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

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URL: https://www.frixlaw.com/law-library/documents/fr%3A98-1904

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** January 28, 1998
- **Citation:** 63 FR 4308

## 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 January 5, 1998, through January 15, 1998.
The last biweekly notice was published on January 14, 1998 (63 FR
2271).

Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By February 27, 1998, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or

[[Page 4309]]

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 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: December 17, 1997.
Description of amendment request: The requested amendment revises
Technical Specification Section 5.6.5, ``Core Operating Limits Report
(COLR).'' The revisions add reference to an additional approved
methodology for correlating departure from nucleate boiling (DNB)
ratios. The added methodology is the Siemens Power Corporation Topical
Report, EMF-92-153(P)(A), ``HTP: Departure from Nucleate Boiling
Correlation for High Thermal Performance Fuel.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change adds a methodology that has been previously
reviewed and approved by the NRC for determining the DNB safety limit.
The new methodology utilizes the High Thermal Performance (HTP)
correlation developed by the fuel manufacturer, Siemens Power
Corporation. The HTP correlation is empirically based and results in a
DNB safety limit that corresponds to a 95% probability at a 95%
confidence level that DNB will not occur. The DNB ratio safety limit is
a conservative design value which is used as a basis for setting core
safety limits. The DNB correlation is not assumed to be an initiator of
analyzed events or transients, and use of the new DNB correlation will
not alter assumptions relative to mitigation of accident or transient
events. The proposed change has been confirmed to ensure that no
previously evaluated accident or transient results in a DNB less than
the DNB correlation safety limit. The HTP DNB correlation assures with
high confidence that, for accidents and transients that do not result
in a DNBR less than the HTP DNBR safety limit, departure from nucleate
boiling and subsequent fuel overheat will not occur in HTP fuel.
Therefore, the proposed change does not involve any increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures, or components or changes in parameters
governing normal plant operation. The proposed change

[[Page 4310]]

will allow use of the new DNB correlation in like manner as the
existing DNB correlation in the analysis of accidents and transients to
assure that the acceptance criteria for current analyses are met.
Therefore, the proposed change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin of
safety?
The proposed change allows use of a DNB correlation that determines
a safety limit that is slightly lower than the currently used DNB
correlation. While the slightly lower DNB correlation safety limit
allows a small increase in margin in analyzing accidents and
transients, the change from the existing DNB correlation to the
proposed DNB correlation is not directly comparable to the margin of
safety. This is because the margin of safety for a particular accident
or transient is that margin that results from the difference between
the DNBR calculated for the particular accident or transient using the
DNB correlation and the DNBR safety limit determined by the DNB
correlation. Since both the safety limit and the accident or transient
calculated DNB use the same DNB correlation, the margin of safety is
consistently calculated and evaluated for acceptability. Since both the
current and proposed DNB correlation closely approximate test data, and
they still meet the 95/95 criterion, and the new DNB correlation does
not result in a DNBR from an accident or transient less than the DNBR
correlation safety limit, the proposed change does not result in a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Gordon E. Edison, Acting.

Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of amendment request: June 6, 1997, as supplemented September
25, 1997.
Description of amendment request: The proposed amendment would
delete the requirement to sample the spray additive tank per Technical
Specification (TS) Table 4.1-2, ``Frequency for Sampling Tests,'' and
delete the sodium hydroxide (NaOH) reference in TS Section 5.2.C.1. The
request to delete the requirement and the reference was inadvertently
omitted as part of the licensee's original submittal dated August 22,
1996, supplemented March 28, 1997, to eliminate the requirement for the
NaOH containment spray additive and spray additive tank.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident previously
evaluated?
Response:
The request to remove the requirement for the spray additive tank
was approved as part of Amendment No. 191 to Operating License No. DPR
26. By letter dated April 23, 1997, the Commission reviewed and
approved the amendment request. However, Consolidated Edison failed to
include the deletion of the requirement to sample the spray additive
tank. The removal of the requirement for the spray additive tank has
been analyzed and approved; therefore, there is no further basis for
continued testing of the tank. Further, the deletion of the requirement
would not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Does the proposed license amendment create the possibility of a
new or different kind of accident from any previously evaluated?
Response:
The proposed changes allow the containment safeguards to mitigate
the consequences of a design basis LOCA [loss-of-coolant accident] in a
manner equivalent to that previously approved. Therefore, the proposed
changes do not create an accident or malfunction of safety equipment of
a different type.
(3) Does the proposed amendment involve a significant reduction in
margin of safety?
Response:
With the proposed changes, all of the safety criteria previously
evaluated are still valid and remain conservative. Therefore, the
proposed amendment does not involve a significant reduction in the
margin of safety.
Based on this review, it appears that the three standards of
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: S. Singh Bajwa, Director.

Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: December 17, 1997.
Description of amendment request: The proposed amendments would
revise Section 6.9.1.9 of the Technical Specifications (TS) to
reference updated or recently approved topical reports, which contain
methodologies used to calculate cycle-specific limits contained in the
Core Operating Limits Report. These topical reports have all been
previously approved by the staff under licensing actions separate from
the current amendment request.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below.
1. Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes do not involve any modification to
existing systems, components, operating limits, or operating procedure.
Therefore, these proposed changes will have no impact on the
consequences or probabilities of any previously evaluated accidents.
2. Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?

[[Page 4311]]

No. No actual plant equipment or operating procedure will be
affected by the proposed changes. Hence, no new equipment failure modes
or accidents from those previously evaluated will be created.
3. Will the change involve a significant reduction in a margin of
safety?
No. Margin of safety is associated with confidence in the design
and operation of the plant. The proposed changes to the TS do not
involve any change to plant design or operation. Thus, the margin of
safety previously analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28242-0001.
NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: December 17, 1997.
Description of amendment request: The proposed amendments would
revise Section 6.9.1.9 of the Technical Specifications (TS) to
reference updated or recently approved topical reports, which contain
methodologies used to calculate cycle-specific limits contained in the
Core Operating Limits Report. These topical reports have all been
previously approved by the staff under licensing actions separate from
the current amendment request.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below.
1. Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes do not involve any modification to
existing systems, components, operating limits, or operating procedure.
Therefore, these proposed changes will have no impact on the
consequences or probabilities of any previously evaluated accidents.
2. Will the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. No actual plant equipment or operating procedure will be
affected by the proposed changes. Hence, no new equipment failure modes
or accidents from those previously evaluated will be created.
3. Will the change involve a significant reduction in a margin of
safety?
No. Margin of safety is associated with confidence in the design
and operation of the plant. The proposed changes to the TS do not
involve any change to plant design or operation. Thus, the margin of
safety previously analyzed and evaluated is maintained.
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: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, North Carolina.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas

Date of amendment request: September 23, 1997.
Description of amendment request: The proposed amendment changes
the Reactor Protective System and Engineering Safety Actuation System
trip set point and allowable values for steam generator low pressure.
The proposed amendment also relocates the RPS and ESFAS response time
tables from the Technical Specifications to the Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does Not Involve a Significant Increase in the Probability or
Consequences of and Accident Previously Evaluated.
The proposed changes included in this amendment request do not
affect the accident initiators in any of the accidents previously
evaluated. The proposed trip setpoints and allowable values for Steam
Generator Pressure--Low are being reduced by this proposed amendment
request. This change is necessary to increase the operating margin
between the full power steam generator pressure and these setpoints.
The change should reduce the probability of an inadvertent Main Steam
Isolation Signal (MSIS) from occurring at power since it will increase
the operating space between the operating pressure and the setpoints.
Therefore, this amendment request will not increase the probability of
any accident previously evaluated.
The secondary system pipe break safety analyses were reanalyzed for
the Steam Generator Pressure--Low setpoint reduction effort. This
effort included the removal of unnecessary analysis conservatisms
resulting in a significant reduction in the associated setpoints. The
proposed changes do not involve any change to the configuration or
method of operation of any plant equipment used to mitigate the
consequences of an accident. The previously evaluated accidents which
were determined to be impacted by this setpoint change were evaluated
with no significant increase in the consequences.
This amendment request contains the relocation of the Reactor
Protective System (RPS) and Engineered Safety Features Actuation System
(ESFAS) response time information from the Technical Specifications
(TS) to the Safety Analysis Report. This proposed change adopts the TS
``line-item improvement'' as recommended in NRC Generic Letter 93-08,
``Relocation of Technical Specification Tables of Instrument Response
Time Limits,'' dated December 29, 1993. The NRC has concluded that 10
CFR 50.36 does not require the response time tables to be retained in
TSs and has issued Generic Letter 93-08 as a line item improvement to
allow their removal. Response time testing will still be required by
the ANO-2 TS after the relocation of the associated response time
information in this amendment request. Relocating the response time
information for the RPS and ESFAS from the TS to the SAR will not alter
these surveillance requirements. Therefore, the relocated response time
portion of this amendment request is considered administrative in
nature and will not affect the probability or consequences of any
accident previously evaluated.
Therefore, this change does not involve a significant increase in
the

