# Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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## Record

- **Collection:** Federal Register
- **Document type:** Uncategorized Document
- **Published:** November 9, 1994

## Text

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 October 17, 1994, through October 28, 1994.
The last biweekly notice was published on October 26, 1994 (59 FR
53834).

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 December 9, 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.

Carolina Power & Light Company, et al.

Docket Nos. 50-325 and 50-324

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County,
North Carolina.
Date of amendments request: September 30, 1994.
Description of amendments request: The amendments would revise the
Technical Specifications to eliminate the scram and isolation trip
functions from the main steam line radiation monitor (MSLRM). This
change would specifically remove the reactor scram, main steam line
isolation valve closure, main steam line drain valve closure, reactor
water sample line isolation, and mechanical vacuum pump line isolation
actuated on a MSLRM High-High Radiation signal. The actuation signal
for isolation of the reactor water sample line will be replaced with a
low condenser vacuum signal. The isolation of the mechanical vacuum
pump line will be changed to a signal from the main stack radiation
monitor.
The MSLRMs will have both High Radiation and High-High Radiation
alarms. The setpoint for the MSLRM High Radiation alarm will be set at
or below 1.5 times the nominal full power background radiation adjusted
for Hydrogen water chemistry operation. The setpoint for the condenser
off-gas radiation monitor will be set at a value of 1.5 times
background radiation, but not less than 1.5 Rem per hour.
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 amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The deletion of the MSLRM trip function from the reactor scram and the
Group 1 isolation initiation logic removes a potential transient
initiation and therefore decreases the probability of plant transients
occurring due to inadvertent scrams resulting from this system.
The deletion of the MSLRM trip function from the Main Steam Drain
Valve, the Reactor Water Sample Isolation Valve, and the Mechanical
Vacuum Pump line isolation logic, does not affect the initiators of any
accident previously evaluated in the Safety Analysis Report. Therefore,
the proposed change does not involve an increase in the probability of
occurrence of any accident previously evaluated.
The NRC staff acceptance criterion for the Control Rod Drop
Accident is that the doses from the accident fall significantly below
the limits given in 10 CFR Part 100. The releases calculated for
accident during plant operations when the Steam Jet Air Ejectors (SJAE)
are operating and when the Mechanical Vacuum Pumps are operating are
within these acceptance limits.
In NEDO-31400, GE shows that the occurrence of a CRDA, with the MSL
high radiation isolation removed, and SJAE in operation, results in
offsite radiological exposures that are small fractions of 10CFR100
guidelines. Since the Brunswick specific CRDA doses are lower than the
[sic] calculated by GE and the GE dose parameters envelope those used
for the Brunswick analysis, it is concluded that the NRC's findings
that the radiological release consequence is within the staff's
acceptance criteria, even without the automatic MSIV trip, is
applicable to Brunswick.
While not specifically addressed in the GE evaluation, Carolina
Power and Light also proposes to eliminate the Main Steam Line Drain
valves, the Reactor Water Sample Line isolation valves, and the
mechanical vacuum line isolation valves from the MSLRM isolation logic.
Main Steam Line Drain Valves B21-F016 and B21-F019 drain to the main
condenser, which is the same flow path as the MSIVs. The discharge of
both the MSIV and MSL drain flow paths is processed through the offgas
system. Any radiation released through the drain valves during a
control rod drop accident will be negligible and, for Brunswick, is
bounded by the NEDO analysis.
The reactor water sample line provides a small amount of reactor
water to the Reactor Building Sample Panel. The discharge of the
Reactor Building Sample Panel is routed through the floor drain sump to
the liquid radwaste system. Any releases through this path would be
negligible and, for Brunswick, is bounded by the NEDO analysis.
The mechanical vacuum pumps are used only when the reactor is at
low power (less than 5%) and there is insufficient steam flow to
operate the Steam Jet Air Ejectors. The increase in radiation will be
detected by the MSLRMs and annunciated in the Main Control Room.
Operators will be instructed, in the annunciator response procedures,
to take action to stop the Mechanical Vacuum Pump(s) and isolate the
Mechanical Vacuum Pump line. The amount of radiation released prior to
isolating the line would represent the most limiting case for this
accident. However, it will still be well within 10 CFR Part 100 limits.
Additionally, the dose received in the Main Control Room as a result of
this accident is within General Design Criteria 19 (SRP 6.4) limits.
Therefore, since elimination of the MSIV [sic, MSLRM] scram and
isolation functions would not result in an increase in exposure above
NRC acceptance limits, the proposed changes will not significantly
increase the consequences of a previously evaluated accident.
2. The proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The function of a MSLRM trip is to detect abnormal fission
product release and isolate the steam lines, thereby stopping the
transport of fission products from the reactor to the main condenser.
The monitors do not perform a prevention function for any kind of
accident. The existence of a MSLRM trip does not prevent the occurrence
of a fuel failure event or any other type of event. The elimination of
these signals, which served only in a mitigative function, does not
create the possibility of a new or different kind of accident from
those previously evaluated. Also, radiation monitors with alarm
functions will remain installed in the plant to warn the operators of a
high radiation condition in the main steam lines, or in the off-gas
system. Thus no new or different accident can be postulated by the
proposed changes.
3. The proposed amendments do not involve a significant reduction
in a margin of safety. As shown in the topical report, the changes
represent an overall improvement in plant safety. Safe operation of the
plant is further enhanced by elimination of the unnecessary scram and
isolation of the reactor vessel. With implementation of these changes,
1) the primary heat sink (main condenser) remains available, 2) large
transients on the reactor vessel, as well as challenges to the ESF, are
avoided, and 3) the Offgas system remains available to control the
pathway of potential releases. As such, the margin of safety is
enhanced by the proposed changes.
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 Acting Project Director: Michael L. Boyle.

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: October 7, 1994.
Description of amendment request: The proposed amendment would
revise the introduction to TS Section 6.9.3.3 to require the approved
revision number for the referenced analytical methods be listed in the
Core Operating Limits Report. The methodology referenced in 6.9.3.3.b.f
(XN-NF-82-49(A)) will be updated to clarify that all supplements are
included. New methodologies ANF-89-151(A) and EMF-92-081(A) will be
added to TS Section 6.9.3.3.b.
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 changes 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. The plant design basis will not
be altered. Therefore, there will be no significant increase in the
probability of an accident previously evaluated.
Consequences are dependent on the type of accident and the
mitigating response of safety related equipment. Furthermore, the
magnitude of consequences are calculated, directly or through
supporting calculations, by use of NRC approved methodologies. The
proposed license amendment 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 any
current limitations or restrictions. Also, this amendment will not
alter the requirement that evaluation of the consequences of an
accident previously evaluated by determined/supported with NRC reviewed
and approved methodologies. The change to TS Section 6.9.3.3.b's
introductory wording satisfies an administrative commitment and the
requirements it adds are administrative in nature. Accordingly the
proposed license amendment will not involve a significant 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 addition of and update to NRC previously reviewed and approved
methodologies in TS Section 6.9.3.3.b will not result in any design 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/or restrictions. The change
to TS Section 6.9.3.3.b's introductory wording satisfies an
administrative commitment and the requirements it adds are
administrative in nature. Therefore, the proposed amendment will not
create the possibility of a new kind of accident from any accident
previously evaluated.
The addition of and update to NRC previously reviewed and approved
methodologies in TS Section 6.9.3.3.b will not result in any design 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/or restrictions. The changes
to TS Section 6.9.3.3.b's introductory wording satisfies an
administrative commitment and the requirements it adds are
administrative in nature. Therefore, the proposed amendment will not
create the possibility of a 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 license amendment is defined as
administrative in nature. No current operational limits, restrictions,
or operating modes of the facility and its equipment, safety related or
otherwise, designed to preserve the margin of safety will be changed or
affected by the proposed amendment. There will be no changes to
setpoints or to the plant design basis. The methodology proposed for
addition to TS Section 6.9.3.3.b and the methodology that will be
updated has been previously reviewed and approved by the NRC. The
change to TS Section 6.9.3.3.b's introductory wording satisfies an
administrative commitment and the requirements it adds are
administrative in nature. 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: William H. Bateman.

Entergy Operations, Inc., et al.

Docket No. 50-416

Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi.
Date of amendment request: October 12, 1994.
Description of amendment request: The proposed amendment requests
the closure and deletion of License Condition 2.C.(26) related to
turbine disk 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:
1. No significant increase in the probability or consequences of an
accident previously evaluated results from this change.
The proposed change would close and delete License Condition
2.C.(26). The approved methodology currently used to evaluate the
probability of rotor failure and the inspection interval will not be
changed. The closure and deletion of the license condition is an
administrative change and will affect any accident previously
evaluated.
The bounding accident for the turbine-generator as analyzed in the
Grand Gulf Nuclear Station (GGNS) Updated Final Safety Analysis Report
(UFSAR) is the occurrence of an external missile resulting from the
failure of a low pressure (LP) turbine disc. The probability of this
incident occurring is less than 1 x 10-5 per year, which is the
NRC acceptable failure criterion for probability.
Any extension to the service interval in the future will be
evaluated in accordance with the current methodology. The original
acceptable levels of failure will be maintained. Therefore, no
significant increase in the probability or consequences of a previously
evaluated accident results from this change.
2. The change would not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed change does not involve a change to the control logic
or operating procedures for the turbine but rather transfers the
control of the LP turbine disc inspection interval from the Operating
License to administrative control. The current approved methodology
will continue to be used when determining future inspection intervals.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The change would not involve a significant reduction in a margin
of safety.
Closing and deleting the current license condition for LP turbine
disc inspections and controlling the inspection interval
administratively has no adverse effects to the margin of safety. The
current approved methodology for failures will continue to be used and
any changes to future inspection intervals will be evaluated by the
methodology. This change does not affect any previous safety analysis
presented in the UFSAR and does not affect the criteria used to
establish safety limits, the basis for limiting safety system settings,
the basis for limiting conditions of operation, a change to the
technical specifications or a change in plant operations.
Therefore, this change does not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502.
NRC Project Director: William D. Beckner.

