant operations
continue to be conducted in a safe manner through the performance
based assessment programs. As stated in Item 1, the proposed
amendment does not physically alter any safety-related systems; nor
does it affect the way in which any safety-related systems perform
their functions. Since the design of the facility and system
operating parameters are not changing, the proposed amendment does
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: David B. Matthews
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: August 1, 1994
Description of amendment request: The proposed amendment would
revise the technical specifications to incorporate a 1.0 volt steam
generator tube interim plugging criteria (IPC) for Unit 1 beginning
with Cycle 7, which will begin in the fall of 1994.
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.
Consistent with Regulatory Guide (RG) 1.121, ``Basis for
Plugging Degraded PWR Steam Generator Tubes, '' Revision 0, August
1976, the traditional depth-based criteria for SG tube repair
implicitly ensures that tubes accepted for continued service will
retain adequate structural and leakage integrity during normal
operating, transient, and postulated accident conditions. It is
recognized that defects in tubes permitted to remain in service,
especially cracks, occasionally grow entirely through-wall and
develop small leaks. Limits on allowable primary-to-secondary
leakage established in Technical Specifications ensure timely plant
shutdown before the structural and leakage integrity of the affected
tube is challenged.
The proposed license amendment request to implement voltage
amplitude SG tube support plate Interim Plugging Criteria for Byron
Unit 1 meets the requirements of RG 1.121. The IPC methodology
demonstrates that tube leakage is acceptably low and tube burst is a
highly improbable event during either normal operation or the most
limiting accident condition, a postulated main steam line break
(MSLB) event.
Adequate SG tube leakage integrity during normal operating
conditions is assured by limiting allowable primary-to-secondary
leakage to 150 gpd per SG or 600 gpd total. Currently, this limit is
administratively controlled. However, a license amendment request
was submitted on 06/03/94 to incorporate this limit into the Byron
Technical Specifications. During normal operating conditions, the
tube support plate constrains the [outer diameter stress corrosion
cracking] ODSCC affected area of the tube to provide additional
strength that precludes burst. Any leakage of a tube exhibiting
ODSCC at the [tube support plate] TSP is fully bounded by the
existing SG tube rupture analysis included in the Byron UFSAR.
Therefore, probability of failure of a tube left in service or
consequences of tube failure during normal operating conditions is
not significantly increased by the application of IPC.
During transients, the TSP is conservatively assumed to displace
due to the thermal-hydraulic loads associated with the transient.
This may partially expose a crack which is within the boundary of
the TSP during normal operations to free span conditions. Burst is
therefore conservatively evaluated assuming the crack is fully
exposed to free span conditions. The structural eddy current bobbin
coil voltage limit for free-span burst is 4.54 volts. This limit
takes into consideration a 1.43 safety factor applied to the steam
line break differential pressure that is consistent with RG 1.121
requirements. With additional considerations for growth rate
assumptions and an upper 95% confidence estimate on voltage
variability, the maximum voltage indication that could remain in
service is reduced to 2.7 volts. For added conservatism, the
allowable indication voltage is further reduced in the proposed
amendment to a 1.0 volt confirmed ODSCC indication limit. All
indications between 1.0 and 2.7 volts will be subject to an RPC
examination. Tubes with RPC confirmed ODSCC indications will be
plugged or sleeved. Any ODSCC indications between 1.0 volt and 2.7
volts which are not confirmed as ODSCC will be allowed to remain in
service since these indications are not as likely to affect tube
structural integrity or leakage integrity over the next operating
cycle as the indications that are detectable by both bobbin and
[rotating pancake coil] RPC inspections.
The eddy current inspection process has been enhanced to address
RG 1.83, ``Inservice Inspection of PWR Steam Generator Tubes,''
Revision 1, July 1975, considerations as well as the EPRI SG
Inspection Guidelines. Enhancements in accordance with NUREG-1477
and Appendix A of the Catawba IPC report (WCAP-13698) are in place
to increase detection of ODSCC indications and to ensure reliable,
consistent acquisition and analysis of data. Based on the
conservative selection of the voltage criteria and the increased
ability to identify ODSCC, the probability of tube failure during an
accident is also not significantly increased due to application of
requested IPC.
For consistency with current offsite dose limits, the site
allowable leakage limit during a MSLB has been conservatively
calculated to be 12.8 gpm. This leakage limit includes maximum
allowable operational leakage from the unaffected SGs and the
accident leakage from the affected SG. As a requirement for
operation following application of IPC, the projected distribution
of crack indications over the operating period must be verified to
result in primary to secondary accident leakage less than the site
allowable leakage limit. Thus, the consequences of a MSLB remain
unchanged.
Therefore, as implementation of the 1.0 volt IPC for Byron Unit
1 does not adversely affect steam generator tube integrity and
results in acceptable dose consequences, the proposed license
amendment request does not result in any significant increase in the
probability or consequences of an accident previously evaluated
within the Byron Updated Final Safety Analysis Report.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Implementation of the proposed SG tube IPC does not introduce
any significant changes to the plant design basis. Use of the
criteria does not provide a mechanism which could result in an
accident outside the tube support plate elevations since industry
experience indicates that ODSCC originating within the tube support
plate does not extend significantly beyond the thickness of the
support plate. This criteria only applies to ODSCC contained within
the region of the tube bounded by the tube support plate.
In addressing the combined effects of Loss of Coolant Accident
(LOCA) coincident with a Safe Shutdown Earthquake (SSE) on the SG
(as required by General Design Criteria 2), it has been determined
that tube collapse of select tubes may occur in the SGs at some
plants, including Byron Unit 1. There are two issues associated with
SG tube collapse. First, the collapse of SG tubing reduces the RCS
flow area through the tubes. The reduction in flow area increases
the resistance to flow of steam from the core during a LOCA which,
in turn, may potentially increase Peak Clad Temperature (PCT).
Second, there is a potential that partial through-wall cracks in
tubes could progress to through-wall cracks during tube deformation
or collapse.
A number of tubes have been identified, in the ``wedge''
locations of the SG TSPs, that demonstrate the potential for tube
collapse during a LOCA + SSE event. Because of this potential, these
tubes have been excluded from application of the voltage-based SG
TSP IPC.
Therefore, neither a single or multiple tube rupture event would
be expected in a steam generator in which IPC has been applied.
ComEd has implemented a maximum primary to secondary leakage
limit of 150 gpd through any one SG at Byron to help preclude the
potential for excessive leakage during all plant conditions. The 150
gpd limit provides for leakage detection and plant shutdown in the
event of an unexpected single crack leak associated with the longest
permissible free span crack length. The 150 gpd limit provides
adequate leakage detection and plant shutdown criteria in the event
an unexpected single crack results in leakage that is associated
with the longest permissible free span crack length. Since tube
burst is precluded during normal operation due to the proximity of
the TSP to the tube and the potential exists for the crevice to
become uncovered during MSLB conditions, the leakage from the
maximum permissible crack must preclude tube burst at MSLB
conditions. Thus, the 150 gpd limit provides a conservative limit to
prompt plant shutdown prior to reaching critical crack lengths under
MSLB conditions.
Upon implementation of the 1.0 volt IPC, steam generator tube
integrity continues to be maintained through inservice inspection
and primary-to-secondary leakage monitoring. Therefore, the
possibility of a new or different kind of accident from any
previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The use of the voltage based bobbin coil probe SG TSP IPC for
Byron Unit 1 will maintain steam generator tube integrity
commensurate with the criteria of RG 1.121 as discussed above. Upon
implementation of the criteria, even under the worst case
conditions, the occurrence of ODSCC at the TSP elevations is not
expected to lead to a steam generator tube rupture event during
normal or faulted plant conditions. The distribution of crack
indications at the TSP elevations result in acceptable primary-to-
secondary leakage during all plant conditions and radiological
consequences are not adversely impacted by the application of IPC.
The installation of SG tube plugs and sleeves reduces the RCS
flow margin. As noted previously, implementation of the SG TSP IPC
will decrease the number of tubes which must be repaired by plugging
or sleeving. Thus, implementation of IPC will retain additional flow
margin that would otherwise be reduced due to increased tube
plugging. Therefore, no significant reduction in the margin of
safety will occur as a result of the implementation of this proposed
license amendment request.
Although not relied upon to prove adequacy of the proposed
amendment request, the following analyses demonstrate that
significant conservatisms exist in the methods and justifications
described above:
LIMITED TUBE SUPPORT PLATE DISPLACEMENT
An analysis was performed to verify [the effect] of limited TSP
displacement during accident conditions (MSLB). Application of
minimum TSP displacement assumptions reduce the likelihood of a tube
burst to negligible levels. Consideration of limited TSP
displacement would also reduce potential MSLB leakage when compared
to the leakage calculated assuming free span indications.
PROBABILITY OF DETECTION
The Electric Power Research Institute (EPRI) Performance
Demonstration Program analyzed the performance of approximately 20
eddy current data analysts evaluating data from a unit with 3/4''
inside diameter and 0.049'' wall thickness tubes. The results of
this analysis clearly show that the detectability of larger voltage
indications is increased which lends creditability for application
of a POD of 0.62 for ODSCC indications larger than 1.0
volt.
RISK EVALUATION OF CORE DAMAGE
As part of ComEd's evaluation of the operability of Byron Unit 1
Cycle 7, a risk evaluation was completed. The objective of this
evaluation was to compare core damage frequency under containment
bypass conditions, with and without the interim plugging criteria
applied at Byron Unit 1.
The total Byron core damage frequency is estimated to be 3.09E-5
per reactor year with a total contribution from containment bypass
sequences of 3.72E-8 per reactor year according to the results of
the current individual plant evaluation (IPE). Operation with the
requested IPC resulted in an insignificant increase in core damage
frequency resulting from MSLB with containment bypass conditions.
Therefore, based on the evaluation above, ComEd has concluded
that this proposed license amendment request does not involve a
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Byron Public Library, 109 N.
Franklin, P.O. Box 434, Byron, Illinois 61010
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
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;Docket Nos.
50-295 and 50-304, Zion Nuclear Power Station, Units 1 and 2, Lake
County, Illinois
Date of amendment request: July 8, 1994
Description of amendment request: The proposed amendment would add
a License Condition to specify that commitments made in response to the
March 14, 1983, NUREG-0737 Order shall be maintained pursuant to the
requirements of 10 CFR 50.59.
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:
Involve a significant increase in the probability or
consequences of any accident previously analyzed:
Commonwealth Edison has addressed all issues made in response to
NUREG-0737. As such, the purpose of the post-TMI Order is no longer
served. The inclusion of the modified Order as a license condition
is administrative in nature and does not allow unregulated decreases
in the level of safety; therefore, this license amendment is
appropriate and safe. The proposed license amendment requires
control of NUREG-0737 commitments through 10 CFR 50.59. If an
unreviewed safety question occurs during the review of a NUREG 0737
item then Commonwealth Edison is obligated to submit a change to the
NRC staff as a license amendment. As a result of the proposed
amendment, there are no physical changes to the facility and all
operating procedures, limiting conditions for operation (LCO),
limiting safety system settings, and safety limits specified in the
Technical Specifications will remain unchanged. Therefore, the
proposed license amendment to modify the post-TMI Order will not
increase the probability or the consequences of any accident
previously analyzed.
Create the possibility of a new or different kind of accident
from any previously evaluated:
Since there are no changes in the way the plant is operated, the
potential for a new or different kind of accident is not created.
The proposed changes are administrative in nature and do not affect
any accident initiators for Dresden, Quad Cities, and Zion Stations.
No new failure modes are introduced.
Involve a significant reduction in a margin of safety:
Plant safety margins are established through LCOs, limiting
safety system settings, and safety limits specified in the Technical
Specifications. As a result of the proposed amendment, there will be
no changes to either the physical design of the plant or to any of
these settings and limits. The proposed changes are administrative
and do not affect the safe operation of the sites. Therefore, there
will be no changes to any of the margins of safety.
Guidance has been provided in 51 FR 7744 for the application of
standards to license change requests for determination of the
existence of significant hazards considerations. This document
provides examples of amendments which are not likely considered to
involve significant hazards considerations.
This proposed amendment does not involve a significant
relaxation of the criteria used to establish safety limits, a
significant relaxation of the bases for the limiting safety system
setting or a significant relaxation of the bases for the limiting
conditions for operations. The proposed changes are administrative
in nature without consequence to the safety of the plant. Therefore,
based on the guidance provided in the Federal Register and the
criteria established in 10 CFR 50.92(c), the proposed change does
not constitute a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: for Dresden, Morris Public
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities,
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021; for
Zion, Waukegan Public Library, 128 N. County Street, Waukegan, Illinois
60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Connecticut Yankee Atomic Power Company, and Northeast Nuclear
Energy Company, Docket Nos. 50-213, 50-245, 50-336, and 50-423
Haddam Neck Plant, and Millstone Nuclear Power Station, Units 1, 2,
and 3, Middlesex County, and New London County, Connecticut
Date of amendment request: June 30, 1994
Description of amendment request: The proposed amendments would
modify the Administrative Controls Section of the Technical
Specifications by replacing the present Nuclear Review Board (NRB) for
the Haddam Neck Plant, and the NRB and Site Nuclear Review Board (SNRB)
with a Nuclear Safety Assessment Board (NSAB) which will serve
Millstone Units 1, 2, and 3, and Haddam Neck.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration (SHC), which is presented below:
... These proposed changes do not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The NSAB is an oversight group which provides independent
assessments of activities at the Haddam Neck Plant and Millstone
Unit Nos. 1, 2, and 3. The members of the NSAB are appointed by the
Executive Vice President - Nuclear to provide oversight and feedback
on the operation of the units. The NSAB adds to the defense-in-depth
provided by the design, operation, maintenance, and quality
oversight of the nuclear units by promoting excellence through the
conduct of its affairs and advising the Executive Vice President -
Nuclear in matters concerning nuclear safety.