[[Page 4312]]

probability or consequences of any accident previously evaluated.
2. Does Not Create the Possibility of a New or Different Kind of
Accident from and Previously Evaluated.
The proposed changes do not involve any physical modifications
(i.e., new systems, new components, etc.) to the plant. The proposed
changes do not involve any change to the configuration or method of
operation of any plant equipment used to mitigate the consequences of
an accident. The results of the accident reanalyzes suggest no
different phenomena or plant behavior than previously considered. The
Steam Generator Pressure Low setpoint change does not create any new or
different system actuations or interactions than evaluated previously.
The relocated response time portion of this amendment request is
considered administrative in nature and is not considered an accident
initiator. Therefore, this change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does Not Involve a Significant Reduction in the Margin of
Safety.
The accidents which were determined to be impacted by the Steam
Generator Pressure Low setpoint change were evaluated to ensure
acceptable results are maintained. The instrument error calculations
supporting the lower Steam Generator Pressure Low setpoint and
allowable values will ensure the present accident analysis assumptions
are still maintained. The methodology used to determine the instrument
loop errors and uncertainties is the same as that used in previous
amendment requests that have been reviewed and approved by the NRC.
Based on these evaluations, the proposed changes do not involve a
significant reduction in a margin of safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the reauested chance does not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
NRC Project Director: John Hannon.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas

Date of amendment request: September 23, 1997.
Description of amendment request: The proposed amendment reduces
the minimum primary system flow that is specified in the technical
specifications to reflect the effects of increased primary system
resistance caused by steam generator tube plugging.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does Not Involve a Significant Increase in the Probability or
Consequences of an Accident Previously Evaluated.
Entergy Operations is proposing a change to the Technical
Specifications for Arkansas Nuclear One--Unit 2 (ANO-2) to accommodate
a larger number of plugged steam generator tubes. The proposed
amendment request will revise the Technical Specifications to
conservatively account for the reduced reactor coolant system (RCS)
flow effects of plugging up to 30 percent of the tubes in either steam
generator. This change will reduce the minimum RCS total flow rate from
120.4 x 106 lbm/hr to 108.4 x 106 lbm/hr until
the steam generators are replaced. The steam generators are currently
scheduled for replacement during the fall of the year 2000. After the
steam generators are replaced, the minimum RCS flow will then return to
the current value of 120.4 x 106 lbm/hr.
The tube plugs that are installed in the steam generators are
passive components by nature. This amendment request does not change
the type of plugs which may be installed in the steam generators nor
does it change the criteria for plugging steam generator tubes.
Reducing the minimum required RCS flow does not change the plant's
required mode of operation or modify any active component. Therefore,
this amendment request will not significantly increase the probability
of the occurrence of a previously evaluated accident.
The installation of steam generator tube plugs removes the affected
tube from service thus reducing the heat transfer surface area and
increasing the steam generator primary side flow resistance. The
increased flow resistance in the affected steam generator leads to a
reduction in the RCS flow available for core cooling. The reduced RCS
flow rate and heat transfer surface area resulted in a change in
several primary and secondary parameters that required reanalysis. The
ANO-2 accident reanalyses supporting the additional steam generator
tube plugging and the reduction in RCS flow have been completed.
The Design Basis Accidents (DBAs) affected by these changes were
reanalyzed to determine if the effects of increased steam generator
tube plugging and the reduced RCS flow could result in exceeding the
acceptance criteria applicable to each of these events. It was
determined that the DBA acceptance criteria would not be exceeded as a
result of increased steam generator tube plugging and reduction in the
minimum RCS flow rate.
Based on the results of the analysis, it is concluded that the
emergency core cooling system design satisfies the acceptance criteria
of 10 CFR 50.46(b) for a spectrum of small break and large break loss
of coolant accidents (LOCAs). The specified acceptable fuel design
limits (SAFDLs) and the RCS pressure boundary limits also are not
violated. The fuel and core performance were also determined to remain
within acceptable limits. Primary and secondary system pressures remain
below their respective pressure limits.
Analyses and evaluations of the DBAs have been performed
demonstrating that the NRC acceptance criteria for these events are
met. The revised analyses and evaluations consider reduced RCS flow,
increased RCS temperatures, and increased steam generator tube plugging
conditions. Although the offsite dose during a steam generator tube
rupture event could increase, the results remain well within 10 CFR
[Part] 100 limits. Therefore, the consequences of a previously
evaluated accident are not significantly increased.
Therefore, this change does not involve a significant increase in
the probability or consequences of any accident previously evaluated.
2. Does Not Create the Possibility of a New or Different Kind of
Accident from any Previously Evaluated.
The proposed amendment reduces the minimum RCS total flow to
account for the effects of steam generator tube plugging. This
amendment request will not change the modes of operation defined in the
Technical Specifications. This change does not add any new equipment,
modify any interfaces with any existing equipment, change the
equipment's function, or the method of operating the equipment. The
proposed change does not change plant conditions in a manner which
could