Florida Power and Light Company

Docket Nos. 50-250 and 50-251

Turkey Point Plant, Units 3 and 4, Dade County, Florida.
Date of amendment request: October 20, 1994.
Description of amendment request: The licensee proposes to change
Turkey Point, Units 3 and 4 Technical Specifications (TS) by revising
TS 1.9, Definitions--CORE ALTERATIONS to only address activities which
may, in actuality, affect core reactivity. In addition, the licensee
proposes to revise TS 3.9.4, Containment Building Penetrations to allow
both containment personnel airlock (PAL) doors to be open during core
alterations and movement of irradiated fuel in containment provided (a)
that at least one PAL door is capable of being closed; (b) the plant is
in Mode 6 with at least 23 feet of water above the fuel; and (c) a
designated individual is available outside the PAL to close the door.
The licensee also proposes a revision to the footnote of TS 3.9.4, to
remove the description of the purpose for imposing administrative
controls.
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
amendments would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The change in the definition of CORE ALTERATIONS would allow the
movement of a temporary source range detector or other small
components, such as cameras, tools, etc., within the reactor vessel
without the activity being considered CORE ALTERATIONS. The potential
exists, however small, that an object can be dropped into the reactor
vessel. However, the justification for this change, is that the
insertion of small components into the reactor vessel will have no
effect on core reactivity since these items displace a small volume of
borated water, and sufficient borated water will surround the
components and provide the necessary neutron absorption to
neutronically isolate the components from the reactor. The consequences
of dropping one of these small components into the vessel are bounded
by the In-Containment Fuel Handling Accident Analysis discussed in
Chapter 14.2.1 of the Turkey Point Updated Final Safety Analysis Report
(UFSAR). Therefore, the proposed change is bounded by the current and
the proposed In-Containment Fuel Handling Accident Analyses and will
not involve a significant increase in the probability or consequences
of an accident previously evaluated.
The proposed change to TS 3.9.4 would allow the containment
personnel airlock (PAL) doors to be open during fuel movement and core
alterations. Currently, a single PAL door is closed during fuel
movement and core alterations to prevent the escape of radioactive
material in the event of a in-containment fuel handling accident. The
PAL is not an initiator of an accident. Whether the PAL doors are open
or closed during fuel movement and core alterations has no affect on
the probability of any accident previously evaluated.
Allowing the PAL doors to be open during fuel movement and core
alterations does not increase the consequences from a fuel handling
accident. The calculated offsite doses are well within the limits of 10
CFR Part 100. In addition, the calculated doses are larger than the
expected doses because the calculation does not incorporate the closing
of the PAL door after the containment is evacuated. The proposed change
should significantly reduce the dose to workers in containment in the
event of a fuel handling accident by reducing the time required to
evacuate the containment. The proposed change will also significantly
decrease the wear on the PAL doors and, consequently, increase the
availability of the PAL doors in the event of an accident.
The proposed change to the footnote of TS 3.9.4 is administrative
in nature, and does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The changes being proposed do not affect assumptions contained in
plant safety analyses or the physical design of the plant, nor do they
affect Technical Specifications that preserve safety analysis
assumptions. Therefore, 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 analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The change in the definition of CORE ALTERATIONS would allow the
movement of a temporary source range detector or other small
components, such as cameras, tools, etc., within the reactor vessel
without the activity being considered CORE ALTERATIONS. The potential
exists however small, that an object can be dropped into the reactor
vessel. However, the justification for this change, is that the
insertion of small components into the reactor vessel will have no
effect on core reactivity since these items displace a small volume of
borated water, and sufficient borated water will surround the
components and provide the necessary neutron absorption to
neutronically isolate the components from the reactor. The consequences
of dropping one of these small components into the vessel are bounded
by the In-Containment Fuel Handling Accident Analysis discussed in
Chapter 14.2.1 of the Turkey Point UFSAR. Therefore the proposed change
is bounded by the current and the proposed In-Containment Fuel Handling
Accident Analyses and will not create the possibility of a new or
different kind of accident.
The proposed change to Specification 3.9.4 affects a previously
evaluated accident, i.e., in-containment fuel handling accident. Both
the current and the proposed In-Containment Fuel Handling Accident
Analysis assume that all of the iodines and noble gases that become
airborne within the containment escape and reach the site boundary and
low population zone with no credit taken for the containment building
barrier or for decay or deposition taken. Since the proposed change
does not involve the addition or modification of equipment nor does it
alter the design of plant systems and the revised analysis is
consistent with the current In-Containment Fuel Handling Accident
Analysis, the proposed change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed change to the footnote of TS 3.9.4 is administrative
in nature and does 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
amendments would not involve a significant reduction in a margin of
safety.
The change in the definition of CORE ALTERATIONS would allow the
movement of a temporary source range detector or other small
components, such as cameras, tools, etc., within the reactor vessel
without the activity being considered CORE ALTERATIONS. The potential
exists however small, that an object can be dropped into the reactor
vessel. However, the justification for this change, is that the
insertion of small components into the reactor vessel will have no
effect on core reactivity since these items displace a small volume of
borated water, and sufficient borated water will surround the
components and provide the necessary neutron absorption to
neutronically isolate the components from the reactor. The consequences
of dropping one of these small components into the vessel are bounded
by the Fuel Handling Accident Analysis discussed in Chapter 14.2.1 of
the Turkey Point UFSAR. Therefore, the proposed change is bound by the
current In-Containment Fuel Handling Accident Analyses and as a result
will not involve a significant reduction in a margin of safety.
The margin of safety as defined by 10 CFR Part 100 has not been
reduced. There is no increase in calculated offsite dose resulting from
a fuel handling accident in containment and the calculated dose is a
small fraction of the limits given in 10 CFR Part 100. The proposed
changes do not alter the bases for assurance that safety-related
activities are performed correctly or the basis for any Technical
Specification that is related to the establishment of or maintenance of
a safety margin. Therefore, operation of the facility in accordance
with the proposed amendments would not involve a significant reduction
in a margin of safety.
The proposed change to the footnote of TS 3.9.4 is administrative
in nature and does not relate to or modify the safety margins defined
in, and maintained by, the Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: Mohan C. Thadani, (Acting)