The proposed modification to the Technical Specifications are
administrative in nature and will establish a new group which will
accomplish the guidance provided in ANSI N18.7-1976. The charter of
the NSAB will be controlled by procedure.
These administrative changes will not increase the probability
of occurrence or the consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed addition of the NSAB and its subcommittees and the
ensuing elimination of the NRB and the SNRB is an administrative
reorganization. There are no changes in the way in which the plants
are physically operated. The administrative changes being
accomplished by the establishment of the NSAB fulfills the function
previously provided by the NRB and the SNRB. The organization of the
NSAB will follow the guidance found in ANSI N18.7-1976 and will be
controlled by procedure.
3. Involve a significant reduction in a margin of safety.
The proposed changes establish the requirements of the NSAB. The
NSAB replaces those activities previously performed by the NRB and
the SNRB. With these changes the new organization will provide more
consistent and clearer feedback to the four units and the Executive
Vice President - Nuclear.
The changes do not directly affect any protective boundaries nor
do they impact the safety limits for the protective boundaries.
These proposed changes are administrative in nature. Therefore,
there can be no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457, for the Haddam Neck Plant, and
the Learning Resource Center, Three Rivers Community-Technical College,
Thames Valley Campus, 574 New London Turnpike, Norwich, Connecticut
06360, for Millstone 1, 2 and 3.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford,
Connecticut, 06141-0270.
NRC Project Director: John F. Stolz
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: August 11, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification Section 6.5.1, Station Nuclear Safety
Committee (SNSC), to change the designation of the Chairman and to
clarify the maximum allowable alternate members for quorum purposes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change does not involve a significant hazards
consideration since:
1. There is no significant increase in the probability or
consequences of an accident.
This is an administrative change. Since this change maintains a
consistent level of chairmanship while continuing to ensure
independence and technical expertise of the SNSC chairman, this
change does not increase the probability or consequences of an
accident.
2. The possibility of a new or different kind of accident from
any previously evaluated has not been created.
This is an administrative change of the designation of the
Chairman of SNSC which does not significantly decrease the level of
senior management which is responsible for chairing SNSC. No new or
different kind of accident has been created.
3. There has been no reduction in the margin of safety.
The independence and technical expertise of the SNSC Chairman
will be preserved. SNSC will continue to be composed of those
individuals most related to matters of nuclear safety. The margin of
safety will not be reduced by this change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Pao Tsin Kuo
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: May 24, 1994
Description of amendment request: The proposed amendments would
transfer the boron concentration in Technical Specification (TS) 3.9.1
for the reactor coolant system and the refueling canal during MODE 6,
and the boron concentration in TS 4.7.13.3 for the spent fuel pool from
the TS to the Core Operating Limits Report (COLR). The application is
submitted in response to the guidance in Generic Letter 88-16 which
addresses the transfer of fuel cycle-specific parameter limits from the
TS to the COLR.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following analysis, performed pursuant to 10 CFR 50.91,
shows that the proposed amendment will not create a significant
hazards consideration as defined by the criteria of 10 CFR 50.92.
1. This amendment will not significantly increase the
probability or consequence of any accident previously evaluated.
No component modification, system realignment, or change in
operating procedure will occur which could affect the probability of
any accident or transient. The relocation of boron concentration
values to the COLR is an administrative change which will have no
effect on the probability or consequences of any previously-analyzed
accident. The required values of boron concentration will continue
to be determined through use of approved methodologies.
2. This amendment will not create the possibility of any new or
different accidents not previously evaluated.
No component modification or system realignment will occur which
could create the possibility of a new event not previously
considered. The administrative change of relocating parameters to
the COLR, in this case boron concentration, cannot create the
probability of an accident.
3. This amendment will not involve a significant reduction in a
margin of safety.
Required boron concentrations will remain appropriate for each
cycle, and will continue to be calculated using approved
methodologies. There is no 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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: February 5, 1993 as supplemented by
letter dated August 1, 1994.
Description of amendment request: The proposed amendment would
revise the Technical Specifications to incorporate a technical review
and control process to supplement the onsite technical review and
approval of new procedures and changes thereto affecting nuclear
safety. This process is discussed in Section 5.5 of the Revised
Standard Technical Specifications, NUREG-1432. This notice supersedes
the notice issued on April 14, 1993 (58 FR 19478), and acknowledges the
clarification in the licensee's August 1, 1994, letter.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change is administrative in nature and provides for
1) procedural reviews through the use of qualified technical review
personnel designated by the PORC [Plant Operating Review Committee]
and 2) procedural approval through the use of group heads designated
by the General Manager Plant Operations as authorized by
administrative controls upon their development. As part of this
process, qualified technical reviewers will be individuals other
than the preparer who will document and implement necessary cross-
discipline reviews prior to approval. The process will be controlled
by administrative controls which will be reviewed by the PORC and
approved by the General Manager Plant Operations.
The procedures governing plant operation will continue to ensure
that plant parameters are maintained within acceptable limits.
Procedures and changes thereto will be reviewed and approved at a
level commensurate with their importance to safety. Therefore, the
proposed changes will not involve a significant increase in the
probability or consequences of any accident previously evaluated.
The proposed changes are administrative in nature. The proposed
changes do not involve physical changes to the plant, changes to
setpoints, or operating parameters. The applicable procedures
governing the operation of the plant will receive reviews and
approvals commensurate with their importance to nuclear safety, and
where appropriate cross-discipline review will be performed.
Therefore, the proposed changes will not create the possibility of a
new or different kind of accident from any previously evaluated.
The proposed changes are administrative in nature. The Waterford
3 safety margins are defined and maintained by the Technical
Specifications in Sections 2-5 which are unaffected. Therefore, the
proposed change will not involve a significant reduction in a margin
of safety.
The licensee's letter dated August 1, 1994, provided a
clarification of the proposed wording of the technical specifications
to assure the personnel performing the technical reviews would have the
necessary technical knowledge base.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 19, 1994
Description of amendment request: The proposed amendment would move
the requirements of Technical Specification 3/4.3.4 Turbine Overspeed
Protection from the technical specifications (TS) and relocate them in
the Updated Final Safety Analysis Report (UFSAR) consistent with the
NRC Final Policy Statement on Technical Specifications Improvements for
Nuclear Power Reactors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change relocates the Turbine Valve Overspeed
Protection requirements from the TS to the Waterford 3 UFSAR
consistent with the NRC Policy Statement on Technical Specification
Improvements. Testing and inspections of the turbine Overspeed
Protection System will remain governed by an approved turbine
maintenance program, described in the UFSAR. This proposed change
has no affect on the current Turbine Overspeed Protection
requirements other then to relocate them to the UFSAR. Thus, the
probability of a turbine missile causing damage to a safety-related
component or structure at Waterford 3 as described in the FSAR
analysis (Reference 5) is not affected. The purpose of the Turbine
Overspeed Protection System is to prevent an overspeed event, the
precursor to a potential turbine fragment missile. Since the purpose
of this system is preventive, it serves no function to mitigate any
accident previously evaluated.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
The proposed change does not involve any change to the
configuration or method of operation of any plant equipment. No new
failure modes or limiting failures have been identified as result of
the proposed change. The proposed change will not alte the operation
of the plant or the manner in which it is operated. Any subsequent
change to the Turbine Oversspeed Protection System requirements will
undergo a review in accordance with the criteria of 10 CFR 50.59 to
ensure that the change does not involve an unreviewed safety
question.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident prveiously
evaluated.
The proposed change will relocate Turbine Overspeed Protection
System requirements from the TS to the Waterford 3 UFSAR on the
basis that the Turbine Overspeed Protection System does not meet the
criteria of the NRC Final Policy Statement on Technical
Specifications Improvements for Nuclear Reactors. The requirements
that will reside in the UFSAR for the Turbine Overspeed Protection
system will ensure that the system remains capable of protecting the
turbine from excessive overspeed. The proposed change will have no
adverse impact on any protective boundary or safety limit.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: July 25, 1994
Description of amendment request: The proposed amendments evise
various Technical Specification sections to implement enhancements
recommended by NRC Generic Letter (GL) 93-05, ``Line-Item Technical
Specification Improvements to Reduce Surveillance Requirements for
Testing During Power Operation,'' for St. Lucie Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The standards used to arrive at a determination that a request
for amendment involves a no significant hazards consideration are
included in the Commission's regulation, 10 CFR 50.92. 10 CFR 50.92
states that no significant hazards considerations are involved if
the 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. Each standard is discussed as
follows:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed amendments conform to the guidance given in
Enclosure 1 of the NRC Generic Letter 93-05. The overall functional
capabilities of the incore detector system, reactor coolant system
pressure isolation valves, safety injection tank, or containment
sump will not be modified by the proposed change. Therefore, the
probability or consequences of an accident are not significantly
increased by the changes.
(2) Use of the modified specification would not create the
possibility of a new or different kind of accident from any
previously evaluated.
The use of the modified specifications can not create the
possibility of a new or different kind of accident from any
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined in the
facility operating license. No new failure mode is introduced due to
the surveillance interval changes and clarifications, since the
proposed changes do not involve the addition or modification of
equipment nor do they alter the design or operation of affected
plant systems.
(3) Use of the modified specification would not involve a
significant reduction in a margin of safety.
The operating limits and functional capabilities of the affected
systems are unchanged by the proposed amendments. Therefore, the
modified specifications which establish new or clarify old
surveillance intervals consistent with the NRC Generic Letter 93-05
line-item improvement guidance do not significantly reduce any of
the margins of safety.
Based on the above, we have determined that the proposed
amendments do not (1) involve a significant increase in the
probability or consequences of an accident previously evaluated, (2)
create the probability of a new or different kind of accident from
any previously evaluated, or (3) involve a significant reduction in
a margin of safety; and therefore do not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Victor McCree, Acting
Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida
Date of amendment request: July 25, 1994
Description of amendment request: The amendment will upgrade
Technical Specification (TS) 3/4.7.1.6 for the Main Feedwater Line
Isolation Valves to be consistent with NUREG-1432, ``Standard Technical
Specifications for Combustion Engineering Plants.'' The changes include
all related requirements of NUREG-1432, Revision O, specification
3.7.3. Accordingly, the proposal is consistent with the Commission's
Final Policy Statement on Technical Specifications Improvements (58 FR
39132).
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:
Pursuant to 10 CFR 50.92, a determination may be made that a
proposed license amendment involves no significant hazards
consideration if 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. Each standard is discussed as
follows:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment will upgrade the existing Limiting
Condition for Operation (LCO) associated with the Main Feedwater
Line Isolation Valves (MFIVs) to be consistent with NUREG-1432,
Standard Technical Specifications for Combustion Engineering Plants.
The MFIVs are not initiators of accidents previously evaluated, but
are included as part of the success paths associated with mitigating
various accidents and transients. The redundancy afforded by two
MFIVs per feedwater line in conjunction with the requirements of the
proposed LCO assure that the feedwater isolation safety function of
these valves can be accomplished considering single failure
criteria. Neither the feedwater system design nor the safety
function of the MFIVs have been altered from those previously
evaluated, and the proposed amendment does not change the applicable
plant safety analyses.
Therefore, operation of the facility in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment will not change the physical plant or the
modes of operation defined in the facility license. The changes are
administrative in nature in that they do not involve the addition of
new equipment or the modification of existing equipment, nor do they
otherwise alter the design of St. Lucie Unit 2 systems. Therefore,
operation of the facility in accordance with the proposed amendment
would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The safety function of the MFIVs is to terminate main feedwater
flow and isolate the safety related portion from the non-safety
related portion of the feedwater system. The proposed amendment, in
conjunction with the redundancy afforded by the feedwater system
design, assures that this safety function can be accomplished
considering single-failure criteria. The bases for required actions
and the action completion times specified for inoperable MFIVs is
consistent with the corresponding specifications in NUREG-1432,
which are equally applicable to St. Lucie Unit 2. The safety
analyses for applicable accidents and transients remain unchanged
from those previously evaluated and reported in the Updated Final
Safety Analysis Report. Therefore, operation of the facility in
accordance with the proposed amendment would not involve a
significant reduction in a margin of safety.