[[Page 4313]]

affect other plant components. Reactor core, RCS, and steam generator
parameters remain within appropriate design limits during normal
operation. The proposed change could not cause any existing equipment
to become an accident initiator. Therefore, this change does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Does Not Involve a Significant Reduction in the Margin of
Safety.
The margins of safety associated with this change are defined in
the fuel and core related analyses, and in each of the transient and
accident analyses affected by the reduced RCS flow. An evaluation of
the affected analyses confirmed that the established acceptance
criteria for specified acceptable fuel design limits, primary and
secondary system over-pressurization, and the acceptance criteria for
the emergency core cooling systems have been satisfied by this license
amendment request. The evaluation concludes that, when considering the
proposed Limiting Conditions for Operation for the minimum RCS total
flow rate, all applicable acceptance criteria limits are met. The
margins of safety associated with the transient and accident analyses
affected by this change will not be significantly reduced. Therefore,
this change does not involve a significant reduction in the margin of
safety.
Therefore, based upon the reasoning presented above and the
previous discussion of the amendment request, Entergy Operations has
determined that the requested change does not involve a significant
hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
NRC Project Director: John Hannon.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: December 10, 1997.
Description of amendment request: To clarify certain sections of
the Technical Specifications (TSs) and Bases which have been
demonstrated to be unclear or conflicting. Administrative changes
include TS 2.3 Bases, Table 3.1.1.G.1, Table 3.1.1.M.2, Section 4.3.C,
and Section 6.1.1. Technical changes include Table 3.3.3, note b,
Section 3.4 Bases, Section 3.8 Bases and Section 4.5 Bases.
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:
With respect to the administrative changes, they are typical of the
example I.c.2.e.i in 51 FR 7744 and therefore, they do 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 the margin of safety; in that
they are purely administrative changes to achieve consistency or
correct an error in the TS.
With respect to technical change, Table 3.1.1, note b:
1. Involve a siginificant increase in the probability or
consequences of an accident previously evaluated; (or)
The proposed change would restore the original value of less than
600 psig. This lower value would not increase the probability of any
accident as it provides a more conservative level below which
protection can be bypassed.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated; (or)
The proposed change would restore the original value of less than
600 psig. The setpoint of a bypass cannot create a different kind of
accident, it can only affect the severity.
3. Involve a significant reduction in a margin of safety; As the
requested change lowers the bypass setpoint, the margin of safety will
be increased.
With respect to Section 3.4 Bases:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated; (or)
The proposed change to the Bases removes a possible area of
confusion from the [TS], and updates the Bases to reflect the results
of newer, approved methodologies. Therefore, no change to any
probability calculation occurs.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated; (or)
The proposed change addresses an existing accident (Small Break
LOCA) and removes outdated and possibly confusing information.
Therefore, no new or different kind of accident is created.
3. Involve a significant reduction in a margin of safety;
The proposed change does not change the way the plant is operated
or the way design Bases are maintained. It only removes an outdated and
possibly confusing paragraph from the Bases, therefore, no margin of
safety is affected.
With respect to Section 3.8 Bases:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated; (or)
The Isolation Condenser Radiation Monitors had no impact o[n] the
operation of any plant system. Additionally, the monitors were not
relied upon for any post accident evaluations. They were removed from
the plant using the 10 CFR 50.59 process. As this request updates the
[TS] Bases to reflect the plant as currently configured, no impact on
the probability or consequences of any previously evaluated accident is
possible.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated; (or)
The Isolation Condenser Radiation Monitors had no impact o[n] the
operation of any plant system. Additionally, the monitors were not
relied upon for any post accident evaluations. They were removed from
the plant using the 10 CFR 50.59 process. As this request updates the
[TS] Bases to reflect the plant as currently configured, no new or
different kind of accident is created.
3. Involve a significant reduction in the margin of safety;
The Isolation Condenser Radiation Monitors had no impact o[n] the
operation of any plant system. Additionally, the monitors were not
relied upon for any post accident evaluations. They were removed from
the plant using the 10 CFR 50.59 process. As this request updates the
[TS] Bases to reflect the plant as currently configured, no reduction
in any margin of safety can occur.
With respect to Section 4.5 Bases:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated; (or)
No change to any procedure, nor any modification to any system is
requested. The same surveillance will be performed at the same
frequency. Only the brand of chemical used to perform the surveillance
will be affected. As an equivalent chemical will be selected, no
increase in the probability or consequences of an accident previously
evaluated can be created.

[[Page 4314]]

2. Create the possibility of a new or different kind of accident
from any accident previously evaluated; (or)
No change to any procedure, nor any modification to any system is
requested. The same surveillance will be performed at the same
frequency. Only the brand of chemical used to perform the surveillance
will be affected. As an equivalent chemical will be selected, no new or
different kind of accident previously evaluated can be created.
3. Involve a significant reduction in the margin of safety;
No change to any procedure, nor any modification to any system is
requested. The same surveillance will be performed at the same
frequency. Only the brand of chemical used to perform the surveillance
will be affected. As an equivalent chemical will be selected, no margin
of safety can be affected.
The 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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ronald B. Eaton.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa

Date of amendment requests: October 3, 1997.
Description of amendment requests: The proposed amendment would
revise the Operating License to allow the start of core offload as soon
as 60 hours after shutdown instead of the 120 hours currently
specified.
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 Operating License Amendment will not significantly
increase the probability or consequences of any previously evaluated
accidents.
The proposed change will allow initiation of core offload earlier
after shutdown than is currently allowed. Thermal-hydraulic analysis
shows that maximum bulk SFP, local water, and fuel clad temperatures
will remain within acceptable limits and, in fact, do not exceed those
previously reviewed and approved for Amendment 195.
Thermal-hydraulic analysis shows the minimum time to action is
calculated at 4.5 hours versus 5.5 hours previously reviewed and
approved for Amendment 195. In the event of a loss of forced cooling
with cask pit isolation gate failure event, the DAEC will use Emergency
Service Water (ESW), a Seismic Category I system, to provide makeup to
the SFP. It is estimated to take no more than 2 hours to provide ESW
makeup to the SFP, therefore the minimum time to action of 4.5 hours is
sufficient time to prevent uncovering the fuel in the SFP.
The DAEC design basis refueling accident, as discussed in Section
15.10.2 of the Updated Final Safety Analysis Report, assumes a twenty-
four hour decay time before core offload begins. The proposed change
does not adversely affect that accident analysis.
Therefore, the proposed change will not result in an increase in
probability or consequences of an accident previously evaluated.
2. The proposed changes will not create a new or different kind of
accident from those previously evaluated.
Thermal-hydraulic analysis shows that the proposed change will not
result in maximum bulk SFP, local water, or fuel clad temperatures
which would initiate bulk pool boiling, challenge fuel rod integrity or
jeopardize the structural integrity of the pool.
As stated above, the minimum time to action of 4.5 hours allows
sufficient time to provide ESW makeup to the SFP. Therefore, this
change does not create the possibility of a new or different type of
accident.
3. The proposed change will not result in a significant reduction
in any margin of safety.
This change will not result in maximum bulk SFP, local water, and
fuel clad temperatures in excess of those previously evaluated and
accepted per Amendment 195. The thermal-hydraulic analysis for Case C
does show a reduction in the minimum time to action by one hour.
However, 4.5 hours does provide sufficient time to provide ESW makeup
to the SFP as this task is estimated to require no more than 2 hours.
Furthermore, this change does not result in any change to the Technical
Specifications. Therefore, this change does not result in a significant
reduction in a margin of safety.
Based upon the above, we have determined that the proposed
amendment will not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, Iowa 52401.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: December 15, 1997.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) 2.1 and 3/4.4.1 to change the
safety limit minimum critical power ratio (MCPR) for the upcoming fuel
operating cycle (Cycle 7) from 1.07 to 1.09 for two recirculation loop
operation and from 1.08 to 1.10 for single loop operation. An obsolete
footnote in TS 3/4.4.1, which states that ``the MCPR Safety Limit of
1.07 will be used through the first operating cycle,'' would be
deleted. The associated Bases 2.1 would be changed to (1) reflect the
new MCPR values, (2) delete certain details (including Bases Table
B2.1.2-1, ``Uncertainties Used in the Determination of the Fuel
Cladding Safety Limit,'' and Bases Table B2.1.2-2, ``Nominal Values of
Parameters Used in the Statistical Analysis of Fuel Cladding Integrity
Safety Limit,'') and (3) substitute for the deleted detail a reference
to General Electric Standard Application for Reactor Fuel (GESTAR II),
NEDE-24011-P-A, and to the cycle-specific analysis. The TS Index would
be changed to reflect deletion of Bases Tables B2.1.2-1 and B2.1.2-2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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 derivation of the revised Safety Limit MCPR was performed using
the NRC approved methodology in GESTAR II. The Safety Limit MCPR is a
TS numerical value that cannot initiate an event. Maintaining
compliance with this

[[Page 4315]]

limit will assure that 99.9 percent of the fuel rods will not
experience transition boiling during transient events. The deletion of
the footnote that is no longer necessary and the revision to the Bases
information are administrative only. The proposed change does not
modify any of the accident initiators described in the USAR [Updated
Safety Analysis Report]. No equipment malfunctions or procedural errors
are created as a result of this change, therefore, no accidents are
affected by it. The change does not adversely impact the integrity of
the fuel cladding, which is the first barrier to the release of
radioactivity to the environment. The change does not affect the
operation of any systems necessary to mitigate the radiological
consequences of an accident or to safely shutdown the plant. Therefore,
this change will not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2. 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.
The Safety Limit MCPR is a TS numerical value designed to prevent
fuel damage from transition boiling. It cannot create the possibility
of a transient or accident. The deletion of the footnote that is no
longer necessary and the revision to the Bases information are
administrative only. The proposed change does not directly impact the
operation of any systems or equipment important to safety. The analyses
show that all fuel licensing acceptance criteria are met. The fuel
cladding, reactor vessel, and reactor coolant system integrity will be
maintained. Therefore, this change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. 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 Safety Limit MCPR calculation was performed using the NRC
approved methodology in GESTAR II. Analyses of limiting USAR transients
establish Operating Limit MCPR values that ensure that the Safety Limit
MCPR is not violated. The revised cycle specific Safety Limit MCPR
preserves the existing margin of safety and will continue to assure
that 99.9 percent of the fuel rods will not experience transition
boiling during transient events. The deletion of the footnote that is
no longer necessary and the revision to the Bases information are
administrative only. Thus, the margin of safety to fuel cladding
failure due to insufficient cladding heat transfer during transient
events is not reduced. Therefore, this 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: S. Singh Bajwa.