Florida Power and Light Company

Docket Nos. 50-250 and 50-251

Turkey Point Plant, Units 3 and 4, Dade County, Florida.
Date of amendment request: October 20, 1994.
Description of amendment request: This supersedes the licensee's
original request dated July 19, 1994, and noticed in the Federal
Register on August 3, 1994 (59 FR 39588). The licensee proposes to
change Turkey Point, Units 3 and 4 Technical Specifications (TS)
4.8.1.1.2e. and 4.8.1.1.2f., which address Emergency Diesel Generator
(EDG) fuel oil testing, by replacing the specific EDG fuel oil
Surveillance Requirements with the requirement to verify new and stored
EDG fuel oil in accordance with the Diesel Fuel Oil Testing Program. In
addition, the licensee proposes the addition of ACTION statements g.
and h., to TS 3.8.1.1, to address the required action in the event the
diesel fuel oil does not meet the Diesel Fuel Oil Testing Program
limits. The Diesel Fuel Oil Testing Program will be described in both
TS 6.8.4 and the BASES Section to the Technical Specifications. In
addition, FPL proposes revising TS 6.8.1 to include the requirement
that written procedures shall be established, implemented and
maintained for implementation of the Diesel Fuel Oil Testing Program.
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
amendments would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The proposed changes to the Technical Specifications will permit
the Technical Specification required testing of Emergency Diesel
Generator (EDG) fuel oil in accordance with the Turkey Point, Units 3
and 4 Diesel Fuel Oil Testing Program. The proposed change will permit
FPL to use more recent editions of the American Society for Testing and
Materials (ASTM) standards currently listed in Technical Specification
Surveillance Requirements 4.8.1.1.2e. and 4.8.1.1.2f. Prior to changing
the Diesel Fuel Oil Testing Program, the proposed change will be
evaluated pursuant to Title 10 Code of Federal Regulations Sec. 50.59
(10 CFR Sec. 50.59), ``Changes, tests, and experiments.'' Title 10 CFR
Sec. 50.59 permits a licensee to make changes in the procedures as
described in the safety analysis report without prior Commission
approval, provided that the proposed changes does not involve an
unreviewed safety question.
Title 10 CFR Sec. 50.59(a)(2) states that a proposed change
involves an unreviewed safety question (i) if the probability of
occurrence or the consequences of an accident or malfunction of
equipment important to safety previously evaluated in the safety
analysis report may be increased. Consequently, since any change to the
Diesel Fuel Oil Testing Program, including the ASTM standard or ASTM
edition standard to be used to evaluate EDG fuel oil acceptability, the
change must be evaluated relative to the more restrictive evaluation
criterion of 10 CFR Sec. 50.59, then 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 EDG fuel oil TS Surveillance Requirements will be
replaced with a requirement to test the EDG fuel oil in accordance with
the Turkey Point Units 3 and 4 Diesel Fuel Oil Testing Program.
ACTION statement g. of TS 3.8.1.1 is added to address the required
action in the event the new fuel oil properties do not meet the Diesel
Fuel Oil Testing Program limits. A failure to meet the American
Petroleum Institute (API) gravity, kinematic viscosity, flash point or
clarity limits is cause for rejecting the new fuel oil prior to the
addition to the Diesel Fuel Oil Storage Tanks, but does not represent a
failure to meet the Limiting Condition for Operation (LCO) of TS
3.8.1.1, since the new fuel oil has not been added to the storage
tanks. Provided these new fuel oil properties are met subsequent to the
addition of the new fuel oil to the storage tanks, 30 days is provided
to complete the analyses of the other fuel oil properties specified in
Table 1 of ASTM-D975-81, except sulfur which may be performed in
accordance with ASTM-D1552-79 or ASTM-D2622-82. In the event the other
new fuel oil properties specified in Table 1 of ASTM-D975-81 are not
met, ACTION statement g. of TS 3.8.1.1 provides an additional 30 days
to meet the Diesel Fuel Oil Testing Program limits. This additional 30
day period is acceptable because the fuel oil properties of interest,
even if they are not within limits, would not have an immediate effect
on EDG operation.
ACTION statement h. of TS 3.8.1.1 is added to address the required
action in the event the stored fuel oil total particulates do not meet
the Diesel Fuel Oil Testing Program limits. Fuel oil degradation during
long term storage shows up as an increase in particulate, due mostly to
oxidation. The presence of particulate does not mean the fuel oil will
not burn properly in a diesel engine. The frequency for performing
surveillance on stored fuel oil is based on stored fuel oil degradation
trends which indicate that particulate concentration is unlikely to
change significantly between surveillances.
Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil
Testing Program, FPL will need to determine if the proposed program
change is at least as, if not more, effective, in detecting
unsatisfactory fuel oil. The EDGs will thus continue to function as
designed and the probability or consequences of previously evaluated
accidents will be unaffected.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The proposed changes to the Technical Specifications will permit
the Technical Specification required testing of Emergency Diesel
Generator fuel oil using more recent editions of the American Society
for Testing and Materials (ASTM) standards currently listed in
Technical Specification Surveillance Requirements 4.8.1.1.2e. and
4.8.1.1.2f. Prior to changing the edition of the previously approved
ASTM standard being used to evaluate the EDG fuel oil, the proposed
edition standard will be evaluated pursuant to 10 CFR Sec. 50.59,
``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits a
licensee to make changes in the procedures as described in the safety
analysis report without prior Commission approval, provided that the
proposed changes does not involve an unreviewed safety question. Title
10 CFR Sec. 50.59(a)(2) states that a proposed change involves an
unreviewed safety question (ii) if a possibility for an accident or
malfunction of a different type than any evaluated previously in the
safety analysis report may be created. Consequently, since any change
to the edition of the ASTM standard to be used to evaluate EDG fuel oil
acceptability must be evaluated relative to the more restrictive
evaluation criterion of 10 CFR Sec. 50.59, then operation of the
facility in accordance with the proposed amendments would not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
ACTION statement g. of TS 3.8.1.1 is added to address the required
action in the event the new fuel oil properties do not meet the Diesel
Fuel Oil Testing Program limits. A failure to meet the API gravity,
kinematic viscosity, flash point or clarity limits is cause for
rejecting the new fuel oil prior to the addition to the Diesel Fuel Oil
Storage Tanks, but does not represent a failure to meet the Limiting
Condition for Operation (LCO) of TS 3.8.1.1, since the new fuel oil has
not been added to the storage tanks. Provided these new fuel oil
properties are met subsequent to the addition of the new fuel oil to
the storage tanks, 30 days is provided to complete the analyses of the
other fuel oil properties specified in Table 1 of ASTM-D975-81, except
sulfur which may be performed in accordance with ASTM-D1552-79 or ASTM-
D2622-82. In the event the other new fuel oil properties specified in
Table 1 of ASTM-D975-81 are not met, ACTION statement g. of TS 3.8.1.1
provides an additional 30 days to meet the Diesel Fuel Oil Testing
Program limits. This additional 30 day period is acceptable because the
fuel oil properties of interest, even if they are not within limits,
would not have an immediate effect on EDG operation.
ACTION statement h. of TS 3.8.1.1 is added to address the required
action in the event the stored fuel oil total particulates does not
meet the Diesel Fuel Oil Testing Program limits. Fuel oil degradation
during long term storage shows up as an increase in particulate, due
mostly to oxidation. The presence of particulate does not mean the fuel
oil will not burn properly in a diesel engine. The frequency for
performing surveillance on stored fuel oil is based on stored fuel oil
degradation trends which indicate that particulate concentration is
unlikely to change significantly between surveillances.
Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil
Testing Program, FPL will need to determine if the proposed program
change is at least as, if not more, effective, in detecting
unsatisfactory fuel oil. Since the proposed changes do not involve a
change in the design of any plant system or component, and since the
proposed changes will need to evaluate the effect of any ASTM standard
edition change on the level of EDG reliability, the change proposed
will 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
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes to the Technical Specifications will permit
the Technical Specification required testing of Emergency Diesel
Generator (EDG) fuel oil using more recent editions of the American
Society for Testing and Materials (ASTM) standards currently listed in
Technical Specification Surveillance Requirements 4.8.1.1.2e. and
4.8.1.1.2f. Prior to changing the edition of the previously approved
ASTM standard being used to evaluate the EDG fuel oil, the proposed
edition standard will be evaluated pursuant to 10 CFR Sec. 50.59,
``Changes, tests, and experiments.'' Title 10 CFR Sec. 50.59 permits a
licensee to make changes in the procedures as described in the safety
analysis report without prior NRC approval, provided that the proposed
changes does not involve an unreviewed safety question. Title 10 CFR
Sec. 50.59(a)(2) states that a proposed change involves an unreviewed
safety question (iii) if the margin of safety as defined in the basis
for any technical specification is reduced. Consequently, since any
change to the edition of the ASTM standard to be used to evaluate EDG
fuel oil acceptability must be evaluated relative to the more
restrictive evaluation criterion of 10 CFR Sec. 50.59, then operation
of the facility in accordance with the proposed amendments would not
involve a significant reduction in a margin of safety.
ACTION statement g. of TS 3.8.1.1 is added to address the required
action in the event the new fuel oil properties do not meet the Diesel
Fuel Oil Testing Program limits. A failure to meet the API gravity,
kinematic viscosity, flash point or clarity limits is cause for
rejecting the new fuel oil prior to the addition to the Diesel Fuel Oil
Storage Tanks, but does not represent a failure to meet the Limiting
Condition for Operation (LCO) of TS 3.8.1.1, since the new fuel oil has
not been added to the storage tanks. Provided these new fuel oil
properties are met subsequent to the addition of the new fuel oil to
the storage tanks, 30 days is provided to complete the analyses of the
other fuel oil properties specified in Table 1 of ASTM-D975-81, except
sulfur which may be performed in accordance with ASTM-D1552-79 or ASTM-
D2622-82. In the event the other new fuel oil properties specified in
Table 1 of ASTM-D975-81 are not met, ACTION statement g. of TS 3.8.1.1
provides an additional 30 days to meet the Diesel Fuel Oil Testing
Program limits. This additional 30 day period is acceptable because the
fuel oil properties of interest, even if they are not within limits,
would not have an immediate effect on EDG operation.
ACTION statement h. of TS 3.8.1.1 is added to address the required
action in the event the stored fuel oil total particulates does not
meet the Diesel Fuel Oil Testing Program limits. Fuel oil degradation
during long term storage shows up as an increase in particulate, due
mostly to oxidation. The presence of particulate does not mean the fuel
oil will not burn properly in a diesel engine. The frequency for
performing surveillance on stored fuel oil is based on stored fuel oil
degradation trends which indicate that particulate concentration is
unlikely to change significantly between surveillances.
Prior to changing the Turkey Point Units 3 and 4 Diesel Fuel Oil
Testing Program, FPL will need to determine if the proposed program
change is at least as, if not more, effective, in detecting
unsatisfactory fuel oil. Since the proposed changes will require a
safety evaluation to assure that the reliability of the EDGs using fuel
oil tested in accordance with the different ASTM standard edition
maintains the current margin of safety, 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 request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: Mohan C. Thadani, Acting.

Florida Power Corporation, et al.