Based on the discussion presented above and on the supporting
Evaluation of Proposed TS Changes, FPL has concluded that this
proposed license amendment involves no significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: Harold F. Reis, Esquire, Newman and
Holtzinger, 1615 L Street, NW., Washington, DC 20036
NRC Project Director: Victor McCree, Acting
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: July 18, 1994
Description of amendment request: The licensee proposes to revise
Technical Specification Table 4.3-1, Reactor Trip System
Instrumentation Surveillance Requirements, Technical Specification
3.3.4, Turbine Governor Valves and Technical Specification 3.7.1.2,
Turbine Driven Auxiliary Feedwater Pump. The purpose of this amendment
is to remove one-time amendments that are no longer necessary. In
addition, six minor editorial changes are proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of a previously evaluated
accident.
The changes proposed to remove the one-time amendments return
the Technical Specifications to the exact wording prior to the one-
time amendments. Returning the Technical Specifications to their
original wording is administrative because the one-time amendments
are no longer applicable. Hence, removing the one-time amendments
would not increase the probability or consequences of an accident.
The other changes are purely editorial in nature, hence, would not
increase the probability or consequences of an accident. Based on
the above, removal of the one-time amendments from the Technical
Specifications will not significantly increase the probability or
consequences of an accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The changes proposed to remove the one-time amendments return
the Technical Specifications to the exact wording prior to the one-
time amendments. Returning the Technical Specifications to their
original wording is administrative because the one-time amendments
are no longer applicable. Therefore, removing the one-time
amendments would not create the possibility of a new or different
kind of accident. The other changes are purely editorial in nature,
hence, would not create the possibility of a new or different kind
of accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The changes proposed to remove the one-time amendments return
the Technical Specifications to the exact wording prior to the one-
time amendments. Returning the Technical Specifications to their
original wording is administrative because the one-time amendments
are no longer applicable. Therefore, removing the one-time
amendments would not involve a significant reduction in a margin to
safety. The other changes are purely editorial in nature, hence,
would not involve a significant reduction in a margin to safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior
College, J. M. Hodges, Learning Center, 911 Boling Highway, Wharton,
Texas 77488
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036
NRC Project Director: William D. Beckner
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: July 12, 1994
Description of amendment request: The proposed amendment changes
the requirement to perform the surveillance test for the channel
functional test Rod Block Monitor, Flow-biased Average Power Range
Monitor and Recirculation Flow instruments from within 24 hours prior
to startup to after the reactor is in the RUN mode, but prior to when
each system is assumed to function in the plant safety analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) The proposed change to the Channel Functional Test frequency
for the RBM will not significantly increase the probability or
consequences for any previously-evaluated event as we are only
matching the mode requirements for performing the SR to the
OPERABILITY requirement for the RBM system, i.e., prior to 30% RTP.
The system will be verified to be OPERABLE prior to when it is
assumed to be OPERABLE in the Updated Final Safety Analysis Report
(UFSAR) for the DAEC.
Allowing the Channel Functional Test for the APRM Flow-Biased
Rod Block Upscale and Downscale trips to be performed ``within 24
hours of entering RUN mode and prior to exceeding 25% RTP'' will not
increase either the probability or consequences of any previously-
analyzed event. The applicable event for the rod block function
during reactor startup is a control Rod Withdrawal Error (RWE),
which is initiated by either an operator error or malfunction within
the Reactor Manual Control System, not by a malfunction within the
APRM system. However, a RWE event that could challenge the fuel
thermal limits is precluded because, as documented in the DAEC UFSAR
(see Section 15.4) and the analysis submitted to support DAEC TS
Amendment No. 120,( NEDC-30813-P,
Average Power Range Monitor, Rod Block Monitor and Technical
Specification Improvement (ARTS) Program for the Duane Arnold Energy
Center, December 1984.), significant margin exists below 25% RTP to
assure the Safety Limit Minimum Critical Power Ratio (SLMCPR) is not
violated by a RWE event. In addition, rod pattern controls are in
place during this period to limit the rod withdrawal sequence, i.e.,
rod worth, such that the fuel thermal limits would not be exceeded.
The Control Rod Drop Accident is unaffected by the requested SR
change as the ``accident'' control rod is assumed to be de-coupled
from its drive mechanism and free-falls from fully inserted to
fully-withdrawal. As the drive for that rod is assumed to be fully-
withdrawn as an initial condition in the event, the APRM rod block
has no role in either preventing or mitigating the rod drop
accident. Thus, revising the SR for the APRM Flow-Biased Rod Block
has no impact upon the Control Rod Drop Accident.
The SRs for the Recirculation Flow Rod Block trips are being
modified for consistency with the APRM Rod Block changes above, as
the sole purpose of this Recirculation Flow signal is to provide the
flow input signal into the APRM Flow-Biased trips. The Recirculation
Flow units are a support system to the APRM Flow-Biased Rod Blocks.
There is no event that is either caused by or mitigated by the
Recirculation Flow Rod Block trips. They are provided solely to
ensure that if the flow signal being input into the APRM circuits is
not valid, a precautionary rod block will be generated as the APRM
Flow-Biased Rod Block setpoint could be in error. Consequently,
allowing the Channel Functional Test for the Recirculation Flow Rod
Block Upscale, Downscale and Comparator trips to be performed
``within 24 hours of entering RUN mode and prior to exceeding 25%
RTP'' will not increase either the probability or consequences of
any previously-analyzed event as these rod blocks are not involved
in either preventing or mitigating any analyzed event.
2) The proposed change to the Channel Functional Test frequency
for the RBM will not introduce any new or different event, as no
changes in system design or operation are being made. We are only
matching the requirement for performing the SR to the OPERABILITY
requirement for the RBM system.
The proposed change to the Channel Functional Test frequency for
the APRM and Recirculation Flow Rod Blocks will not introduce any
new or different event, as no changes in either system design or
operation are being made. In fact, by allowing the Channel
Functional Test to be performed in an operating state which does not
require extensive use of jumpers and/or relay blocks, we reduce the
possibility of an error being made that could cause an inadvertent
actuation of an ESF or disabling of an ESF.
3) The proposed change matches the mode requirement for
performing the SR to the OPERABILITY requirement for the RBM system,
i.e., prior to 30% RTP. The system will be verified to be
OPERABLE prior to when it is assumed to be OPERABLE in the UFSAR
accident analysis. Thus, the margin of safety for the RBM is not
reduced.
As stated in the BASES for TS Chapter 3/4.2, the margin of
safety for the APRM rod block is to prevent violation of the SLMCPR
in RUN by a RWE event. The analysis of the RWE event during Startup
(See DAEC UFSAR Section 15.4.2) and during Power Operation (Ibid),
demonstrates that violations of the SLMCPR are not possible in RUN
below 25% RTP when normal control rod patterns are followed (which
are reinforced by procedural and/or automatic rod pattern controls).
Because the proposed change to the SR for the APRM Flow-Biased Rod
Block will still ensure that the trip will be OPERABLE prior to
exceeding 25% RTP, this change will not reduce the existing margin
of safety.
Again, the Recirculation Flow units are a support system to the
APRM flow-biased circuits. The Recirculation Flow Rod Blocks are
merely precautionary, they do not prevent or mitigate any accident.
Therefore, the proposed revision to the Recirculation Flow Rod block
SR frequency will not reduce the margin of safety for the same
reasons given above for the APRM Rod Blocks.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036.
NRC Project Director: John N. HannonIES Utilities Inc., Docket No.
50-331, Duane Arnold Energy Center, Linn County, Iowa
Date of amendment request: July 29, 1994
Description of amendment request: The proposed amendment would
revise the Technical Specifications by allowing the processing and
implementation of an ISI or IST request for relief from the ASME Code
under 10 CFR 50.59 without prior NRC approval, provided that the relief
request has been reviewed and approved by the plant staff and plant
safety committee.
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 any increase in the probability of
occurrence or consequences of an accident previously evaluated. The
Inservice Inspection and Testing Programs, pursuant to 10 CFR
50.55a, are described in the Technical Specifications. The proposed
amendment, in accordance with NUREG-1433 and draft NUREG-1482,
permits relief from an ASME Code requirement in the interim between
the time of submittal of a relief request and NRC approval of the
relief. The changes being proposed do not affect assumptions
contained in plant safety analyses or change the physical design
and/or operation of the plant, nor do they affect Technical
Specifications that preserve safety analysis assumptions. Any relief
from the approved ASME Section XI Code requirements that is
implemented prior to NRC review and approval will require evaluation
under the 10 CFR 50.59 process to determine that no TS changes or
unreviewed safety questions exist. This evaluation process will
ensure that the impact of any Code relief is thoroughly evaluated
and that the structures, systems and components remain in
conformance with assumptions made in the safety analysis. Therefore,
operation of the facility in accordance with the proposed amendment
would not affect the probability or consequences of an accident
previously evaluated.
2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The
Inservice Inspection and Testing Programs, pursuant to 10 CFR
50.55a, are described in the Technical Specifications. The proposed
amendment, in accordance with NUREG-1433 and draft NUREG-1482,
permits relief from an ASME Code requirement in the interim between
the time of submittal of a relief request and NRC approval of the
relief. The changes being proposed will not change the physical
plant or the modes of operation defined in the Facility License. The
changes do not involve the addition or modification of equipment nor
do they alter the design or operation of plant systems. Any relief
from the approved ASME Section XI Code requirements that is
implemented prior to NRC review and approval will require evaluation
under the 10 CFR 50.59 process to determine that no TS changes or
unreviewed safety questions exist. This evaluation process will
ensure that the impact on any Code relief is thoroughly evaluated
and that the structures, systems and components remain in
conformance with assumptions made in the safety analysis. Therefore,
operation of the facility in accordance with the proposed amendment
would not create the possibility of a new or different kind of
accident previously evaluated.
3) Operation of the facility in accordance with the proposed
amendment would not involve any reduction in a margin of safety. The
Inservice Inspection and Testing Programs, pursuant to 10 CFR
50.55a, are described in the Technical Specifications. The proposed
amendment, in accordance with NUREG-1433 and draft NUREG-1482,
permits relief from an ASME Code requirement in the interim between
the time of submittal of a relief request and NRC approval of the
relief. The changes being proposed do not alter the bases for
assurance that safety-related activities are performed correctly or
the basis for any TS that is related to the establishment of or
maintenance of a safety margin. Any relief from the approved ASME
Section XI Code equirements that is implemented prior to NRC review
and approval will require evaluation under the 10 CFR 50.59 process
to determine that no TS changes or unreviewed safety questions
exist. This evaluation process will ensure that the impact on any
Code relief does not affect the ability of structures, systems or
components to perform their design function, affect compliance with
any TS requirements or reduce the margin of safety. Therefore,
operation of the facility in accordance with the proposed amendment
would not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401
Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea,
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC
20036
NRC Project Director: John N. Hannon
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: July 26, 1994
Description of amendment request: The proposed amendment would
revise the existing limiting condition for operation (LCO) 3.12.A.2.c
to allow for increased flow capacity of the control room emergency
filter system. By increasing the maximum allowed makeup capacity of
this system, additional margin is provided for the positive
pressurization of the control room envelope.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Evaluation
This license amendment request involves the upgrading of the
Control Room Emergency Filter System from 341 cubic feet per minute
(CFM) plus or minus 10% to a maximum of < 1000 CFM. By establishing
a new maximum flowrate for this system, additional filtered makeup
air can be supplied to the Control Room, thus increasing the
positive pressure in the Control Room envelope.
The purpose ofthe Control Room Emergency Filter System is to
remove radioactive iodine and other radioactive materials from the
makeup air during design basis accidents. Therefore, any change to
this system will not increase the probability of an accident
previously evaluated. Radiological calculations show that the
increased flowrate of this system will not result in a significant
increase in Control Room operator dose during a design basis
accidents, and these doses remain well below the established limits.
Therefore, the consequences of an accident previously evaluated are
not significantly increased. The addition of the new Surveillance
Requirement provides a Technical Specification required periodic
demonstration of the positive pressurization function of the system.
This requirement has previously been implemented per existing
surveillance procedures as part of the overall Control Room
Emergency Filter System operability demonstration, and does not
represent a new requirement. This proposed change does not introduce
any new modes of plant operation nor affect any operational
setpoints. The change does not degrade the performance of any safety
system assumed to function in the accident analysis. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility for a new or
different kind of accident from any accident previously evaluated?
Evaluation
This license amendment request involves the upgrading of the
Control Room Emergency Filter System from 341 cubic feet per minute
(CFM) plus or minus 10% to a maximum of < 1000 CFM. This proposed
change involves a physical modification to the Control Room
Emergency Filter System where the filter fan is replaced with a new
fan with greater capacity. To accommodate the additional flow
capacity of the system, an additional charcoal tray is installed in
the charcoal adsorber unit. The District has evaluated the potential
effects of this modification and has determined that the increased
air flowrate is within the system capacity and that radiological
doses, through the filter system, during the design basis accident
are largely unaffected. Because this is a modification of an
existing system with no direct interface with other systems
responsible for prohibiting or mitigating design basis events, the
District has concluded that this proposed change cannot create the
possibility for a new or different kind of accident. This proposed
change does not involve the creation, deletion, or modification of
the function of any structure or system, except as described above,
nor does this change introduce or change any mode of plant
operation. This proposed change does not create the possibility for
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change create a significant reduction in
the margin of safety?