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: November 13, 1997.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) to (1) modify the low
temperature overpressure protection (LTOP) requirements; (2) modify the
reactor coolant system (RCS) heatup and cooldown limits; and (3) make
changes to correct various items based on the licensee's review of the
current TSs. The supporting TS Bases sections would also be changed to
reflect the proposed TS changes.
The affected TSs are: TS 3.1.2.1, ``Flow Paths--Shutdown;'' TS
3.1.2.2, ``Flow Paths--Operating;'' TS 3.1.2.3, ``Charging Pump--
Shutdown;'' TS 3.1.2.4, ``Charging Pumps--Operating;'' TS 3.1.2.5,
``Boric Acid Pumps--Shutdown;'' TS 3.1.2.6, ``Boric Acid Pumps--
Operating;'' TS 3.1.2.8, ``Borated Water Sources--Operating;'' TS
3.4.1.3, ``Coolant Loops and Coolant Circulation--Shutdown;'' TS 3.4.3,
``Relief Valves;'' TS 3.4.9.1, ``Reactor Coolant System;'' TS 3.4.9.2,
``Pressurizer;'' TS 3.4.9.3, ``Overpressure Protection Systems;'' TS
3.5.3, ``ECCS Subsystems--Tavg 2.2 in2 has been
established. This curve and rate may also be used when RCS cold leg
temperature is below 230 deg.F to demonstrate compliance with Appendix
G limits when unanticipated temperature excursions occur.
The current action statements of Technical Specification 3.4.9.1
will be separated by Mode and will be modified. Similar changes will be
made to the action statements of Technical Specification 3.4.9.2. A
time limit of 72 hours will be placed on the performance of the
engineering evaluation. If this evaluation is not performed in this
time period, or the evaluation does not allow continued operation, the
plant will be required to enter Mode 5 ([less than or equal to] 200
deg.F), instead of the current requirement to be q;''
(2) correct the wording in TS 4.1.1.1.2 ``Reactivity Control Systems
Shutdown Margin--Tavg [less than or equal to] 200 deg.F;''
(3) correct the mode applicability from Mode 3 to Modes 1 and 2 in TS
3.1.3.4 ``Reactivity Control Systems--Rod Drop Time;'' (4) correct the
terminology used to refer to the power dependent insertion limit alarm
in TS 4.1.3.6 ``Reactivity Control Systems--Regulating CEA [Control
Element Assembly] Insertion Limits;'' (5) add a footnote for Mode 4
operability requirement clarification to TS 3.5.3 ``Emergency Core
Cooling Systems, ECCS Subsystems--Tavg q) used to verify compliance
with Technical Specification 3.2.4 is associated with the quadrant of
highest power production with respect to the average of the four
quadrants, instead of the quadrant that deviates the most (increases or
decreases) from the average of the four quadrants. This is consistent
with the method by which power distribution factors are calculated and
applied in the accident analysis and how the Core Power Distribution
Monitoring System calculates Tq. The proposed change will
not alter the way Tq is calculated by the Core Power
Distribution Monitoring System, nor will it alter any of the power
distribution assumptions used in the accident analysis. Therefore, this
change will not significantly increase the probability or consequences
of an accident previously evaluated.
Surveillance Requirement (SR) 4.1.1.1.2 requires that the
difference between predicted and measured core reactivity values be
maintained within [plus or minus] 1.0% [delta]k/k, and that an
adjustment be made between the measured and predicted core reactivity
conditions prior to exceeding 60 EFPD [effective full power days]
following a refueling outage. The proposed change will not affect the
requirement to maintain predicted and measured core reactivity values
within [plus or minus] 1.0% [delta]k/k. However, it will no longer be
necessary to make an adjustment prior to exceeding 60 EFPD provided the
[plus or minus] 1.0% [delta]k/k requirement is met. Historically, this
difference has been small at Millstone Unit No. 2 (less than
approximately [plus or minus] 1.0% [delta]k/k) and an adjustment has
not been necessary to ensure the [plus or minus] 1.0% [delta]k/k
requirement is met. The fact that no adjustment (normalization) will be
necessary when reactivity differences are small will not affect the
ability to identify reactivity anomalies. Therefore, this change will
not significantly increase the probability or consequences of an
accident previously evaluated.
The proposed change to Technical Specification 3.1.3.4 will change
the applicability from Mode 3 to Modes 1 and 2. This is necessary to
allow performance of SR 4.1.3.4 at the conditions in the accident
analysis, and also specified in the [Limiting] Condition [for]
Operation (LCO). CEA [Control Element Assembly] drop time is important
for the mitigation of accidents that are initiated while the reactor is
critical. To ensure the CEA drop time assumed in the accident analysis
is valid, it is necessary to verify CEA drop time with plant conditions
consistent with those expected when the reactor is critical. This
proposed change will allow this verification, and thereby ensure the
CEAs will function as designed to mitigate design basis accidents.
Therefore, this change will not significantly increase the probability
or consequences of an accident previously evaluated.
The proposed change to SR 4.1.3.6 will modify the terminology used
to refer to the power dependent insertion limit (PDIL) alarm to agree
with plant terminology. This change will not alter equipment operation
or any technical aspect of the SR. The information added to the Bases
will specify what equipment provides the PDIL alarm. These changes will
eliminate any confusion with alarm terminology. Therefore, this change
will not significantly increase the probability or consequences of an
accident previously evaluated.
Technical Specification 3.5.3 requires an operable flowpath capable
of taking a suction from the refueling water storage tank (RWST) on a
safety injection actuation signal (SIAS), and automatically
transferring suction to the containment sump on a sump recirculation
actuation signal (SRAS) in Mode 4. In Mode 4, the automatic SIAS
generated by low pressurizer pressure and high containment pressure,
and the automatic SRAS generated by low RWST level, are not required to
be operable. Automatic actuation in Mode 4 is not required because
adequate time is available for plant operators to evaluate plant
conditions and respond by manually operating engineered safety