Docket No. 50-302

Crystal River Nuclear Generating Plant, Unit No. 3, Citrus County,
Florida.
Date of amendment request: September 30, 1994.
Description of amendment request: The proposed amendment would
revise the Crystal River 3 (CR3) Nuclear generating Plant Technical
Specifications (TS) to allow an increase in the rated thermal power
(RTP) for CR-3 from the current 2544 level to 2568 Megawatt thermal
(Wt). Accordingly, in TS 1.1, ``Definitions,'' would be revised to
indicate the new power level of 2568 MWt. The proposed change would not
require any hardware modifications.
Basis for proposed no significant hazards consideration
determination: Currently, CR-3 is operating at a maximum RTP of 2544
MWt. The licensee proposes to operate at a maximum RTP of 2568 MWt, an
increase of 24 MWt over the current licensed power of 2544 MWt.
The licensee states that the Babcock and Wilcox (B&W) 177 Fuel
Assembly (FA) Nuclear Steam Supply System (NSSS) in the CR3 design is
capable of operating at a thermal power level of 2772 MWt. Due to
limitations in the secondary area of the plant, the licensee requests
authorization to operate at 2568 MWt which is less than the design
level of 2772 MWt. The licensee performed a detailed engineering study
on this power increase.
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the probability
of occurrence or consequences of an accident previously evaluated. The
thermal-hydraulic and nuclear characteristics of the reactor core were
originally designed for a rated thermal power of 2568 MWt or higher.
Therefore, the proposed thermal power increase to the reference power
level of 2568 MWt does not change the original design assumptions and
analyses for the reactor core. Most of the design basis accidents and
transients were originally evaluated at the proposed power level. As
described more fully in this submittal, those transients and accidents
that were not originally evaluated at 2568 MWt were re-evaluated using
CR-3 FSAR [Final Safety Analysis Report] Chapter 14 accident sequence
of events, reactor protection criteria, and approved calculational
methods. Based on this evaluation and initial plant design evaluations,
FPC [Florida Power Corporation, the licensee for CR3] has determined
that the probability and consequences of design basis transients and
accidents are not significantly increased and that the radiological
consequences from the design basis transients and accidents remain well
below 10 CFR 100 limits.
FPC has also reviewed CR-3 balance of plant and safety related
systems to determine which systems and components could be affected by
the proposed power increase. The changes to the reactor coolant system
and secondary conditions and parameters are discussed in this
submittal. These changes are minor in nature. The only Technical
Specification change is to revise the reference power to 2568 MWt. No
facility modifications will be required. FPC evaluated the systems and
components and concluded that these systems and components will
continue to perform within their design parameters with the unit
operating at 2568 MWt.
Based on the foregoing, the proposed amendment does not
significantly increase the probability or consequences of an accident
previously evaluated.
2. The proposed thermal power increase does not create the
possibility of a new or different kind of accident from previously
evaluated accidents. As noted above, the thermal-hydraulic and nuclear
characteristics of the reactor core were originally designed for
operation at the proposed thermal power. Therefore, operation at the
proposed power level does not introduce new or different performance
characteristics that create the possibility of a new or different kind
of accident.
FPC has also reviewed CR-3 safety-related systems and balance of
plant systems to determine which systems could be affected by the
proposed power increase and the resultant minor changes in plant
parameters and operating conditions. Systems that could be affected
were evaluated using the licensing basis criteria described in the CR-3
FSAR to assure their adequacy at the increased power level. Included in
these evaluations were plant features that are not power level related
or directly affected by an increase in power level, as well as,
associated issues such as environmental considerations. Equipment
performance and plant operation were evaluated with respect to actual
performance versus projected operating conditions to identify any
hardware modifications required to achieve the upgraded power. Based on
this evaluation, FPC has determined that all systems will continue to
perform within their design parameters at 2568 MWt and that no physical
modifications to these systems will be necessary to accommodate a 2568
MWt rating. Only minor re-calibration of plant instrumentation to
reflect the increased power will be needed. The proposed power level
does not introduce any new performance characteristics or modes of
operation for plant systems and components, and does not introduce any
new failure modes.
Based on the foregoing, the proposed amendment does not create the
possibility of a new or different kind of accident.
3. The proposed amendment does not involve a significant reduction
in a margin of safety. The thermal-hydraulic and nuclear
characteristics of the reactor core were originally designed for
operation at the proposed power level. Most of the design basis
transients and accidents were originally analyzed assuming a power
level of 2568 MWt or higher. As described more fully in this submittal,
those transients and accidents that were not originally analyzed at
2568 MWt were re-evaluated using CR-3 FSAR Chapter 14 accident sequence
of events, reactor protection criteria, and approved calculational
methods. FPC has determined that operation with the proposed thermal
power will be bounded by the original analyses. In addition, FPC's
evaluation of affected plant systems and components revealed that plant
systems and components will continue to operate within their design
parameters with no significant change in a margin of safety.
Based on the foregoing, the proposed amendment does not involve a
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 32629
Attorney for licensee: A. H. Stephens, General Counsel, Florida
Power Corporation, MAC-A5D, P. O. Box 14042, St. Petersburg, Florida
33733.
NRC Project Director: Mohan C. Thadani, (Acting).

Indiana Michigan Power Company

Docket Nos. 50-315 and 50-316

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan.
Date of amendment request: August 3, 1994.
Description of amendment requests: The proposed amendments would
allow the radiological effluent technical specifications (TS) to be
relocated to other controlled documents. Procedural details contained
in the current radiological effluents TS have been relocated to either
the Offsite Dose Calculation Manual (OCDM) or the Process Control
Program (PCP), as applicable. Proposed revisions to the OCDM and PCP
have been prepared in accordance with the proposed changes to the
administrative controls section of the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1
The changes described above in no way negatively impact the
requirements of the T/Ss. Separating the turbine room sump releases
from the others is purely a clarification of the method we handle
releases. The six ground monitoring wells added to the T/S table
updates our current monitoring practice. With the six extra wells to
monitor, we exceed the monitoring requirements of the T/Ss. Therefore,
it is concluded that the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
Criterion 2
No changes to the LCOs for either T/S are proposed as part of this
amendment request. The proposed change does not involve any physical
changes to the plant or any changes to plant operations. The changes
merely propose to update our methods of implementing the T/S with our
current practices. Thus, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3
The changes described above in no way negatively impact the
requirements of the T/Ss. Separating the turbine room sump releases
from the others is purely a clarification of the method we handle
releases. The six ground monitoring wells added to the T/S table
updates our current monitoring practice. With the extra wells to
monitor, we exceed the monitoring requirements called for in the T/Ss.
Therefore, it is concluded that the proposed changes do not involve a
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: John N. Hannon.

Niagara Mohawk Power Corporation

Docket No. 50-410

Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.
Date of amendment request: October 5, 1994.
Description of amendment request: The proposed license amendment
would revise the applicability requirements of Technical Specification
(TS) 3.7.3 to require operability of the Control Room Outdoor Air
Special Filter Train System in Operational Conditions 1, 2, 3 and **
(when irradiated fuel is being handled in the reactor building and
during CORE ALTERATIONS and operations with a potential for draining
the reactor vessel and uncovering irradiated fuel) rather than in all
Operational Conditions and * * *. The applicability requirements for
Action Statement b of TS 3.7.3 and for the Radiation Monitoring
Instrumentation required operable by TS Tables 3.3.7.1-1 and 4.3.7.1-1
would be changed in a similar manner. The proposed amendment would also
add a notation to Action Statement b.1 of TS 3.7.3 stating that the
provisions of Specification 3.0.4 are not applicable provided an
operable control room filter train is in the emergency pressurization
mode of operation. The licensee stated that these proposed changes are
consistent with the requirements of the NRC's Improved Standard
Technical Specifications (NUREG-1433) and with Generic Letter 87-09,
``Section 3.0 and 4.0 of the Standard Technical Specifications (STS) on
the Applicability of Limiting Conditions for Operation and Surveillance
Requirements.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The Control Room Outdoor Air Special Filter Train System is not an
initiator or precursor to an accident. The Control Room Outdoor Air
Special Filter Train System responds to a release of radioactivity to
the outside environment as detected in the air supply to the control
room by providing a radiologically controlled environment within the
control room. In operational conditions 4 and 5, the probability and
consequences of a design basis accident are reduced due to the pressure
and temperature limitations in these operational conditions. Therefore,
maintaining the chiller subsystem operable is not required in
operational conditions 4 and 5, except for the * * * operational
condition. Therefore, a change to applicability and action statements
of LCO [Limiting Condition For Operation] 3.7.3 cannot affect the
probability of a previously evaluated accident.
All accidents which take credit for operation of the Control Room
Outdoor Air Special Filter Train System in the emergency pressurization
mode of operation are analyzed and presented in Chapter 15 of the USAR
[Updated Safety Analysis Report]. These accidents can only occur in
operational conditions 1, 2, 3 and * * *.
Accordingly, the proposed change in the applicability of LCO 3.7.3
from all operational conditions (i.e., 1, 2, 3, 4, 5 and * * *) to
operational conditions 1, 2, 3 and * * * does not significantly
increase the consequences of an accident previously evaluated. The
proposed change to action statement b of LCO 3.7.3 and to Tables
3.3.7.1-1 and 4.3.7.1-1 of LCO 3.3.7.1 is consistent with the above
change.
Sections 15.7.4 and 15.7.5 of the USAR evaluate a fuel handling
accident and a spent fuel cask drop accident, respectively. The
radiological evaluation of these accidents considers the unfiltered
radioactivity that enters the control room prior to the automatic
operation of the Control Room Outdoor Special Filter Train System in
the emergency pressurization mode of operation. The radiological
consequences of these accidents are within the limits of GDC [General
Design Criterion]-19.
With one control room filter train inoperable and prior to entering
the operational condition, the proposed change to action statement b.1
of LCO 3.7.3 would require an operable control room filter train be
placed in the emergency pressurization mode of operation. During an
accident involving the release of radioactivity to the environment, an
operable control room filter train would already be running in the
emergency pressurization mode and performing its safety function,
thereby preventing the entry of unfiltered radioactivity into the
control room. Therefore, if a fuel handling accident or a spent fuel
cask drop accident were to occur and release radioactivity, the control
room personnel radiological doses would be less than the doses depicted
in the USAR. Accordingly, the Technical Specification change to action
statement b.1 does not significantly increase the consequences of a
previously evaluated accident.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This amendment does not involve any accident precursors or
initiators. In addition, this amendment does not require any changes to
plant equipment.
During an accident involving the release of radioactivity to the
environment an operable control room filter train would already be
running in the emergency pressurization mode and performing its safety
function. Furthermore, the operating status of a running control room
filter train would be unaffected by the receipt of an automatic start
signal due to high radiation in either air intake to the Control Room
Outdoor Air Special Filter Train System. Therefore, the proposed
amendment will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed change in the applicability of LCO 3.7.3 from all
operational conditions (i.e., 1, 2, 3, 4, 5 and * * *) to operational
conditions 1, 2, 3 and * * * is consistent with the safety analysis
contained in the USAR. The proposed changes to action statement b of
LCO 3.7.3 and to Tables 3.3.7.1-1 and 4.3.7.1-1 of LCO 3.3.7.1 is
consistent with the above change.
Entry into the ** operational condition for LCO 3.7.3 with one
control room filter train inoperable and the other control room filter
train operable and operating in the emergency pressurization mode
provides a comparable level of safety to two operable non-running
control room filter trains. The remedial measure prescribed by
Technical Specification action statement b.1 (placing an operable
control room filter train in the emergency pressurization mode of
operation) for which the exception to LCO 3.0.4 is proposed provides a
sufficient level of protection to permit operational mode changes and
safe long-term operation of NMP2 [Nine Mile Point Unit 2] consistent
with the licensing basis described in the USAR. Therefore, the proposed
change to action statement b.1 is consistent with Generic Letter 87-09,
``Sections 3.0 and 4.0 of the Standard Technical Specifications (STS)
on the Applicability of Limiting Conditions for Operation and
Surveillance Requirements.'' Accordingly, this change will not
significantly reduce the margin of safety.
This proposed amendment is consistent with the Improved Standard
Technical Specifications, NUREG-1433. Accordingly, as determined by the
analysis above, this proposed 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: 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: Ledyard B. Marsh.