Evaluation
This license amendment request involves the upgrading of the
Control Room Emergency Filter System from 341 cubic feet per minute
(CFM) plus or minus 10% to a maximum of < 1000 CFM. By establishing
this new maximum flowrate for the Emergency Filter System,
additional filtered makeup air can be supplied to the Control Room
envelope, thus improving the margin for positive pressurization with
respect to adjacent areas. Recent tests, utilizing the Main Control
Room Air Conditioning System, have provided information that
supports an increase in the Emergency Filter System flowrate, from
341 CFM to the proposed maximum system capability of approximately
1000 CFM. These tests indicate that an increase of positive pressure
can be achieved by this increased flowrate. This positive pressure
increase provides additional margin of Control Room envelope
positive pressure.
The District has performed radiological calculations to
determine the increase in Control Room operator dose during the 30-
day design basis LOCA event, as a result of increased system air
flow. These calculations show increasing the Control Room Emergency
Filter System to a maximum of 1000 CFM results in a dose of 1.799
Rem whole body and 12.81 Rem thyroid. These doses are not
significantly different than the doses received at a system flowrate
of 341 CFM, which is 1.745 Rem whole body and 11.39 Rem thyroid.
These doses are well within the limits of 10 CFR 20, 10 CFR 50,
Appendix A, General Design Criteria 19, and the guidance provided in
NUREG 0800, which require that doses be limited to less than 5 Rem
whole body, or its equivalent to any part of the body including 30
Rem thyroid, for the duration of any design basis accident. The
above calculated values have also been evaluated and determined to
be within the Updated Safety Analysis Report (USAR) Section XII
requirement of 0.5 Rem in any eight-hour period, whole body from the
reactor building. Increasing the Control Room Emergency Filter
System maximum flowrate has a minimal effect on quantifiable dose
rates, while increasing positive pressurization in the Control Room
envelope. By increasing the positive pressurization in the Control
Room envelope, the possibility of non quantifiable radiation dose to
the Control Room operators, through inleakage, is reduced. This
proposed change does not involve any change to instrument setpoints
or operation. Therefore, the District has concluded that this
proposed change does not create a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305
Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499
NRC Project Director: William D. Beckner
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: July 21, 1994
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 2.2.2 (Reactor Coolant System),
3.2.8/4.2.8 (Pressure Relief Systems - Safety Valves), and the
associated Bases to reduce the number of reactor head safety valves
required operable from 16 valves to 9 valves. The Nine Mile Point
Nuclear Station Unit No. 1 (NMP1) reactor vessel was designed to
American Society of Mechanical Engineers Boiler and Pressure Vessel
Code (ASME Code), Section I-1962 and Code Case 1271N. In order to show
compliance with the ASME Code, it was assumed that a main steam
isolation valve (MSIV) closure occurred without scram. This assumption
demonstrated that it was necessary to have 16 reactor head safety
valves. However, the licensee now states that Section 5.2.2.II.A of
NUREG-0800 (Standard Review Plan) requires that safety valves shall be
designed with sufficient capacity to limit the pressure to less than
110 percent of the reactor coolant system design pressure during the
most severe abnormal operational transient with credit for reactor
scram. The licensee now proposes to use the high neutron flux scram in
analyzing this event. The licensee states that this results in a
reduction in the number of safety valves required operable from 16
valves to 9 valves. The setpoints of the valve groups would remain
unchanged. Testing of the safety valves for setpoint and partial lift
would be changed to be in accordance with the NMP1 Inservice Test
Program which is based upon ASME Code, Section XI, 1983, including
Summer Addenda.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes reduce the number of safety valves from
sixteen (16) to nine (9). The function of the safety valves is to
provide code overpressure protection. This design basis event has
been reanalyzed using the methodology documented in NEDE-24011-P-A,
General Electric Standard Application for Reactor Fuel (GESTAR). The
reanalysis takes credit for the high neutron flux scram in order to
reduce the number of valves. Since peak pressure remains below the
safety limit, the consequences of the event remain the same. The
resultant peak pressure is below the pressure safety limit of 1375
psig. The only event initiator that involves safety valves is the
spurious actuation of one valve. Since the number of valves has been
reduced, the probability of a spurious actuation has been reduced.
Testing in accordance with ASME Section XI will ensure that the
safety valves lift at the required setpoints although the frequency
of testing has been reduced. Therefore, operation of Nine Mile Point
Unit 1 in accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change represents physical changes to the plant as
described in the NMP1 Final Safety Analysis Report (Updated). The
proposed changes however, do not alter the method of providing
overpressure protection, i.e., safety valves. The valves continue to
function to limit peak pressure below the safety limit. Therefore,
maintaining reactor vessel integrity. The reduction in number of
safety valves results from taking credit for the high neutron flux
scram in the safety valve actuation transient as allowed by NUREG-
0800. The initiating event, MSIV closure, remains unchanged.
Although the frequency of testing has been reduced, testing in
accordance with the requirements of Section XI of the ASME Code will
ensure valves lift at the required setpoints. Thus, no potential
initiating events are created which would cause any new or different
kinds of accidents. Therefore, operation of Nine Mile Point Unit 1
in accordance with the proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The overpressure safety limit of 1375 psig remains unchanged. In
addition to the initial conditions associated with the safety valve
actuation transient, NUREG-0800 allows the use of the high neutron
flux scram. This results in the reduction from sixteen (16) to nine
(9) safety valves with peak pressure in the vessel still below the
overpressure safety limit. The margin of safety is defined as the
range between the safety limit (1375 psig) and the failure point of
the vessel. Thus, since peak pressure is below the safety limit, the
margin of safety has not been reduced. Additionally, testing in
accordance with ASME Section XI ensures operation at the required
setpoints and does not result in reduction with margin of safety.
Therefore, the operation of Nine Mile Point Unit 1 in accordance
with the proposed change will not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Pao Tsin Kuo
Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: July 21, 1994
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 2.1.2 (Fuel Cladding Integrity),
3.1.7 (Fuel Rods), 3.4.6/4.6.2 (Protective Instrumentation), and the
associated Bases to allow the use of Range 10 on the Intermediate Range
Neutron Flux Monitors (IRMs) with the Reactor Protection System (RPS)
low pressure trip for main steam line isolation valve closure not in
bypass. Changes are also being proposed to TS Tables 3.6.2.a/4.6.2.a
(Instrumentation that Initiates Scram) and TS Tables 3.6.2.g/4.6.2.g
(Instrumentation that Initiates Control Rod Withdrawal Block) to extend
the calibration frequency of the Source Range Neutron Flux Monitors
(SRMs) and the IRMs from prior to startup and shutdown to once per
operating cycle. In addition, the proposed amendment would change the
Instrument Channel Test interval for the SRMs and IRMs from prior to
startup and shutdown to once per week. The licensee stated that these
changes are in accordance with NUREG-1433 (Improved Standard Technical
Specifications for BWR/4). Associated changes to TS Setpoints, Bases,
References, and Notes for TSs 2.1.2, 3.1.7, and 3.6.2/4.6.2 are also
being proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of Nine Mile Point Unit 1, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes expand the IRM operating range, deletes the
coincident APRM [Average Power Range Monitor] downscale scram trip
and extend the calibration interval for the SRM/IRM System
setpoints. The expansion of the startup operating range is required
to achieve the 1/2 decade overlap between the IRMs and the APRMs.
Proper overlap improves plant safety by ensuring a smooth transition
between IRMs and APRMs. The evaluation of operation in IRM range 10
demonstrates that the addition of range 10 along with the RPS low
pressure isolation activated ensures that the fuel cladding
integrity safety limits would not be exceeded.
The increased IRM/APRM overlap reduces the probability of
multiple APRM channels downscale in the transition between the IRM
and APRM Systems and thus eliminates the need for re-activation of
the IRM scram when in the run mode. The scram is replaced by an
overlap surveillance which requires that the IRMs overlap by at
least 1/2 decade with the APRMs during normal shutdown. This
surveillance ensures that the IRM/APRM overlap is maintained which
is the basis for deletion of the APRM downscale scram. With the
improved overlap, the probability of multiple APRM channels being
downscale is reduced such that it is no longer a credible event and
therefore, the APRM rod block in combination with proper operating
procedures, provides the same level of protection. Thus, normal
plant operation is not affected by these changes and the probability
of previously analyzed accidents is not increased.
The new surveillance intervals and setpoints were calculated
using the General Electric approved methodology documented in NEDC-
31336. The methodology in NEDC-31336 provides assurance that safety
system actuation (i.e., reactor scram or control rod withdrawal
block) will occur prior to the associated system parameter, neutron
flux, from exceeding its analytical limit. Thus, plant response to
previously analyzed accidents remains within previously determined
limits.
Therefore, the operation of Nine Mile Point Unit 1, in
accordance with the proposed amendment, will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The operation of Nine Mile Point Unit 1, in accordance with the
proposedamendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The addition of IRM range 10, deletion of the APRM downscale
scram tripand extension of the surveillance interval for the SRM/IRM
instrumentation, does not involve an initiation or failure not
considered in the Final Safety Analysis Report (Updated). The
proposed changes do not alter the plant configuration and the
initial conditions used for the design basis accident are still
valid. Thus, no potential initiating events are created which would
cause any new or different kinds of accidents. Therefore, operation
of Nine Mile Point Unit 1 in accordance with the proposed amendment,
will not create the possibility of a new or different kind of
accident from any previously analyzed.
The operation of Nine Mile Point Unit 1 in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The addition of IRM range 10 ensures sufficient overlap with the
APRM System such that switching between startup and run can be
easily accomplished. The requirement for having the low reactor
pressure isolation in effect when operating in IRM range 10 is to
prevent a potential depressurization event. Analysis has shown that
the margin between the existing safety limits and those events
previously analyzed has not been reduced. The deletion of the
coincident APRM scram trip has also been shown not to result in a
decrease in the margin of safety as the APRM downscale control rod
withdrawal block provides adequate protection. The analytical limits
associated with the SRM/IRM instrumentation have been reconstituted
in conjunction with extending the surveillance interval to once per
operating cycle. The results using the methodology defined in NEDC-
31336 required that various setpoints in the Technical
Specifications be changed, however, these changes do not reduce the
margins between any existing safety limits and previously analyzed
events. Therefore, operation of Nine Mile Point Unit 1, in
accordance with the proposed amendment, will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Pao Tsin Kuo
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of amendment request: June 23, 1994
Description of amendment request: The proposed change would reword
Technical Specification 3.7, ``Containment Systems,'' to permit
operation with one of the two circuits of the reactor building
ventilation logic temporarily inoperable. In addition, Section
3.7.C.1.b will be reworded to not permit movement of irradiated fuel,
or movement of any loads over irradiated fuel, without secondary
containment integrity.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed changes in accordance with
10CFR50.92 and concluded that the changes do not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed changes do not involve a SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The function of a reactor building ventilation isolation is to
limit fission product release in the event of a design basis
accident (DBA). Two dampers in series are provided in both the
supply and exhaust lines of the reactor building ventilation so that
no single failure would result in the failure to isolate these
secondary containment penetrations. The proposed change permits one
of the two circuits of the reactor building ventilation isolation
logic to be temporarily inoperable.
The Millstone Unit No. 1 DBAs that could be affected by the
proposed LCO [Limiting Condition for Operation] are those in which
the secondary containment is credited to isolate and contain fission
products. The probability of occurrence of any of these accidents is
not altered by changes to the operation of secondary containment.
There is a small potential increase in the consequence of an
accident previously evaluated. This is due to the small increase in
probability that a release could occur while Millstone Unit No. 1 is
operating in the proposed LCO.
The increase in public risk due to the proposed LCO is
negligible. The associated increase in risk is similar to the
increase in public risk permitted by other LCOs. A PRA
[probabilistic risk assessment] has determined that the probability
of an event with a radioactive release in the reactor building,
concurrent with a failure of both operable ventilation dampers to
function, while in the 7-day LCO, is very small.
The wording change to Technical Specification 3.7.C.1.b would
result in more conservative restrictions, in that secondary
containment integrity would be required when any load (e.g., new
fuel) is being moved over irradiated fuel. As currently written, the
Technical Specification only requires secondary containment
integrity when the fuel cask or irradiated fuel is being moved.
Therefore, qualitatively, this change would have a positive impact
on the probability and consequences of accidents involving spent
fuel.
Based upon the above, the proposed changes do not constitute 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 previously analyzed.
The proposed addition of a new LCO which allows operation with
one circuit of the reactor building ventilation isolation logic
inoperable for up to seven days, would only affect the reliability
of the secondary containment isolation. No new equipment is being
added, and no new type of operation is being introduced. This change
allows a short (seven day) period of operation when the reactor
building ventilation is not single failure proof. The reactor
building ventilation functions to mitigate the consequences of
accidents. Failure to function, therefore, does not create the
possibility of a different kind of accident.
The proposed change to Section 3.7.C.1.b increases the
restrictions on load handling, thereby decreasing the possibility of
any kind of accident involving irradiated fuel.
Therefore, the proposed changes to the Technical Specifications
do not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed changes would permit temporary plant operation with
a small decrease in the reliability of secondary containment
isolation. However, the reliability of the reactor building
ventilation isolation would remain high enough with the proposed
LCO, that the impact on the protective boundaries and the margin of
safety would be insignificant.