[[Page 4320]]

features components. Since the manual actuation (trip pushbuttons)
portions of the safety injection and sump recirculation actuation
signal generation are required to be operable in Mode 4, credit can be
taken for remote manual operation to generate the SIAS and SRAS which
will position all components to the required accident position. The
proposed change to Technical Specification 3.5.3 will add a footnote
(***) to explain how these requirements are met in Mode 4. This change
will not reduce operability or surveillance requirements for the
Emergency Core Cooling System (ECCS) subsystem required to be operable
by Technical Specification 3.5.3. The ECCS will continue to function as
designed to mitigate design basis accidents. Therefore, this change
will not significantly increase the probability or consequences of an
accident previously evaluated.
The proposed change to Technical Specification 3.6.3.2 will revise
the wording of the LCO and SR by changing ``locked closed'' to ``sealed
closed,'' and deleting the requirement to be electrically deactivated.
The action statement will also be revised to reflect these proposed
changes. These changes will not affect the requirement for the
containment purge valves to be closed in Modes 1 through 4. Therefore,
the proposed changes will not significantly increase the probability or
consequences of an accident previously evaluated.
The proposed change to SR 4.6.1.7 will change the surveillance
frequency from ``prior to each reactor startup'' to ``at least once per
31 days.'' This change, which will require the surveillance to be
performed more often (assuming a normal plant startup sequence) will
provide additional assurance that the containment purge valves are
sealed closed. In addition, this change will ensure consistency between
the SR and the applicability of this specification, and also with the
requirements to verify containment integrity in accordance with
Technical Specification 3.6.1.1. Therefore, the proposed change will
not significantly increase the probability or consequences of an
accident previously evaluated.
The change in numbering of SR 4.6.1.7 to SR 4.6.3.2 is an
administrative change only. It will not affect any technical aspect of
the SR. Therefore, the proposed change will not significantly increase
the probability or consequences of an accident previously evaluated.
The proposed changes to Technical Specifications 3.8.2.1 and
3.8.2.1A will modify the nomenclature used to refer to the vital A.C.
buses to be consistent with the terminology used by Operations
Department personnel and the nomenclature contained in their
procedures. These changes will not alter equipment operation or any
technical aspects of these specifications. These proposed changes are
administrative changes only. The A.C. buses will continue to function
as designed to mitigate design basis accidents. Therefore, these
changes will not significantly increase the probability or consequences
of an accident previously evaluated.
The proposed changes to Technical Specifications 5.1.3, 5.2.3,
5.3.2, and 5.7.1 will remove the word ``original.'' Reference to
original design is not appropriate since these items can be changed by
approved processes. However, these changes will still require the items
addressed by these specifications to be designed and maintained in
accordance with the Final Safety Analysis Report (FSAR). The proposed
changes have no affect on the current approved plant design. Therefore,
these changes will not significantly increase the probability or
consequences of an accident previously evaluated.
Technical Specification 5.9 will be deleted. The required
provisions for shoreline protection have been completed, and this
Technical Specification is no longer necessary. The removal of this
outdated specification will not impact any current requirements.
Therefore, this change will not significantly increase the probability
or consequences of an accident previously evaluated.
The proposed changes do not alter how any structure, system, or
component functions. There will be no effect on equipment important to
safety. The proposed changes have no effect on any of the design basis
accidents previously evaluated. Therefore, this License Amendment
Request does not impact the probability of an accident previously
evaluated, nor does it involve a significant increase in the
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes do not alter the plant configuration (no new
or different type of equipment will be installed) or require any new or
unusual operator actions. They do not alter the way any structure,
system, or component functions and do not alter the manner in which the
plant is operated. The proposed changes do not introduce any new
failure modes. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change to the definition of Tq will make
the Technical Specification definition consistent with the approved
calculation methodology. This will ensure the core power distribution
is consistent with accident analysis assumptions. The proposed change
to the wording of SR 4.1.1.1.2 will not affect the acceptance criteria
of [plus or minus] 1.0% [delta]k/k, which ensures the accident analysis
accurately reflects core reactivity conditions. The proposed change in
the applicability of Technical Specification 3.1.3.4 will allow
verification of CEA drop time at plant conditions assumed in the
accident analysis. This will ensure the CEAs will function as assumed.
The proposed change to SR 4.1.3.6 will modify the terminology used to
refer to the PDIL alarm to agree with plant terminology. This change
will not alter equipment operation or any technical aspect of the SR.
Adding the footnote to Technical Specification 3.5.3 will not change
any technical aspects of this specification. One ECCS subsystem will be
available for accident mitigation. The proposed change in wording of
Technical Specification 3.6.3.2 will not affect the requirement for the
containment purge valves to be closed in Modes 1 through 4. The
proposed change in the frequency of performance for SR 4.6.1.7 will
provide greater assurance that the containment purge valves are closed
to prevent the potential release of radioactive material through these
penetrations during accident conditions. The proposed changes in
terminology in Technical Specifications 3.8.2.1 and 3.8.2.1A will not
change any technical requirements for the equipment covered. The
equipment will still function as assumed. Modifying the Bases of
Technical Specifications are necessary to be consistent with the
proposed changes will not change any requirements of these
specifications. The modification to Technical Specifications 5.1.3,
5.2.3, 5.3.2, and 5.7.1 will not affect the requirement to maintain
these items in accordance with requirements contained in the FSAR.
Deleting Technical Specification 5.9 will not affect any requirements
since the requirements contained in this specification have already
been completed.
The proposed changes do not affect any of the assumptions used in
the accident analysis, nor do they affect any operability requirements
for equipment