Niagara Mohawk Power Corporation

Docket No. 50-410

Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York.
Date of amendment request: October 21, 1994.
Description of amendment request: The proposed amendment would add
a footnote to Technical Specification (TS) 4.8.1.1.2.e.8 which would
permit performance of the 24-hour functional test of the emergency
diesel generators (EDGs) during power operation. TS 4.8.1.1.2.e.8
currently requires the 24-hour functional test of the EDGs be performed
at least once per 18 months during shutdown; the proposed amendment
would permit this testing to be performed during power operation
provided the other two EDGs are operable. If either of the other two
EDGs become inoperable, the test would be aborted.
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 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to permit the 24 hour functional test of the
diesels to be performed during power operation does not increase the
chances for a previously analyzed accident to occur. The function of
the diesels is to supply emergency power in the event of a loss of
offsite power. Operation of the diesels is not a precursor to any
accident. Furthermore, the diesel generator being tested will remain
operable and will be available to supply emergency loads within the
required time. In addition, the two remaining diesel generators will be
operable during the test. Consequently, if an offsite disturbance were
to occur that affected the operability of the diesel being tested, the
two remaining diesels would be capable of feeding the loads necessary
for safe shutdown of the plant. This addresses the concerns raised in
Information Notice 84-69 regarding the operation of emergency diesel
generators connected in parallel with offsite power. In summary, the
proposed changes do not adversely affect the performance or the ability
of the diesel generators to perform their intended function.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed amendment to the 24 hour functional surveillance test
will not affect the operation of any safety system or alter its
response to any previously analyzed accident. The diesel will
automatically transfer from the test mode if necessary to supply
emergency loads in the requried time. The test mode is used for the
monthly surveillance of the diesel generators as well, therefore, no
new plant operating modes are introduced. In the event the diesel fails
the functional test it will be declared inoperable and the actions
required for an inoperable diesel will be performed. The remaining two
diesel generators will be operable and are capable of feeding the loads
necessary for safe shutdown of the plant.
Therefore, 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 2, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed amendment will not reduce availability of the diesel
generator being tested to provide emergency power in the event of a
loss of offsite power. If a loss of offsite power or a loss of coolant
accident occurs during the surveillance test, the emergency bus would
de-energize and shed load. The diesel generator would then transfer
from the test mode to the emergency mode. It would then be available to
automatically supply emergency loads. In addition, the two remaining
generators will be operable during the test. Consequently, if an
offsite disturbance were to occur that affected the operability of the
diesel begin tested, the two remaining diesels would be capable of
feeding the loads necessary for safe shutdown of the plant. The time
required for the diesel being tested to pick up emergency loads will
not be affected by performing the 24 hour functional test during power
operation.
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 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: Ledyard B. Marsh.

Northeast Nuclear Energy Company et al.

Docket No. 50-336.

Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut.
Date of amendment request: October 18, 1994.
Description of amendment request: The proposed amendment would
require three type A overall Integrated Containment Leakage Tests be
conducted at approximately equal intervals during shutdowns during each
10-year service period. For the third Type A test for the second 10-
year period, it would be conducted during the thirteenth refueling
outage extending the second 10-year service period to the end of the
thirteenth refueling outage. The amendment would also change the
Containment Leakage Bases by reflecting the conditions of a proposed
exemption to 10 CFR 50, Appendix J, that would remove the requirement
that the third Type A test for each 10-year period be conducted when
the plant is shutdown for the 10-year plant inservice inspection.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
* * * The proposed changes do not involve a SHC [significant
hazards consideration] because the change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Type A tests are performed to ensure that the total leakage from
containment does not exceed the maximum allowable primary containment
leakage rate at the design pressure. This ensures compliance with the
dose limits of 10 CFR 100.
The proposal to revise Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 2 Technical Specifications will increase the
flexibility for scheduling the Type A tests. It does not modify the
maximum allowable leakage rate at the design containment pressure, does
not impact the design basis of the containment, and does not make any
physical or operational changes to existing plant structures, systems,
or components.
The first two Type A tests of the second 10-year service period for
Millstone Unit No. 2 have been conducted. The results of these tests
demonstrate that Millstone Unit No. 2 has maintained control of
containment integrity by maintaining margin between the acceptance
criterion and the ``As-Found'' and ``As-Left'' leakage rates.
Historically, Type A tests have a relatively low failure rate where
Type B and C testing (local leakage rate tests) could not detect the
leakage path. Most Type A test failures are attributed to failures to
Type B or C components (containment penetrations and isolation valves).
Type B and C components are tested per Surveillance Requirement
4.6.1.2.d for the Millstone Unit No. 2 Technical Specifications. These
tests are required to be conducted at intervals no greater than 24
months, and the acceptance criterion for the combined leakage rate for
all penetrations and valves subject to the Type B and C tests is 0.6
La. These local leakage rate tests provide assurance that
containment integrity is maintained. The relatively low ``As-Left''
Type B and C total leakage resulting from the past outage indicates
that the leakage has been maintained within the technical specification
acceptance criterion. The Type B and C tests will continue to be
performed in accordance with the requirements of Surveillance
Requirement 4.6.1.2.d. However, on September 26, 1994, NNECO submitted
a request for a one-time technical specification change, request for
enforcement discretion, and a request for a scheduler exemption from
Appendix J to 10 CFR 50 regarding the Schedule for Type B and C
testing. The NRC verbally granted enforcement discretion on September
24, 1994, and written enforcement discretion on September 30, 1994. The
schedular exemption request was granted on October 12, 1994.
The previous Type A, B, and C tests demonstrate that Millstone Unit
No. 2 has maintained control of containment integrity by maintaining a
conservative margin between the acceptance criterion and the ``As-
Found'' and ``As-Left'' leakage results. Based on this, the Millstone
Unit No. 2 containment is considered to be in sound condition. No
operations are known to have occurred which would suggest any
substantial degradation of these results.
Based on the above, the proposal to revise Surveillance Requirement
4.6.1.2.a of the Millstone Unit No. 2 Technical Specifications does not
involve a significant increase in the probability or consequences of an
accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposal to revise Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 2 Technical Specifications will increase the
flexibility in scheduling the Type A tests. It does not make any
physical or operational changes to existing plant structures, systems,
or components. In addition, the proposal does not modify the acceptance
criterion for the Type A tests. Maintaining the leakage through the
containment boundary to the atmosphere within a specific value ensures
that the plant complies with the requirements of 10 CFR 100. The
containment boundary serves as an accident mitigator; it is not an
accident initiator. Therefore, the proposal to revise Surveillance
Requirement 4.6.1.2.a does not create the possibility of a new or
different kind of accident from any previously analyzed.
3. Involve a significant reduction in the margin of safety.
The proposal to revise Surveillance Requirement 4.6.1.2.a of the
Millstone Unit No. 2 Technical Specifications will increase the
flexibility for scheduling the Type A tests. It does not modify the
maximum allowable leakage rate at the design containment pressure, does
not impact the design basis of the containment, and does not make any
physical or operational changes to existing plant structures, systems,
or components.
The first two Type A tests of the second 10-year service period for
Millstone Unit No. 2 have been conducted. The results of these tests
demonstrate that Millstone Unit No. 2 has maintained control of
containment integrity by maintaining margin between the acceptance
criterion and the ``As-Found'' and ``As-Left'' leakage rates.
Additionally, the results of the last Type B and C tests had
significant margin with respect to the acceptance criterion. Based on
the previous Type A, B, and C tests, the Millstone Unit No. 2
containment is considered to be in sound condition. No operations are
known to have occurred which would suggest any substantial degradation
of these results.
Based on the above, the proposal 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: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee.