The proposed change to Section 3.7.C.1.b increases the
restrictions on load handling, decreasing the possibility of any
kind of accident involving irradiated fuel. Therefore this change
would not constitute a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz
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: July 22, 1994
Description of amendment request: The proposed amendment would
revise the technical specifications to (1) change the title of Figure
3.1-5 to be consistent with the applicable Limiting Condition For
Operation (LCO), (2) relocate the Chemical and Volume Control System
(CVCS) valve position requirements to the Reactivity Control Systems -
Shutdown Margin specifications, and (3) consolidate action statements
to be expressed in the LCOs rather than in Surveillance Requirements,
and also clarify the requirements for calculating the heat flux hot
channel factor FQ(z) when using the base load option.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
... The proposed changes would not involve an SHC [significant
hazards consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes are clarifications or relocation of
existing technical specification requirements and do not
substantively affect plant operation. Since they do not affect plant
operations, they cannot be initiators of any events.
The safety analysis of the plant is unaffected by the proposed
changes. Since the safety analysis is unaffected, the calculated
radiological releases associated with the accident analyses are not
affected. Therefore, the proposed changes will not increase the
probability or consequences of previously evaluated accidents.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
As previously stated, the proposed changes are clarifications or
relocation of existing technical specifications and do not
substantively affect plant operation. No new failure modes are
introduced. Since the proposed modifications do not affect plant
operations, they cannot be initiators of new events.
3. Involve a significant reduction in a margin of safety.
The proposed changes are clarifications or relocation of
existing technical specifications and are not substantive changes.
The correction of the title in Figure 3.1-5 will ensure consistency
throughout the technical specifications. The relocation of the CVCS
valves requirements from the RCS [Reactor Coolant System] - Cold
Shutdown Specification to the Reactivity Control Systems - Shutdown
Margin specification will ensure the CVCS valves requirements are
located in the most appropriate location and will help the operators
from the commission of errors or omission of actions due to
inappropriately located material. The final change will revise the
action statement sections of the specification pertaining to heat
flux hot channel factor to ensure all actions in these
specifications are clearly displayed and not contained in the
corresponding surveillance requirements. Therefore, since these
changes are editorial in nature, the proposed modification will have
no impact on the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford,
Connecticut 06141-0270.
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket No. 50-171, Peach Bottom
Atomic Power Station, Unit 1, York County, Pennsylvania
Date of Application for Amendment: May 9, 1994
Brief description of amendment: This Licensee Amendment Request
(LAR) proposes to revise the Peach Bottom Atomic Power Station, Unit 1,
Possession-Only License and Technical Specifications (TS) to reflect
the name change of Philadelphia Electric Company to PECO Energy
Company, to provide proper reference to 10 CFR Part 20 requirements (56
FR 23360), and to reduce the required frequency for performing periodic
inspections in the containment vessel below ground level for water
accumulation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
a. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed changes do not alter the operation of equipment
assumed to be an initiator or any analyzed event or assumed to be
available for the mitigation of accidents or transients. Proposed
changes 1 and 2 are administrative in nature. Proposed change 3 to
reduce the required frequency for performing the periodic inspection
for water accumulation in the containment vessel below ground level
does not impact the probability of ground water intrusion into the
containment building. Proposed change 3 maintains adequate assurance
that integrity of the containment building with respect to ground
water entry will be maintained. The design of Unit 1 makes it very
difficult for ground water to reach the exterior of the containment
liner to start the metal corrosion process. The concrete layer
between the rock and the containment liner serves as a barrier to
prevent water migration to the liner shell. A cathodic protection
system provides protective current to the containment liner as well
as nearby underground piping. The steel containment liner of Unit 1
should not corrode under the present environmental conditions or any
anticipated future conditions even without an operating cathodic
protection system. Monthly inspections from May 1990 (following
issuance of Amendment No. 7 to the Possession-Only License No. DR-12
on April 25, 1990) through April 1994 have not detected any water in
the containment building. Prior to Amendment No. 7, the inspection
of Unit 1 was performed semi-annually. A review of these semi-annual
inspections dating back to October 1981 determined that water has
never been detected in the accessible areas below ground level in
the containment building. The TS limit water accumulation in the
containment sump to 500 gallons. Twelve and one-half years of
inspections have confirmed the reliability of the design of Unit 1
to maintain integrity against any ground water intrusion. There is
no reason, based on the review of inspection data, why the
inspection could not be performed semi-annually rather than monthly.
Therefore, these proposed changes do not increase the probability or
consequences of an accident previously evaluated.
b. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because implementation of the proposed changes do not involve any
physical changes to plant systems, structures, or components. The
proposed changes do not affect the plant SAFSTOR status. Therefore,
the possibility of a new or different kind of accident from any
accident previously evaluated is not created.
c. Does the proposed amendment involve a significant reduction
in a margin of safety?
The proposed changes do not involve a significant reduction in a
margin of safety because the proposed changes do not affect the
plant SAFSTOR status. Because proposed changes 1 and 2 are
administrative in nature, they do not involve a question of safety.
The semi-annual inspection of the accessible areas below ground
level in the containment building for water accumulation, as
proposed by change 3, is adequate to ensure containment building
integrity is maintained with respect to ground water. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
Pennsylvania 19101
NRC Branch Chief: John H. Austin
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 amendment request: August 4, 1994
Description of amendment request: The proposed changes would revise
Sections 3.4 and 3.5 of the Technical Specifications. The Section 3.4
revision would reduce the maximum allowable percent of rated power
associated with inoperable Main Steam Safety Valves (MSSVs). This
change would modify Table 3.4-1 and the associated Basis such that the
maximum power level allowed for operation with inoperable MSSVs is
below the heat removing capability of the operable MSSVs. The Section
3.5 revision would correct administrative errors in the action
statements associated with Items 2.a and 2.c of Table 3.5-4.
Additionally, the proposed changes to Item 2.b of Table 3.5-3 and Item
2.b of Table 3.5-4 would clarify the action statements associated with
inoperable high containment pressure (Hi-Hi Level) instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Consistent with the criteria of 10 CFR 50.92, the enclosed
application is judged to involve no significant hazards based on the
following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. This proposed technical specification change
would modify Table 3.4-1 and the associated basis such that the
maximum power level allowed for operation with inoperable MSSVs is
below the heat removing capability of the operable MSSVs. This
proposed technical specification change will be more conservative
than the current technical specifications. Proposed changes to Items
2.a and 2.c of Table 3.5-4 would restore the original intent of the
specifications and remove undue restrictions on the plant. Proposed
changes to Item 2.b of Table 3.5-3 and Item 2.b of Table 3.5-4
clarify the action statements associated with inoperable high
containment pressure (Hi-Hi Level) instrumentation.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
The proposed license amendment does not create the possibility
of a new or different kind of accident from any accident previously
evaluated. The proposed change incorporates more conservative limits
on the maximum power level allowed for operation with inoperable
MSSVs, restores the original intent of items 2.a and 2.c of Table
3.5-4, and clarifies action statements associated with item 2.b of
Table 3.5-3 and item 2.b of Table 3.5-4.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed amendment would not involve a significant reduction
in a margin of safety. This proposed technical specification change
would modify Table 3.4-1 and the associated basis such that the
maximum power level allowed for operation with inoperable MSSVs is
below the heat removing capability of the operable MSSVs. This
proposed technical specification change will be more conservative
than the current technical specifications. Proposed changes to Items
2.a and 2.c of Table 3.5-4 would restore the original intent of the
specifications and remove undue restrictions on the plant. Proposed
changes to Item 2.b of Table 3.5-3 and Item 2.b of Table 3.5-3 and
Item 2.b of Table 3.5-4 clarify the action statements associated
with inoperable high containment pressure (Hi-Hi Level)
instrumentation.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Pao Tsin Kuo
Power Authority of The State of New York, Docket No. 50-286,
Indian PointNuclear Generating Unit No. 3, Westchester County, New
York
Date of amendment request: August 4, 1994
Description of amendment request: The proposed changes would revise
the fuel oil availability requirements for the Emergency Diesel
Generators (EDGs) from Section 3.7 of the Technical Specifications
(TSs). This TS change would require that 30,026 gallons of fuel oil be
available onsite in addition to the oil in the EDG storage tanks. TS
3.7.F.4 is also being changed to require a total of 7056 gallons of
fuel in the EDG fuel oil storage tanks. In addition, several
administrative changes are being proposed to remove the word
``available'' from the phrase ''... gallons of fuel available...'' in
Section 3.7.A.5 (for the individual storage tanks) to avoid confusion
regarding the amount of usable fuel in the tanks.
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: Consistent with the criteria
of 10 CFR 50.92, the enclosed application is judged to involve no
significant hazards based on the following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously analyzed?
Response:
The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously analyzed.
The change in the minimum required volume for the EDG fuel oil
storage tanks ensures that two EDGs can power minimum safeguards
equipment for 48 hours. The new required levels allow for
temperature effects on fuel density and calibration uncertainties.
The change to the minimum amount of fuel that must be stored onsite
is based on a new fuel consumption profile and ensures that
sufficient oil is present, even in the unlikely event that one EDG
storage tank (and its associated day tank) is unavailable. The
change to specification 3.7.F.4 is consistent with the newly
calculated amount of usable fuel and instrument uncertainties.
The deletion of the word ``available'' from Section 3.7.A.5
(concerning the individual storage tanks) and the change to
Reference 2 of Section 3.7 are administrative in nature and do not
involve a significant increase in the probability or consequences of
a previously analyzed accident.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the changes do not affect current plant configuration or how
the plant operates. The proposed change in the minimum required
volume for the EDG fuel oil storage tanks ensures an adequate amount
of usable fuel and allows for temperature effects on fuel density
and calibration uncertainties. The change to the minimum amount of
fuel that must be stored onsite is based on a new fuel consumption
profile and ensures that sufficient oil is present, even in the
unlikely event that one EDG storage tank (and its associated day
tank) is unavailable. These changes do not alter how the fuel
storage tanks operate and therefore do not create the possibility of
a new or different kind of accident. Specification 3.7.F.4 is being
changed consistent with the revised calculation.
The deletion of the word ``available'' from Section 3.7.A.5
(concerning the individual storage tanks) and the change to
Reference 2 of Section 3.7 are administrative in nature and do not
create the possibility of a new or different kind of accident.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed changes do not involve a significant reduction in a
margin of safety. The proposed change in the minimum required volume
for the EDG fuel oil storage tanks ensures the required amount of
usable fuel is available for two EDGs to operate minimum safeguards
for 48 hours, and it allows for temperature effects on fuel density
and calibration uncertainties. The change to the minimum amount of
fuel that must be stored onsite is based on a new fuel consumption
profile and ensures that sufficient oil is present, even in the
unlikely event that one EDG storage tank (and its associated day
tank) is unavailable. Specification 3.7.F.4 is being changed
consistent with the revised calculation.
The deletion of the word ``available'' from Section 3.7.A.5
(concerning the individual storage tanks) and the change to
Reference 2 of Section 3.7 are administrative in nature and do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Pao Tsin Kuo
Power Authority of the State of New York, Docket No. 50-333,
James A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: August 4, 1994
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) to revise the primary
containment atmosphere monitoring and drywell to torus differential
pressure requirements. Specifically, TS 3.7.A.6 would be revised to
adopt primary containment inerting/deinerting requirements that are
consistent with NUREG-1433, ``Standard Technical Specifications -
General Electric Plants, BWR/4.'' TSs 4.7.A.6.a and 4.7.A.7.a would be
revised to provide frequencies for the verifications of primary
containment oxygen concentration and pressure differential between the
drywell and torus. TSs 3.7.A.7.a.(1) and 3.7.A.7.a.(3) would be revised
to provide requirements for establishing and maintaining differential
pressure between the drywell and torus that are consistent with NUREG-
1433. Several administrative changes to Tables 3.2-8 and 4.3-8 were
also proposed to improve the overall quality of the TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the FitzPatrick plant in accordance with the
proposed Amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes revise primary containment atmosphere
monitoring requirements. The proposed changes adopt reference plant
operating conditions (i.e., 15% rated thermal power) for inerting/
de-inerting requirement as well as for the drywell to torus
differential pressure monitoring consistent with the NRC guidance
provided in the Standard Technical Specifications. The FitzPatrick
Technical Specifications currently allow a 24 [hour] grace period
following startup or before shutdown in which the primary
containment does not have to be inerted. During this 24 hour time
period required leak inspections as well as inerting or shutdown
evolutions are completed. Making the 24 hour ``window'' contingent
upon core thermal power will allow [operators] to place the mode
switch in run sooner, removing startup neutron monitoring
instrumentation scrams (i.e., APRM 15% and IRM upscale/inop). This
reduces the probability of spurious trips due to spiking of this
instrumentation. The proposed changes do not involve physical
modification to the plant nor involve any accident initiators.
Therefore, the probability of an accident occurring remains
unchanged. Accident analyses contained in FSAR [Final Safety
Analysis Report] Chapter 14 assume that a LOCA [Loss-of-coolant
accident] occurs from full power. The consequences of a LOCA below
15% rated thermal power would be less severe and would produce less
hydrogen.
The proposed changes to Tables 3.2-8 and 4.2-8 will eliminate
the reference to Specifications 3.7.A.9 by moving the primary
containment atmosphere monitoring requirements from Specification
3.7.A.9 to Table 3.2-8, Note F. Note F is also revised such that if
recorder 279CR-101A or B is inoperable, a daily monitoring and
logging of the appropriate parameter on the associated indicator on
panel 279CX-101A, B is acceptable in lieu of taking grab samples.
The monitoring will be performed using indicators on 279CX-101A and
B which are Regulatory Guide 1.97 qualified analyzers. The proposed
new Note K is added for completeness. These changes are
administrative in nature and will improve the overall quality of the
technical specifications. Therefore, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes revise primary containment atmosphere
monitoring requirements by adopting STS [Standard Technical
Specifications] guidance regarding inerting/de-inerting
requirements. Consistent with this change the drywell to torus
differential pressure monitoring requirement is being revised.
Adopting the STS reference plant operating condition of 15% rated
thermal power adds operational flexibility. The proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated because the plant safety
analyses assume that a LOCA occurs at full power. In addition,
several changes are proposed to Tables 3.2-8 and 4.2-8 which
simplify hydrogen/oxygen monitoring requirements by moving the
primary containment monitoring requirements from Specification
3.7.A.9 to Table 3.2-8. These changes are administrative in nature
and will result in the overall improvement to the Technical
Specifications. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. involve a significant reduction in a margin of safety.
The proposed changes revise the primary containment atmosphere
inerting/de-inerting requirements and the drywell to torus
differential pressure monitoring requirement. The proposed change
will allow inerting within 24 hours of exceeding 15% rated thermal
power during startup and de-inerting 24 hours prior to reducing
thermal power to less than 15% of rated before a plant shutdown.
These requirements are consistent with the guidance provided in the
STS. This proposed change does not affect the assumptions or
conclusions contained in the plant safety analyses which assume that
a LOCA occurs from full power. The consequences of a LOCA below 15%
rated thermal power would be less severe and would produce less
hydrogen. The proposed changes to Tables 3.2-8 and 4.3-8 are
administrative in nature. Therefore, the proposed changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Pao Tsin Kuo
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: July 27, 1994
Description of amendment request: The amendment request proposes to
revise the Allowed Out-of-service Times (AOTs) for: inoperable Station
Service Water System (SSWS) pumps, inoperable Safety Auxiliaries
Cooling System (SACS) pumps, and inoperable Emergency Diesel Generators
(EDGs). In addition, this request is also proposing to allow online
maintenance of the EDGs.
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:
PSE&G has, pursuant to 10 CFR 50.92, reviewed the proposed
amendment to determine whether our request involves a significant
hazards consideration. We have determined that operation of the Hope
Creek Generating Station in accordance with the proposed changes:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
LCR 94-08
Station Service Water System (SSWS) Changes
Engineering evaluations of the SSWS/Safety Auxiliaries Cooling
System (SACS) demonstrate that adequate heat removal capability is
maintained in the post LOCA/LOP period with either two SSWS/SACS
pumps in one loop or with one SSWS/SACS pump in each independent
loop. The risk evaluations contained in the Probabilistic Safety
Assessment analyses of the SSWS determined that the probability of
an accident previously evaluated does not significantly change by
increasing the SSWS pump AOT from 7 days to 30 days. The evaluations
demonstrated that the relative risk remained low with an increased
(and more appropriate) AOT due to capabilities of the Hope Creek
SSWS to accommodate active failures.
Increasing the SSWS pump AOT does not involve physical
alteration of any plant equipment and does not affect analysis
assumptions regarding functioning of required equipment designed to
mitigate the consequences of accidents. Further, the severity of
postulated accidents and resulting radiological effluent releases
will not be affected by the increased AOT.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Safety Auxiliaries Cooling System Changes
Engineering evaluations of the SSWS/SACS demonstrate that
adequate heat removal capability is maintained in the post LOCA/LOP
period with either two SSWS/SACS pumps in one loop or with one SSWS/
SACS pump in each independent loop. The risk evaluations contained
in the Probabilistic Safety Assessment analysis of the SACS
determined that the probability of an accident previously evaluated
does not significantly change by increasing the SACS pump AOT from
72 hours to 30 days. Similarly, the provision of a 72 hour AOT for
one SACS pump inoperable in each SACS loop does not significantly
change the probability of an accident previously evaluated. The
evaluations demonstrated that the relative risk remained low with an
increased (and more appropriate) AOTs due to capabilities of the
Hope Creek SACS to accommodate active failures.
Increasing the SACS pump AOTs does not involve physical
alteration of the plant equipment and does not affect analysis
assumptions regarding functioning of required equipment designed to
mitigate the consequences of accidents. Further, the severity of
postulated accidents and resulting radiological effluent releases
will not be affected by the increased AOTs.
The proposed changes to ACTION Statement a.2 of Technical
Specification 3.7.1.1 precludes overly conservative and improper
operator action (initiation of plant shutdown procedures) to comply
with the requirements in the situation in which one of the affected
EDGs (an EDG cooled by the inoperable SACS loop) is not realigned to
OPERABLE SACS loop. Currently, Hope Creek can simultaneously be in
the ACTION Statement for Technical Specifications 3.7.1.1 and
3.8.1.1. Simultaneous entry into these ACTION Statements bounds the
conditions of the plant when the proposed requirements of the
Technical Specification 3.7.1.1, ACTION Statement a.2 are met. For
this reason, the proposed changes will not increase the
probabilities or consequences of an accident previously evaluated.
Technical Specification 3.7.1.1, ACTION Statements b., c. and d.
are being revised to require that the RHR loop or safety related
equipment must be declared inoperable when two SACS pumps in the
associated SACS loop are inoperable. This change permits one SACS
pump to be inoperable without affecting the operability of the
associated RHR loop or safety related equipment. Engineering
evaluations demonstrate that two SACS loops with one pump and two
heat exchangers per loop can provide the required heat removal
capability in the post DBA LOCA/LOP scenario and maintain safe
shutdown conditions. Therefore, a SACS loop with one OPERABLE SACS
pump should still be considered as a 100% functional SACS loop,
capable of supplying sufficient cooling for RHR and safety related
equipment required by Specifications 3.4.9.1, 3.4.9.2, 3.5.2,
3.9.11.1 and 3.9.11.2. For this reason, the proposed changes will
not increase the probabilities or consequences of an accident
previously evaluated.
In conclusion, the above SACS changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
LCR 94-11
Emergency Diesel Generator AOT Extensions.
The Hope Creek offsite and onsite power systems are highly
reliable. The risk evaluations contained in the Probabilistic Safety
Assessment analyses of the onsite power system determined that the
probability of an accident previously evaluated does not
significantly change by increasing the diesel generator AOT from 72
hours to 30 days for one inoperable diesel generator or from 2 hours
to 72 hours for two inoperable diesel generators. The evaluations
demonstrated that the relative risk remained low with an increased
(and more appropriate) AOT due to capabilities of the four channel
onsite Class-1E electrical system design at Hope Creek.
Increasing the diesel generator AOT does not involve physical
alteration of any plant equipment and does not affect analysis
assumptions regarding functioning of required equipment designed to
mitigate the consequences of accidents. Further, the severity of
postulated accidents and resulting radiological effluent release
will not be affected by the increased AOT.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
LCR 94-12
Emergency Diesel Generator Online Maintenance
The proposed changes would require that the requisite number of
diesel generators be in an operable condition, but would eliminate
the restriction that the 18 month maintenance inspection and other
surveillance tests be performed only while the unit is shutdown.
Because all operational conditions (governed by the operability of
the equipment prescribed as necessary in Technical Specification
3.8.1.1) and the associated actions are defined elsewhere in the
Technical Specifications, the removal of this restriction would not
involve as significant increase in the probability or consequences
of an accident previously evaluated.
2. Will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
LCR 94-08
Station Service Water System (SSWS) Changes
Extending the SSWS pump AOTS does not necessitate physical
alteration of the plant or changes in parameters governing normal
plant operation. Thus, this change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated for Hope Creek.
Safety Auxiliaries Cooling System Changes
The changes to the SACS do not necessitate physical alteration
of the plant or changes in parameters governing normal plant
operation. Thus, these changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated for Hope Creek.
LCR 94-11
Emergency Diesel Generator AOT Extensions
Extending the diesel generator AOTs does not necessitate
physical alteration of the plant or changes in parameters governing
normal plant operation. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated for Hope Creek.
LCR 94-12
Emergency Diesel Generator Online Maintenance
The proposed revisions will not change the method in which any
of the 4.8.1.1.2.h surveillance activities are to be performed, only
the prescriptive operational condition is being removed. Since the
operational conditions and the associated actions are defined
elsewhere in the Technical Specifications, the removal of this
restriction will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Will not involve a significant reduction in a margin of
safety.
LCR 94-08
Station Service Water System (SSWS) Changes
As discussed above, the Probabilistic Safety Assessment analyses
determined that the change in core damage frequency for extended
SSWS pump AOT is insignificant. Therefore, this change does not
result in a significant reduction in a margin of safety.
Safety Auxiliaries Cooling System Changes
As discussed above, the Probabilistic Safety Assessment analyses
determined that the change in core damage frequency for the SACS
changes are insignificant. Therefore, these changes do not result in
a significant reduction in a margin of safety.
LCR 94-11
Emergency Diesel Generator AOT Extensions
As discussed above, the Probabilistic Safety Assessment analyses
determined that the change in core damage frequency for extended
diesel generator AOTs is insignificant. Therefore, this change does
not result in a significant reduction in a margin of safety.
LCR 94-12
Emergency Diesel Generator Online Maintenance
The margin of safety for the emergency power system depends on
the proven, historical reliability of the diesel generators and the
surveillances verifying the power circuits between the offsite and
the onsite power systems. The elimination of the restrictions for
performance of the maintenance tear down inspection would remain
within the action parameters of Technical Specification 3.8.1.1.
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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Mohan C. Thadani, Acting
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: July 28, 1994
Description of amendment request: The proposed Technical
Specification changes contained herein represent changes to Section 3/
4.8.1 ``AC Sources.'' The revised specification removes the
surveillance requirements, methodology and frequency for Emergency
Diesel Generator (EDG) fuel oil from the Technical Specifications and
relocates them in a controlled plant procedure, VSH.SS-CA.ZZ-0013(Q)
``Procedure for Testing Diesel Fuel and 2 Fuel Oil at
Artificial Island for PSE&G Nuclear Operations.'' The changes also
delete an unnecessary lab test for the fuel oil and extend the
surveillance frequency from once per 92 days to once per 184 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to delete a test not required by Regulatory
Guide 1.137 or ASTM-D975-77 will not result in degradation of fuel
oil quality below acceptable limits. Based on established fuel oil
quality history, the proposed increase in surveillance frequency
from once per 92 days to once per 184 days will not significantly
decrease confidence in fuel oil quality and EDG operability, nor
will the relocation of fuel oil quality surveillance from the
Technical Specifications to the Diesel Fuel Oil Testing Program have
any effect on established plant practices in regards to the testing
of EDG fuel oil. The proposed changes involve no hardware changes,
no changes to the operation of any systems or components, and no
changes to existing structures. Therefore, these changes will not
alter or impact previously evaluated accidents.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed changes are procedural in nature concerning fuel
oil testing and, therefore, will not directly impact the operation
of any plant safety related component or equipment. Any reduction in
fuel oil quality will not be significant or result in a decrease in
EDG operability. Therefore, these changes will not create a new or
unevaluated operating condition.
3. Will not involve a significant reduction in a margin of
safety.
The proposed changes concern how EDG fuel oil quality is to be
determined, how frequently this determination is to be performed,
and how to control the process for determining fuel oil
acceptability, and therefore EDG operability. There are no
associated safety margins and the only margin of concern is that of
fuel oil combustibility due to the presence of either contaminants
or particulate buildup from long term storage. Based on historical
data, PSE&G believes that EDG fuel oil quality will not be affected
or impacted by the proposed changes. Therefore, the proposed
amendment does not involve any reduction in a safety margin.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Mohan C. Thadani, Acting
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: May 20, 1994.
Description of amendment requests: This is a proposal to revise the
Units 2 and 3 Technical Specification (TS) 3/4.7.3, ``Component Cooling
Water System,'' and the corresponding Bases to support the addition of
the component cooling water surge tank backup nitrogen supply (BNS)
system. The amendment is necessary to establish new operability and
surveillance requirements for the system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The Component Cooling Water (CCW) system removes heat to
mitigate the consequences of those design basis accidents included
in chapter 15 of the Updated Final Safety Analysis Report (UFSAR). A
CCW system failure is not an accident initiating event as listed in
the UFSAR, Table 15.0-2. The addition of the Backup Nitrogen Supply
(BNS) system does not change the CCW system function and does not
interface with any system which relates to the initiating events
listed in Table 15.0-2 of the UFSAR. The BNS system is designed to
Quality Class II, Seismic Category I requirements and will increase
CCW reliability by minimizing CCW system voiding during and after a
Design Basis Event (DBE). Failure of the BNS system will not by
itself result in an accident or have any effect on normal plant
operation.