Northeast Nuclear Energy Company, et al.

Docket No. 50-423

Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut.
Date of amendment request: September 30, 1994.
Description of amendment request: The licensee has proposed to
revise the Technical Specifications (1) to clarify the definition of
core alterations, (2) to change the verbiage in the Limiting Condition
For Operation (LCO) addressing the refueling operations, (3) to make
changes to three surveillance requirements involving source range
instrumentation, and (4) to change the LCO regarding the Residual heat
Removal and coolant circulation water levels to be consistent with the
guidance provided in NUREG-1431.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do 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.
Boron Dilution in Mode 6--A boron dilution in Mode 6 is precluded
by technical specification requirements to close and lock all dilution
source valves. There is a provision for dilution valves to be opened
under administrative controls; in this case, cautionary measures will
be taken to control and monitor the reactivity addition. Deletion of
the source range analog operational test prior to core alterations will
not impact an accident previously evaluated since the sources range
monitors are verified operable prior to entry into Mode 6 and every 7
days thereafter. The change in definition for a core alteration means
that components which do not effect reactivity may be moved within the
reactor vessel without any additional condition such as direct
supervision of an SRO.
Since a boron dilution would not be initiated by movement of
nonfuel components within the reactor vessel, it is not impacted by the
change in definition of a core alteration.
Inadvertent Loading of a Fuel Assembly--Movement of a fuel assembly
would be performed as a core alteration under the supervision of an
SRO, therefore, it would not be impacted by the change to the
definition of a core alteration. The change to the source range
monitors also will not affect the probability of occurrence of a
misloaded fuel assembly since this accident is precluded by
administrative controls, as well as the source range monitors. Also,
there will be no degradation in the reliability or accuracy of the
source range monitors due to this change. The deletion of the
requirement to perform the analog channel operational test within eight
hours prior to core alterations will not impact performance of the
monitors, since they have to be checked prior to entry into Mode 6 and
every 7 days thereafter.
Fuel Handling Accident--Movement of fuel will not affect this
accident, because it will still be considered a core alteration.
Therefore, there is no effect on the probability of a fuel handling
accident. The source range monitors are not involved in the occurrence
of a fuel handling accident. The fuel handling accident is the only
accident considered here with radiological consequences. It will not be
impacted by the proposed changes.
Loss of RHR in Mode 6--The probability of this accident will not be
changed since the new requirement is the same as before. As before, RHR
may be secured for up to one hour per eight-hour period and boron
dilution operations may not be performed with RHR secured (although
this requirement is being added to the notes, the requirement is also
given elsewhere in the technical specifications). Additionally, the
existing reactor coolant system (RSC) temperature limits must still be
met.
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 analyzed.
All required systems will continue to operate as before. Therefore,
there is no possibility of a new or different kind of accident. The
deletion of the source range analog channel operational test within
eight hours prior to core alterations will not affect the performance
of the monitors since they will have had this test completed prior to
entry into Mode 6 and every 7 days thereafter. The change in definition
of a core alteration cannot create the possibility of a new type of
accident because those initiating events for accidents will remain
classified as core alterations.
3. Involve a significant reduction in the margin of safety.
The margin of safety for the above listed accidents will remain as
before.
a. Boron dilution in Mode 6--This accident calculates the time from
receipt of a shutdown margin monitor dilution alarm until the core
reaches criticality. Since this time is not changed, there is no
reduction in the margin of safety. In this case, the dilution is
precluded by administrative controls which will not be impacted by the
proposed changes.
b. Inadvertent Loading of a Fuel Assembly--Technical Specification
3.9.1.1 protects against this accident by requiring sufficient boron in
the RCS to prevent criticality for any core configuration including two
stuck RCCAs [rod cluster control assemblies] in the fully withdrawn
position. Since this requirement will not change, the margin of safety
will not change.
c. Fuel Handling Accident--The margin of safety for the
radiological limits is not changed.
d. Loss of RHR--Changes are editorial due to the revised definition
of a core alteration. There is no change to the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L.M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee.

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 amendment requests: October 3, 1994.
Description of amendment requests: The proposed amendment would
revise Prairie island Nuclear Generating Plant Technical Specification
4.6, ``Periodic Testing of Emergency Power Systems.'' Specifically, the
proposed amendment would modify the emergency diesel generator (EDG)
24-hour load test requirements to provide a indicated load range of
103-110% of the continuous rating. The proposed amendment would also
rephrase various EDG test requirements to provide clarity and delete
the requirement to verify that the auto-connected loads do not exceed
3000 kw (Unit 2 5100kw).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously evaluated.
Changing the specification from ``unit'' to ``diesel generator''
does not change the intent of the specification, it merely clarifies
the original intent and therefore cannot involve a change in the
probability or consequences of an accident.
Changing the 22-hour lower range limit from a load of 90% to an
indicated load of 92% removes possible ambiguity from the specification
but does not change the actual requirement, therefore it cannot involve
a change in the probability or consequences of an accident.
Removing the 22-hour upper range limit from the specification does
not reduce the conservatism of the test since operating at a higher
load provides more evidence of the ability of the machine to carry the
accident loads. For this reason, this change will not involve any
increase in the consequences of an accident. Also, increasing the load
at which the diesel generator is tested cannot affect the probability
of an accident.
The NRC staff has pointed out, in Generic Letter 88-15, the hazards
of testing the Diesel Generator at a load greater than the design
rating. The proposed change is intended to ensure that the design
rating is not inadvertently exceeded. Since the recent installation of
two additional emergency diesel generators, the highest anticipated
event loads are: Unit 1-2414kW, Unit 2-3813 kW. For these diesel
generators, then, 103% of the continuous ratings:
Unit 1, 103% of 2750 kW (continuous rating) = 2832.5 kW
represents 117.3% of the highest anticipated event load and;
Unit 2, 103% of 5400 kW (continuous rating) = 5562 kW
represents 145.9% of the highest anticipated event load.
A test load of 103%, therefore would still be significantly greater
than the load required during accident conditions. Since an adequate
level of electrical load carrying capacity of the diesel generators
(and thus their accident mitigating functions) would still be
demonstrated by the surveillance test, the consequences of an accident
would be unaffected by the proposed change. The probability of
occurrence of a previously evaluated accident would be unaffected since
testing a diesel generator at load between 103 and 110 percent instead
of at load between 105 and 110 percent could not cause or contribute to
the initiation of an accident. For these reasons, this change could
have no effect on the probability or consequences of an accident
previously evaluated.
Allowing momentary transients outside of the test band does not
affect the conduct of the test, it merely allows momentary swing
outside the specified band to not invalidate the test. Not allowing
momentary transients would not prevent them, it would only require
conducting the test longer until the specified time period was achieved
without moving outside the band. Since the machine will not be operated
any differently, this specification change cannot affect the
probability or consequences of an accident previously evaluated.
Proposed changes A, B, C, D, and the first part of E [identified as
such in the submittal] are intended to clarify the meaning of the
existing specifications without changing the requirements. For this
reason, these proposed changes to the Technical Specifications will not
change the manner in which the plant is operated or maintained. These
administrative changes, therefore, will effect on the probability or
consequences of an accident previously evaluated.
The second part of E (verification of the bypass of diesel
generator trips during a simulated safety injection signal vs
concurrent safety injection and loss of offsite power signals) does not
change the intended function which is to be tested but, rather, reduces
the special conditions (temporary electrical jumpers to simulate the
loss of offsite power) in which the plant needs to be placed in order
to perform the test.
Proposed change F (removal of the verification that the auto-
connected load do not exceed 3000 or 5100 kW) does not reduce the
assurance of the ability of the diesel generators to perform the
accident mitigation functions since this verification is performed by
other, more pertinent, means.
Therefore, these changes cannot increase the probability or
consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a new
or different king of accident from any accident previously analyzed.
Changing the specification from ``unit'' to ``diesel generator''
does not change the intent of the specification, it merely clarifies
the original intent and therefore cannot create the possibility of a
new or different kind of accident.
Changing the 22-hour lower range limit from a load of 90% to an
indicated load of 92% removes possible ambiguity from the specification
but does not change the actual requirement.
Removing the 22-hour upper range limit from the specification does
not change the manner in which the surveillance is performed. It only
affects whether the time spent above 100% load can be counted toward 22
hours in the 22-hour portion of the test. This change would not allow
any new modes of operation nor does it allow any modification to the
plant.
As stated above, testing a diesel generator at a load between 103
and 110% instead of between 105 and 110% could not cause or contribute
to the initiation of an accident.
Allowing momentary transients outside of the test band does not
affect the conduct of the test, it merely allows momentary swings
outside the specified band to not invalidate the test. Not allowing
momentary transients would not prevent them, it would only require
conducting the test longer until the specific time period was achieved
without moving outside the band.
Therefore, for these reasons, operation of the facility in
accordance with the proposed amendment will not create the possibility
of a new or different kind of accident from any accident previously
analyzed.
As stated above [for changes A-F], the proposed changes will not
cause a change in the way in which the plant is operated or maintained,
excepted for the reduction of the special conditions in which the plant
needs to be placed in order to test the bypass of the diesel generator
trips. Therefore, these administrative changes will not create the
possibility of a new or different kind of accident from any accident
previously analyzed.
3. The proposed amendment will not involve a significant reduction
in a margin of safety.
Changing the specification from ``unit'' to ``diesel generator''
does not change the intent of the specification, it merely clarifies
the original intent and therefore cannot affect the margin of safety.
Changing the 22-hour lower range limit from a load of 90% to an
indicated load of 92% removes possible ambiguity from the specification
but does not change the actual requirement and therefore cannot affect
the margin of safety.
The margin of safety is not affected by removal of the 22-hour
upper range limit on the operation of the diesel generators during
surveillance testing since the margin of safety is related to the
magnitude of the accident loads and the maximum capacity of the machine
to carry load and this margin would be unaffected by this change.
The capacity of each diesel generator to carry electrical load can
not be diminished by being tested at a lower load. Also, load testing
to less than 105% but more than 103% does not lessen the confidence in
the ability of the diesel generators to carry adequate load for this
facility since these diesel generators have significantly greater load
capacity than required by Standard Review Plan guidance in this regard
(the guidance allows peak accident load up to 100% of the continuous
rating versus Unit 1 diesel generators peak accident load of 87.8% and
Unit 2 diesel generators peak accident load of 70.6%). Therefore, this
change will not involve a significant reduction in the margin of
safety.
Allowing momentary transients outside of the test band does not
affect the conduct of the test, it merely allows momentary swings
outside the specified band to not invalidate the test. Not allowing
momentary transients would not prevent them, it would only require
conducting the test longer until the specified time period was achieved
without moving outside the band. Since the machine will not be operated
any differently per the new specification, the margin of safety is
unaffected.
As stated above [for changes A-F], the proposed changes will not
cause a change in the way in which the plant is operated or maintained,
except for the reduction of the special conditions in which the plant
needs to be placed in order to test the bypass of the diesel generator
trips. Therefore, these administrative 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 requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: John N. Hannon.