The proposed revision of Technical Specification (TS) 3/4.7.3
will not change the CCW system operation. This amendment request
retains the original CCW TS requirements and adds provisions
specifically limited to the BNS system. The proposed revisions
provide an 8-hour Allowed Outage Time (AOT) for one or both trains
of the BNS system inoperable to avoid unnecessary plant power
reductions. If the 8-hour AOT for BNS system inoperability is not
met, the associated CCW train(s) must be declared inoperable. The 8-
hour AOT followed by either the 72-hour AOT for one train of CCW
inoperable or the 1-hour AOT provided by TS 3.0.3 for both trains of
CCW inoperable results in overall AOTs of 80 and 9 hours,
respectively. The results of a conservative Probabilistic Risk
Assessment demonstrate that for the overall 80-hour and 9-hour AOTs
the increases in core damage risk per year are 6.5E-7 and 8.6E-7,
respectively. This results in less than a 3% increase in the annual
core damage risk for Units 2 and 3.
The proposed revisions to TS 3/4.7.3 include surveillance
requirements to provide assurance that the BNS system remains
OPERABLE when required to support CCW operation. Therefore,
operation of the facility in accordance with this proposed TS change
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed amendment create the possibility of a new or different kind
of accident from any accident previously evaluated?
Response: No
The BNS system does not change the CCW system function and does
not interface with any system related to the initiating events
listed in Table 15.0-2 of the UFSAR. The BNS system is designed to
Seismic Category I requirements and will minimize CCW system voiding
and the potential for a subsequent water hammer by maintaining the
CCW surge tank pressure during and after a DBE. No new High Energy
Line Break considerations apply because the nitrogen bottle pressure
is reduced at the bottle header and all connections are less than
one inch in diameter. The BNS system is independent from all systems
possibly related to the initiating DBEs listed in the UFSAR Table
15.0-2.
The proposed TS 3/4.7.3 revision does not change the existing
CCW system requirements. This proposed change adds operability and
surveillance requirements for the BNS system to support CCW system
operability and provide additional assurance that plant operation is
consistent with the design basis. Failure of the BNS system will not
by itself result in an accident or have any effect on normal plant
operation. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Will operation of the facility in accordance with this
proposed amendment involve a significant reduction in a margin of
safety?
Response: No
The addition of the BNS system enhanced the CCW system by
minimizing the possibility for water hammer following certain
postulated events. Surveillance and nitrogen bottle change-out
procedures assure that the BNS system is available to perform its
safety-related function. The redundant cooling capacity of the CCW
system is maintained by providing an independent dedicated BNS
system for each CCW critical loop, assuming a single failure.
The safety function of the BNS system is limited to the
minimization of void formation in the CCW system under a specific
set of coincident circumstances following a DBE. The proposed
revision to TS 3/4.7.3 allows the BNS system to have one or both
trains inoperable for 8 hours before the associated CCW train(s)
must be declared inoperable. The BNS system AOTs do not affect plant
operation because the BNS system is not normally in operation. The
BNS system action statements are not normally entered for normal
bottle change out since the BNS system is designed with one more
bottle than is required for seven days of BNS system operation.
Therefore, the proposed changes do not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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, P. O. Box 19557, Irvine, California 92713
Attorney for licensee: James A. Beoletto, Esquire, Southern
California Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: Theodore R. Quay
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: May 20, 1994
Description of amendment request: The proposed amendment would
modify the Technical Specifications to incorporate improvements
endorsed by the NRC Final Policy Statement on Technical Specification
Improvements for Nuclear Power Reactors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed Technical Specification changes involve relocating
requirements that are not conditions or limitations on reactor
operation necessary to obviate the possibility of an abnormal
situation or event giving rise to an immediate threat to the public
health and safety. The proposed changes were identified through the
application of criteria designed to cull those requirements that are
not important to operational safety from the Technical
Specifications. In this process, selected provisions of the
Technical Specifications identified for relocation were retained if
necessary to support a Technical Specification that was to be
retained. Thus, only specification requirements that have little or
no operational safety significance are proposed for relocation. In
addition, those requirements that would be relocated will be
included in the Final Safety Analysis Report (FSAR) and, therefore,
will be controlled and implemented as FSAR commitments. In this
manner, those requirements that have no operational safety
significance but involve maintaining the plant in its as-designed
state, (for example, through surveillance programs) would be
controlled.
In addition, the criteria for identifying requirements to be
retained in Technical Specifications specifically call out, for
retention, those structures, systems, or components that are
required to mitigate accidents previously evaluated.
Based on the above, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed changes involve relocating Technical Specification
requirements to another licensee-controlled document, i.e. FSAR
Chapter 16. No changes or physical alterations of the plant are
involved. Also, no changes to the operation of the plant or
equipment are involved. Therefore, the proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed changes involve relocating Technical Specification
requirements to the FSAR. The requirements to be relocated were
identified by applying the criteria endorsed in the Commission's
Policy Statement. Thus, those specifications that would be relocated
do not impose constraints on design and operation of the plant that
are derived from the plant safety analysis report or from
probabilistic safety assessment (PSA) information and do not belong
in the Technical Specifications in accordance with 10 CFR 50.36 and
the purpose of the Technical Specifications stated in the Policy
Statement. Therefore, relocation of these requirements does not
involve a significant reduction in the margin of safety.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
NRC Project Director: John N. Hannon
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: June 21, 1994
Description of amendment request: The proposed amendment would
delete the sections describing the On-Site Review Committee (ORC) and
the Nuclear Safety Review Board (NSRB) from the Technical
Specifications. This change also removes reference to the Manager,
Nuclear Safety and Emergency Preparedness. Additionally, the change
reflects an organizational restructuring which addresses the
Independent Safety Engineering Group (ISEG) reporting to the Manager,
Quality Assurance.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The changes are administrative and equivalent descriptions and
requirements for these oversight committees are contained in FSAR
Section 13.4.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
These changes do not involve any physical alterations to the
plant. There is no new type of accident or malfunction created and
the method and manner of plant operation will not change. The
changes are administrative and equivalent descriptions and
requirements for these oversight committees are contained in FSAR
Section 13.4.
3. Involve a significant reduction in a margin of safety.
The margin of safety remains unaffected since no design change
is made and plant operation remains the same. The changes are
administrative and equivalent descriptions and requirements for
these oversight committees are contained in FSAR Section 13.4.
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: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
NRC Project Director: John N. Hannon
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: July 12, 1994
Description of amendment request: The proposed amendment would
modify the technical specifications (TS) to remove instrument response
time limit tables for the reactor protection system (RPS) and isolation
actuation and emergency core cooling system (ECCS) from the TS. The
affected instrument response time limit tables would be located in the
Final Safety Analysis Report (FSAR).
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 RPS, Isolation Actuation and ECCS Instruments provide
signals to the actuation logic for safety equipment needed to
mitigate accidents and transients. The proposed change relocates the
instrument response times from the Technical Specifications to the
FSAR but will not affect the operability or surveillance
requirements of the affected instruments. The instruments will
continue to be proven operable on the schedule provided in the
Technical Specifications.
The FSAR change process and Plant Operations Committee review
responsibilities ensure that changes to the response time limits
cannot be made without adequate review and approval. Since
operability confirmation as required by the Technical Specifications
(surveillance testing requirements) will not be affected by the
change and the limits themselves cannot be altered without adequate
review and approval, there is no possibility of a significant
increase in the probability of an accident previously approved as a
result of this change.
The instruments provide signals to the actuation logic of
equipment used to mitigate the consequences of an accident. However,
since no changes are being made in the methods or frequencies of
proving operability the systems will not be degraded or be made
susceptible to degradation that could go unidentified. As discussed
above, changes to the limits will not be made without adequate
review and approval. Hence, this change will not affect the
capability of the plant to mitigate a previously evaluated accident.
Because the mitigative capability is not affected there is no
significant increase in the consequences of a previously evaluated
accident as a result of this change.
For the above reasons, the change does not represent 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 relocates only the tables containing the
instrument response times for the RPS, Isolation Actuation and ECCS
response time limits from the Technical Specifications to the FSAR.
The change does not affect how these instruments will function.
Relocation of this information does not represent a change in the
configuration or operation of the plant. No new hardware is being
added to the plant as part of the proposed change. Plant procedures
are not affected by the change. The Technical Specification sections
for the surveillance testing of these instruments will not be
affected. Therefore, the Technical Specifications will continue to
require that the same operability and surveillance requirements be
met for the affected instruments.
Consequently, the possibility of a new or different kind of
accident from any accident previously analyzed is not introduced as
a result of this change.
3. Does the change involve a significant reduction in a margin
of safety?
The margin of safety established by the response time limits is
in ensuring that the RPS, Isolation Actuation and ECCS systems will
respond in time to support the assumptions of the accident analysis.
Relocating the response time limits to the FSAR does not alter the
operability or the surveillance requirements applicable to the
affected instruments. These instruments will continue to be tested
for operability and therefore remain capable of responding to
accident events within the time limits required by the accident
analysis. The administrative change control provisions for the FSAR,
the plant procedures implementing the requirements of 10 CFR 50.59
and the administrative sections of the Technical Specifications are
adequate to control changes to the response time limits such that
they cannot be altered in a manner that would adversely affect plant
safety.
Therefore, for these reasons, the 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: Theodore R. Quay
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: March 23, 1994, as supplemented on July
26, 1994
Description of amendment request: The proposed amendments consist
of two parts: Part one, would revise ``Moderator Temperature
Coefficient (MTC)'' Technical Specifications (TSs) to allow the use of
a slightly positive MTC for the core design. The licensee has stated
that a positive MTC will reduce the burnable rod requirements and
improve operational flexibility. Because of using a positive MTC, the
TSs would be revised to permit a higher boron concentration in the
refueling water storage tank, the reactor coolant system (RCS)
accumulators, and the refueling cavity, in order to ensure adequate
shutdown margin is maintained at all times. Part two, would revise the
TSs to reduce the required RCS flow to offset any reduction in flow due
to increased steam generator tube plugging. Additionally, the
associated Bases for the above TSs would be revised to describe the
basis for the TS requirements.
Because Byron, Unit 1, and Braidwood, Unit 2, will be in refueling
outage in the fall of 1994, the proposed TS changes will apply to them.
Byron, Unit 2 and Braidwood, Unit 1 will continue to operate in
accordance with the current TSs. The licensee's submittal identified
the appropriate unit applicability of the TSs pertaining to the
positive MTC and the required RCS flows.Date of publication of
individual notice in Federal Register: August 15, 1994 (59 FR 41802)
Expiration date of individual notice: September 14, 1994
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, Byron, Illinois 61010; and for Braidwood, the
Wilmington Township Public Library, 201 S. Kankakee Street, Wilmington,
Illinois 60481.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: August 5, 1994
Brief description of amendment request: The proposed amendment
would modify Technical Specification Table 4.8.1.1.2-1, ``Diesel
Generator Test Schedule,'' to exclude selected valid failures of the
Division 1 diesel generator from contributing to an accelerated testing
frequency.Date of publication of individual notice in Federal Register:
August 16, 1994 (59 FR 42080).
Expiration date of individual notice: September 15, 1994
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606
NRC Project Director: John N. Hannon
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: January 14, 1994
Brief description of amendment request: The proposed amendment
would increase the storage capacity in each spent fuel pool from their
current 2040 fuel assemblies to 4117 fuel assemblies. In addition, the
proposed amendment would extend the ``fuel core reserve'' capability
from year 1998 to 2013.
Date of publication of individual notice in Federal Register:
August 8, 1994 (59 FR 40376)
Expiration date of individual notice: September 7, 1994
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC 20555, and at the local public document
rooms for the particular facilities involved.
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona
Date of application for amendments: July 1, 1994, as supplemented
by letter dated August 11, 1994
Brief description of amendment: The proposed amendment changes the
minimum cold-leg temperature for core power levels between 90 percent
and 100 percent to 552 degrees Fahrenheit for Unit 2 (which is a
reduction of 10 degrees Fahrenheit from the previous technical
specification (TS) requirement). This TS change permits reactor
operation at full power with a lower reactor coolant temperature to
minimize potential steam generator tube degradation. The cold-leg
temperature reduction at power levels above 90 percent was previously
granted for Units 1 and 3 by letter dated June 7, 1994.
Date of issuance: August 12, 1994
Effective date: August 12, 1994
Amendment No.: 65
Facility Operating License No. NPF-51: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 13, 1994 (59 FR
35767) The additional information contained in the supplemental letter
was clarifying in nature, was within the scope of the initial notice,
and did not affect the NRC staff's proposed no significant hazards
determination.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 12, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Commonwealth Edison Company, Docket No. STN 50-456, Braidwood
Station, Unit No. 1, Will County, Illinois
Date of application for amendment: June 20, 1994, as supplemented
on August 18, 1994.
Brief description of amendment: The amendment revises Technical
Specification (TS) 4.4.5.4.a(11) by moving a footnote into the body of
the text. Additionally, Item 3 of TS Section 4.4.5.4.a(11) has been
revised to remove the licensee's previous calculation of primary-to-
secondary leakage of 26 gallons per minute (gpm) at the end of 100
calendar days of operation in the present fuel cycle. In place of this
value, the licensee's revised calculated value of less than 9.1 gpm at
the end of Cycle 5 is inserted, including a reference to the basis for
this revised estimate (i.e; WCAP-14046). Finally, Section 3.4.8.a is
revised to remove the limit on operating time in the present fuel
cycle. The maximum permissible dose equivalent Iodine-131 concentration
in the footnote to Section 3.4.8.a remains at 0.35 microcuries per gram
of coolant as proposed by the licensee in its letter dated August 18,
1994. The net result of these revisions is to remove the limitation on
permissible operating time from the Braidwood 1 TSs.
Date of issuance: August 18, 1994
Effective date: August 18, 1994
Amendment No.: 54
Facility Operating License No. NPF-72. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 11, 1994 (59 FR
35389) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 18, 1994. The staff has
found that its prior determination of no significant hazards
consideration is not affected by the licensee's submittal of August 18,
1994.No significant hazards consideration comments received: No
Local Public Document Room location: Wilmington Township Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of application for amendment: June 28, 1989, as supplemented
May 1 and September 26, 1991, March 18, August 24, and August 28, 1992,
May 19, 1993, May 5 and July 7, 1994.
Brief description of amendment: This amendment adds new operational
requirements, action statements, and surveillance requirements to
assure the availability of shutdown cooling to the primary coolant
system during certain operational conditions.Date of issuance: August
12, 1994 Effective date: August 12, 1994, with full implementation
within 90 days
Amendment No.: 161
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 7, 1990 (55 FR
8221) and August 18, 1993 (58 FR 43924). The May 5 and July 7, 1994
letters provided clarifying information within the scope of the August
18, 1993, notice and did not affect the staff's proposed no significant
hazards consideration findings. The Commission's related evaluation of
the amendment is contained in a Safety Evaluation dated August 12,
1994.No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: October 22, 1993
Brief description of amendments: These amendments revise the
Appendix A TSs relating to surveillance test intervals and allowed
outage time for the analog instrumentation channels of the reactor trip
system and the engineered safety feature actuation system.
Date of issuance: August 8, 1994
Effective date: August 8, 1994
Amendment Nos.: 181 and 61
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34660) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 8, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear
One,Unit No. 1, Pope County, Arkansas
Date of amendment request: March 3, 1994
Brief description of amendment: The amendment removed restrictions
from the Arkansas Nuclear One, Unit No. 1 technical specifications that
prohibit use of the auxiliary building crane to move spent fuel
shipping casks.
Date of issuance: August 4, 1994
Effective date: August 4, 1994
Amendment No.: 173
Facility Operating License No. DPR-51. Amendment revised the
Technical Specifications
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17598) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 4, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russelville, Arkansas 72801
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket No.
50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County,
Georgia
Date of Application for Amendment: July 19, 1994, as Supplemented
August 4, 1994.
Brief description of amendment: The amendment revised Technical
Specification 3.3.6.6, ``Traversing Incore Probe System,'' for Hatch
Unit 2 to permit the traversing incore probe (TIP) system to be
considered operable with less than four operable TIP units. Date of
issuance: August 8, 1994
Effective date: August 8, 1994
Amendment No.: 134 (Unit 2)
Facility Operating License No. NPF-5: The amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes. (59 FR 37516 dated July 22,
1994). The notice provided an opportunity to submit comments on the
Commission's proposed no significant hazards consideration
determination. No comments have been received. The notice also provided
for an opportunity to request a hearing by August 22, 1994, but
indicated that if the Commission makes a final no significant hazards
consideration determination, any such hearing would take place after
issuance of the amendment. The August 4, 1994, letter provided
additional information that did not change the scope of the July 19,
1994, application and initial proposed no significant hazards
consideration determinations.The Commission's related evaluation of the
amendment, finding of unusual circumstances, and a final no significant
hazards consideration determination are contained in a Safety
Evaluation dated August 8, 1994.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: May 6, 1994
Brief description of amendment: The amendment changes the monthly
operational test of the reactor trip bypass breakers from monthly to
monthly staggered, such that each breaker is tested every 62 days.
Also, it changes the word Breakers in the Functional Unit title to
Breaker.
Date of issuance: August 12, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 93
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32233) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 12, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Community-Technical College, Thames Valley Campus, 574 New London
Turnpike, Norwich, Connecticut 06360.
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of application for amendment: June 24, 1994
Brief description of amendment: The amendment revises the Technical
Specifications to change the Administrative Controls section to require
an individual who serves as the Operations Manager to either hold a
Millstone Unit 2 Senior Reactor Operator (SRO) license or have an SRO
license at another pressurized water reactor. If the Operations Manager
does not hold a Millstone Unit 2 SRO license, then an individual
serving as the Assistant Operations Manager would be required to
possess an SRO license at Millstone Unit 2.
Date of issuance: August 11, 1994
Effective date: As of the date of issuance.
Amendment No.: 178
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: Yes (59 FR 34872, July 7, 1994).
That notice provided an opportunity to submit comments on the
Commission's proposed no significant hazards consideration
determination. No comments have been received. The notice also provided
for an opportunity to request a hearing by August 8, 1994, but
indicated that if the Commission makes a final no significant hazards
consideration determination any such hearing would take place after
issuance of the amendment. The Commission's related evaluation of the
amendment, finding of exigent circumstances, and final determination of
no significant hazards consideration are contained in a Safety
Evaluation dated August 11, 1994.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of application for amendments: September 21, 1992, as revised
December 29, 1992, November 24, 1993, May 17, 1994, and June 21, 1994.
Brief description of amendments: The amendments revise Technical
Specifications and associated Bases for surveillance test intervals and
allowed outage times for the engineered safety features and reactor
protection system instrumentation consistent with the NRC staff
position as documented in NRC letters to the Westinghouse Owners Group.
The amendments also update operation modes to be consistent with
Westinghouse Standard Technical Specification operational modes and
also include several editorial changes to the Prairie Island Technical
Specifications that are unrelated to the changes described above.
Date of issuance: August 10, 1994
Effective date: August 10, 1994
Amendment Nos.: 111 & 104
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10012). The May 17, 1994, and June 21, 1994, letters provided
clarifying information within the scope of the March 2, 1994, notice.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated August 10, 1994.No significant hazards
consideration comments received: No.
Local Public Document Room location: Minneapolis Public
Library,Technology and Science Department, 300 Nicollet Mall,
Minneapolis, Minnesota 55401.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: December 28, 1993
Brief description of amendment: The amendment to the technical
specifications revised the surveillance test frequency from monthly to
quarterly for several channel functional tests for reactor protective
system and engineered safety feature instrumentation and controls based
on Generic Letter 93-05.
Date of issuance: August 17, 1994
Effective date: August 17, 1994
Amendment No.: 163
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 2, 1994 (59 FR
10013) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 17, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: March 24, 1994
Brief description of amendments: The amendments revise Technical
Specification Sections 3.11.1.4, 6.9.1.8, and 6.14.1, and TS Definition
1.24 to change the frequency for submitting the Semiannual Radioactive
Effluent Release Report to the NRC from semiannually to annually.
Date of issuance: August 10, 1994
Effective date: August 10, 1994
Amendment Nos. 73 and 35
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 12, 1994 (59 FR
24751) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 10, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Public Service Electric and Gas
Company, Delmarva Power and Light Company, and Atlantic City
Electric Company, Docket No. 50-277, Peach Bottom Atomic Power
Station, Unit No. 2, York County, Pennsylvania
Date of application for amendment: April 1, 1993, as supplemented
by letters dated April 7, July 16, and August 20, 1993, and June 8,
1994
Brief description of amendment: These amendments implement an
expanded power-to-flow operating domain supported by the Average Power
Range Monitor, Rod Block Monitor, Technical Specifications Improvement/
Maximum Extended Load Line Limit Analysis (ARTS/MELLLA) (NEDC-32162P,
Revision 1, February 1993) submitted with the licensee's April 1, 1993,
application.
Date of issuance: August 10, 1994
Effective date: Following startup from Refueling Outage 2R10.
Amendment No.: 192
Facility Operating License No. DPR-44: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 21, 1993 (59 FR
39058) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 10, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Philadelphia Electric Company, Public Service Electric and Gas
Company Delmarva Power and Light Company, and Atlantic City
Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: April 27, 1994
Brief description of amendments: The amendments modify the existing
Limiting Conditions for Operation, surveillance requirements and bases
to reflect new containment monitoring system hydrogen/oxygen analyzers.
The new analyzers are to be installed in Unit 2 during the scheduled
September 1994 refueling outage and will support the Containment
Atmospheric Dilution system and the Containment Atmospheric Control
system.
Date of issuance: August 10, 1994
Effective date: Prior to the startup of Unit 2 following refueling
outage 2R10.
Amendments Nos.: 193 and 197
Facility Operating License Nos. DPR-44 and DPR-56: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29629) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 10, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: June 25, 1993
Brief description of amendment: The amendment revised the Plant
Operations Review Committee (PORC) composition and quorum description,
presented the membership composition through a set of requirements
defining the necessary management titles to functional titles, changed
the term ``designated alternate'' to ``designee,'' and removed the
requirements in Specification 6.5.2.5 to have the Nuclear Safety Review
Committee meetings ``at least once per calendar quarter during the
initial year of operation following fuel loadings and...thereafter.''
Date of issuance: August 17, 1994
Effective date: Date of issuance, to be implemented within 90 days
Amendment No. 65
Facility Operating License No. NPF-58. This amendment revised the
Facility Operating License.
Date of initial notice in Federal Register: September 15, 1993 (58
FR 48390) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 17, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa
County, Virginia
Date of application for amendments: November 19, 1993
Brief description of amendments: The amendments revise the NA-1&2
TS to allow the substitution of solid stainless steel or zirconium
alloy filler rods for a limited number of failed fuel rods in fuel
assemblies. This will allow the use of reconstituted fuel assemblies,
which were scheduled for reload, without requiring reload core design
and selection of a replacement assembly during a refueling outage.
Date of issuance: August 9, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 186 and 167
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67863) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated August 9, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: February 26, 1993, as
supplemented on March 9, 1993.
Brief description of amendments: The amendments revised Technical
Specification Section 15.3.1.A.3, ``Limiting Conditions for Operation,
Components Required for Redundant Decay Heat Removal Capability.'' The
amendments clarified the exception for when one decay heat removal
method must be in operation. In addition, the amendments changed the
applicable Basis (page 15.3.1-3c) to improve the clarity and
consistency of this section. Date of issuance: August 16, 1994
Effective date: Immediately, to be implement within 20 days
Amendment Nos.: 149 and 153
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43939) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 16, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: May 24, 1994
Brief description of amendment: The amendment relocates the TS
requirements related to seismic monitoring instrumentation from the TS
to the Updated Safety Analysis Report (USAR). The requirements of these
TS will be maintained and controlled pursuant to Appendix A to 10 CFR
100 and other applicable regulations, including 10 CFR 50.59,
``Changes, tests, and experiments.''
Date of issuance: August 11, 1994
Effective date: August 11, 1994, to be implemented within 120 days
of issuance.
Amendment No.: 75
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 6, 1994 (59 FR
34671) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 11, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: October 27, 1993
Brief description of amendment: The amendment revises Technical
Specification 4.6.1.2.a, Overall Integrated Containment Leakage Rate,
to provide one-time relief from the requirements to perform the
surveillance at intervals of 40 months plus or minus 10 months. The
schedule for the third Type A test is extended to the eighth refueling
outage, approximately 54 months after the second test, in order to have
it coincide with the 10-year inservice inspections.
Date of issuance: August 12, 1994
Effective date: August 12, 1994, to be implemented within 30 days
days of issuance.
Amendment No.: 76
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64616) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 12, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: February 24, 1994
Brief description of amendment: The amendment revises Technical
Specification 4.7.1.2.1.a to require that the turbine-driven and motor-
driven auxiliary feedwater pumps be tested at least quarterly on a
staggered basis instead of the previously required testing once per 31
days on a staggered basis. The revised surveillance frequency is
consistent with the guidance issued in Generic Letter 93-05, ``Line-
Item Technical Specification Improvements to Reduce Surveillance
Requirements for Testing During Power Operation.'' The Bases to TS 3/
4.7.7, Emergency Exhaust System - Auxiliary Building, and TS 3/4.9.13,
Emergency Exhaust System - Fuel Building, are also revised to eliminate
the reference to the use of automatic control for the emergency exhaust
system heaters.
Date of issuance: August 16, 1994
Effective date: August 16, 1994, to be implemented within 30 days
of issuance.
Amendment No.: 77
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17610) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 16, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: April 19, 1994
Brief description of amendment: The amendment revises Technical
Specification Table 3.6-1, ``Containment Isolation Valves,'' by
deleting reference to two (2) valves. The technical specification
change reflects a planned modification which removes the essential
service water (ESW) containment air cooler return line isolation valve
bypass valves and associated piping.
Date of issuance: August 16, 1994
Effective date: August 16, 1994
Amendment No.: 78
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32239) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated August 16, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Dated at Rockville, Maryland, this 24th day of August 1994.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects I/II, Office of Nuclear Reactor
Regulation
[Doc. 94-21325 Filed 8-30-94; 8:45 am]
BILLING CODE 7590-01-F