Omaha Public Power District

Docket No. 50-285

Fort Calhoun Station, Unit No. 1, Washington County, Nebraska.
Date of amendment request: October 7, 1994.
Description of amendment request: The proposed amendment to the
Technical Specifications (TSs) would (1) delete the surveillance
requirements contained in TS 3.6(3)a for the raw water backup valves to
the containment cooling coils, (2) delete the surveillance requirements
contained in TS 3.2, Table 3-5, item 6, for raw water valves, and (3)
revise the basis of TS 2.4 to reflect these changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The deletion of surveillance requirements contained in Technical
Specifications (TS) 3.2, Table 3-5, Items 6 and 3.6(3)a does not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
TS 3.6(3)a requires the Raw Water (RW) backup valves to the
containment air coolers to be tested each refueling outage. In 1990,
during the process of reviewing several open items created by the
design basis reconstitution project, an engineering analysis determined
that RW direct cooling of the containment air cooling coils should not
be used after an accident that has created elevated temperature
conditions inside containment. The high containment air temperatures,
in conjunction with the low back pressure in the containment cooling
coils when in the RW direct cooling mode, introduces the possibility of
vaporization inside the coils. Therefore, the use of RW direct cooling
for the containment air coolers has been discontinued in post-Loss of
Coolant Accident (LOCA) or post-Main Steam Line Break (MSLB)
situations. The issue of not being able to utilize RW direct cooling to
the containment air cooling coils was reported to the NRC in LER-90-25,
dated October 29, 1990 and LER-90-25 Revision 1, dated December 17,
1990.
Raw water direct cooling of the containment air coolers is possible
if the containment atmospheric temperatures are less that 150 deg.F. If
RW direct cooling of the containment air coolers was utilized after a
LOCA or MSLB accident, it could only be used for long-term containment
atmospheric cooling. These conditions are essentially equivalent to
that associated with conditions in containment during normal plant
operation. RW direct cooling of the containment air coolers is not a
required post-accident function to maintain containment pressure below
60 psig. Since these valves are not required to perform a post-accident
function, deletion of the requirements to test these valves does not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
TS 3.2, Table 3-5, Item 6 requires that valves in the RW system be
tested every refueling outage. The valves tested by this surveillance
that could perform a safety function are already tested in accordance
with TS 3.3(1). Therefore testing of these valves under TS 3.2, Table
3-5, Item 6 is redundant to TS 3.3(1)a.
(2) The proposed changes do not create the possibility of a new or
different kind of accident from any previously analyzed.
There will be no physical alterations to the plant configuration,
changes to setpoint values, or changes to the implementation of
setpoints or limits as a result of this proposed change. Valves that
are required to be repositioned during an accident to mitigate the
consequences will still be tested on a refueling frequency. The
proposed change only deletes unnecessary or redundant testing
requirements from the TS. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from any
previously analyzed.
(3) The proposed changes do not involve a significant reduction in
a margin of safety.
The proposed changes delete unnecessary or redundant surveillance
requirements within the TS. The deletion of TS 3.2, Table 3-5 Item 6,
only deletes testing requirements that are already required to be
conducted by TS 3.3(1)a. The deletion of the requirement to test the RW
backup valves to the containment air coolers in TS 3.6(3) only deletes
an unnecessary surveillance. RW direct cooling of the containment air
coolers is not required to maintain containment pressure below the
design limit of 60 psig. Therefore, the proposed changes do not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875
Connecticut Avenue, N.W., Washington, D.C. 20009-5728.
NRC Project Director: Theodore R. Quay.

Pennsylvania Power and Light Company

Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania.
Date of amendment request September 26, 1994.
Description of amendment request: The amendment would remove the
requirement for operability of the Average Power Range Monitors (APRMs)
while the plant is in Operational Condition 5. However, the requirement
for the APRMs to be operable during a shutdown margin demonstration,
when the mode switch is in Startup, will remain unchanged.
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
construction, which is presented below:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Not requiring APRMs to be OPERABLE in OPCON 5 will not increase the
probability of inadvertent reactor critically during refueling
operations. Refueling Interlocks, NMS [Neutron Monitoring System] (SRMs
[Source Range Monitor], IRMs [Intermediate Range Monitor]), and
procedural restrictions provide assurance that inadvertent criticality
does not occur due to the simultaneous withdrawal or removal of two
control rods or due to the inadvertent insertion of a fuel bundle into
a core location with a control blade removed.
The FSAR [Final Safety Analysis Report] Section 15.4.1 discusses
the potential for a control rod withdrawal error during refueling and
start-up operations. The discussion concludes that the withdrawal of
one control rod does not require a safety action because the total
worth of one control rod is not sufficient to cause criticality. The
attempted withdrawal of two control rods, assuming an operator error
and a single active failure, would result in a control rod block
initiated by the Refueling Interlocks. The safety-related IRM
subsystem, which is required by Technical Specifications to be OPERABLE
while in OPCON 5, is designed to generate a rod block or reactor scram
on high neutron flux and is therefore a backup protective system for
the Refueling Interlocks during refueling.
The Safety-related IRM subsystem of the NMS is required by
Technical Specifications to be OPERABLE during OPCON 5 to support the
safety design bases of the NMS and RPS [Reactor Protection System]. The
SRM is not a safety-related subsystem but is important to plant safety
and is required by Technical Specifications to be OPERABLE in OPCON 5.
The SRM subsystem provides the plant operator with neutron flux levels
from startup conditions to the IRM operating range. The SRMs and IRMs
are designed to respond to local core conditions and would indicate and
respond (control rod block or scram) to an accident condition to
mitigate the transient. Thus, the APRMS are not necessary to be
OPERATOR in OPCON 5. The proposed Technical Specification change will
not alter the current requirements that the APRMs be OPERABLE during
shutdown margin demonstrations in OPCON 5 when the mode switch is in
Startup.
The proposed Technical Specification change would reduce the APRM
operability requirement in OPCON 5 and would not affect the FSAR
evaluation of the inadvertent criticality due to the withdrawal or
removal of the highest worth control rod or due to the insertion of
fuel bundles in uncontrolled cells. The FSAR concludes that the
Refueling Interlocks and plant procedures provide assurance that
inadvertent criticality does not occur during refueling.
The consequences of an accident will not be increased by the
proposed Technical Specification change because of the existing lines
of defense which prevent an inadvertent criticality event during
refueling, e.g., administrative restrictions, refueling procedures,
licensed plant operators, SRMs, Refueling Interlocks, and IRMs.
Furthermore, should the number of operator IRM or SRM channels be less
than that required by Technical Specifications, the Technical
Specifications require that core alteration activities be suspended and
all insertable control rods be inserted into the core.
Therefore, the proposed changes do not result in an increase in the
probability or consequences of an accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes to the Technical Specifications will remove
the APRM operability requirement while in OPCON 5 (except for shutdown
margin demonstration testing); however, the SRMs and IRMs will still be
required to be OPERABLE in OPCON 5.
The IRMs are safety-related and are designed to detect and respond
to increases in neutron flux within the local core regions. Any
inadvertent increases in neutron flux during refueling would originate
at a local core location, i.e., rod withdrawal error or fuel bundle
insertion. Technical Specifications require IRM operability and will
generate an RPS scram or control rod block if neutron flux increased to
the setpoint. Therefore, removing the APRMs operability requirement in
OPCON 5 would not effect any safety related equipment or equipment
important to safety.
The APRMs provide core power information to the control room
operator and also provide trip signals to the RMCS [Reactor Manual
Control System] and RPS as required. The absence of an APRMs input
signal will not affect these systems during refueling operations.
Removing the APRMs operability in OPCON 5 will not affect the
response of safety-related equipment as previously evaluated in the
FSAR. The proposed changes to the Technical Specifications do not
affect any safety-related equipment or equipment important to safety.
The proposed changes to the Technical Specifications would remove
the APRMs operability requirement during refueling operations.
Technical Specifications require IRM operability and will generate an
RPS scram or control rod block if neutron flux increased to the
applicable setpoint.
No new types of accidents would be introduced since the SRMs and
IRMs are available and required to be OPERABLE in OPCON 5. Both SRMs
and IRMs would indicate and provide a control rod block or scram
signal, as appropriate, to an increase in neutron flux to mitigate a
transient event. Furthermore, should the number of OPERABLE IRM or SRM
channels be less than that required by Technical Specifications, the
Technical Specifications require that core alteration activities be
suspended and all insertable control rods be inserted into the core.
Therefore, the proposed Technical Specification changes do not
create the possibility of a new or different kind of accident from any
accident previously evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
For the reasons discussed in items 1 and 2 above and because the
Technical Specification Bases do not discuss or require APRMs
operability during OPCON 5, Refueling, the proposed Technical
Specification changes do not involve a significant reduction in a
margin of safety.
The NRS staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration. Local
Public Document Room location: Osterhout Free Library, Reference
Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.

Philadephia Electric Company

Docket Nos. 50-352 and 50-353

Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania.
Date of amendment request: August 22, 1994.
Description of amendment request: The amendment consists of five
(5) sections of Technical Specifications changes which reflect the
Improved Standard Technical Specifications (NUREG-1433):

Section 1: Control Rod Block Instrumentation,
Section 2: Standby Liquid Control System Operability in Mode 5,
Section 3: Scram Discharge Volume Valve Testing,
Section 4: Optional Method of Scram Timing, and
Section 5: Definition of Core Alteration.

Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Section 1: Control Rod Block Instrumentation
1. The proposed Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed TS changes can be divided into two general categories,
the deletion of the ``S/U'' requirements, and the change in frequency
of the SRM [Source Range Monitor] and IRM [Intermediate Range Monitor]
Calibration and Functional Tests. In each case in which the ``S/U''
requirement has been deleted, the normal surveillance frequency
specified for the required Operating Condition remains. The equipment's
associated probability of failure remains unchanged. In the case of the
surveillance frequency changes proposed for the SRMs and IRMs, the
probability of an accident evaluated in the SAR [Safety Analysis
Report] occurring does not increase since there is no credit taken in
the SAR for those Control Rod Block functions with respect to an
accident. As such, the proposed changes will not result in an increase
in the probability of occurrence of an accident previously evaluated in
the SAR. The proposed TS changes do not alter the method of operation
or performance of the equipment in carrying out associated Control Rock
Block functions. Thus, the consequences of an accident previously
evaluated in the SAR are not increased.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed TS changes do not alter the configuration of the plant
or the way that the plant is operated. The equipment can perform no
other function than it is presently capable of, or cause or permit any
other accident than is now possible. Thus, the possibility of an
accident of a different type than previously evaluated in the SAR
cannot be created.
Therefore, the proposed TS changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The proposed TS changes do not involve a significant reduction
in a margin of safety.
Since the proposed TS changes affect only the surveillance
frequency intervals and do not change the plant configuration or
associated instrument setpoints, there is no quantitative or
qualitative reduction in the margin of safety. Thus, the margin of
safety as defined in the bases of any Technical Specification is not
reduced.
Therefore, the proposed TS changes do not involve a reduction in a
margin of safety.
Section 2: Standby Liquid Control System Operability in Mode 5
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed TS change will remove the SLCS operability requirement
in OPCON 5. The purpose of the SLCS is to bring the reactor to and
maintain it in a cold shutdown condition from normal power operations
following failure to scram during power operations. Initiation of the
SLCS is not a precursor to any accident. Therefore, inoperability of
the SLCS in OPCON 5 cannot increase the probability of an accident
previously evaluated.
The proposed TS change does not involve a physical change in any
system's configuration and no new modes of operation are introduced.
The SLCS has not analyzed function OPCON 5. The probability of fuel
failure will not be increased by this change. Shutdown margin, in
conjunction with TS requirements and procedural controls, will assure
that an inadvertent criticality event will not occur during refueling.
In addition, the Reactor Protection System (RPS) and Control Rod System
will provide protection in the unlikely event that an inadvertent
criticality should occur.
Therefore, the proposed TS change does not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed TS changes does not involve a physical change in any
system's configuration and no new modes of operation are introduced.
The SLCS's only purpose is to mitigate the consequences of a failure to
scram during power operation. In OPCON 5, the SLCS has no analyzed
function, therefore, the proposed TS change will not create the
possibility of a new or different kind of accident from any previously
evaluated.
3. The proposed TS change does not involve a significant reduction
in a margin of safety.
The purpose of the SLCS is to bring the reactor to and maintain it
in a cold shutdown condition from normal power operations following a
failure to scram during power operations. The SLCS is not designed to
terminate an inadvertent criticality during OPCON 5. Shutdown margin,
either demonstrated or analytically determined, in conjunction with
Technical Specifications and procedural controls, will assure that an
inadvertent criticality event will not occur during refueling
operations. In addition, the RPS and Control Rod System, which are
extremely reliable, will provide protection in the unlikely event that
an inadvertent criticality does occur. Therefore, the proposed TS
change does not involve a reduction in a margin of safety.
Section 3: Scram Discharge Volume Valve Testing
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of an
accident previously evaluated.
The Scram Discharge Volume (SDV) is not an accident initiator.
Deletion of the requirement that the SDV be determined OPERABLE by
testing the SDV vent and drain valves when control rods are scram
tested from a normal control and configuration of less than or equal to
50% rod density at least once per 24 months, as proposed, will have no
effect on the probability or consequences of an accident previously
evaluated.
This proposed TS will have a negligible impact on the conditions
experienced by the vent and drain valves as they stroke closed, since
the SDV is initially vented to the atmosphere, and the valves close
before the SDV becomes pressurized, even during a scram at full reactor
power. Reactor pressure and Control Rod Drive (CRD) discharge flow
conditions do not influence the SDV vent and drain closure rates, since
the SDV is of sufficient volume and initially vented such that peak
pressure prior to the SDV complete isolation will not be substantial.
In addition, lower coolant temperatures expected during testing at
shutdown conditions will also have a negligible impact on the
performance of the test. Although, there could be some variation in the
performance [of] the SDV vent and drain valves to re-open when
performing the test during shutdown conditions, as opposed to
conducting the test during power operation, the ability of the valves
to re-open is demonstrated after each reactor scram during power
operation.
In the event and SDV vent or drain valve failed to open, increasing
SDV level during reactor operation would cause 1) an alarm in the Main
Control Room (MCR), 2) a control rod block, and finally a reactor scram
initiated by the Reactor Protection System (RPS) if action is not taken
to drain the SDV. Therefore, the ability to shut down the reactor is
not impaired. If a SDV vent or drain valve fails to close, the
redundant valve's closure would provide the required function. If both
valves failed to close, a loss of reactor coolant in the form of water
discharged from the CRD system would occur. The amount of water
discharged will be relatively small, and is more of a concern from the
standpoint of contamination to the Secondary Containment rather than a
loss of reactor water inventory. A structural failure of the SDV, which
bounds this case of an open SDV vent or drain line, has been previously
evaluated in NUREG-0808, ``Generic Safety Evaluation Report Regarding
Integrity of BWR Scram System Piping.'' In this evaluation, the NRC
concluded that, for a bounding leakage case corresponding to a rupture
of the SDV, the offsite doses would be well within the limits of
10CF100, and that adequate core cooling would be maintained.
Deletion of the requirement that the SDV be determined OPERABLE by
testing the SDV vent and drain valves, as proposed in this TS Change
Request, will have an insignificant effect on the probability of
occurrence of malfunction of any plant equipment. The conditions in the
SDV at the time of vent and drain valve closure are not appreciably
different w

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A94-27613. Public record. Not legal advice.