[[Page 4321]]

important to plant safety. Therefore, these proposed changes will not
result in a significant reduction in the margin of safety as defined in
the Bases for Technical Specifications covered in this License
Amendment Request.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: Phillip F. McKee.

Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: December 12, 1997.
Description of amendment request: The proposed amendment would
revise the facility Technical Specifications (TSs) regarding normal
working hours of plant staff to provide for shift duration of 12 hours.
It would also revise the TSs to maintain existing ``once per shift''
surveillance requirements at 8-hour intervals.
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:
Does the proposed licensing amendment involve a significant
increase in the probability or consequences of an accident previously
evaluated?
Response:
Establishing operating personnel work hours at ``a normal 8 to 12
hour day, nominal 40-hour week'' allows normal plant operations to be
managed more effectively and does not adversely affect performance of
operating personnel. Overtime remains controlled by site administrative
procedures in accordance with NRC Policy Statement on working hours
(Generic Letter 82-12). If 8 hour shifts are maintained in part or
whole, then acceptable levels of performance from operating personnel
is assured through effective control of shift turnovers and plant
activities. No physical plant modifications are involved and none of
the precursors of previously evaluated accidents are affected.
Therefore, this change will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Editorial changes clarify sections 6.2.2.6.b. and 6.2.2.6.c.
without changing the intent or meaning. [...] Changes to sections
4.5.F.3., 4.5.F.4., 4.5 Bases, and 4.7.A.7.a. do not change the intent
or meaning of the Technical Specifications, do not change operating
procedures, and are consistent with surveillance requirements.
[Therefore, the proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.]
Does the proposed license amendment create the possibility of a new
or different kind of accident from any accident previously evaluated?
Establishing operating personnel work hours at ``a normal 8 to 12
hour day, nominal 40-hour week'' allows normal plant operation to be
managed more effectively and does not adversely affect performance of
operating personnel. If 8 hour shifts are maintained in part or whole,
then acceptable levels of performance from operating personnel is
assured through effective control of shift turnovers and plant
activities. Overtime remains controlled by site administrative
procedures in accordance with the NRC Policy Statement on working hours
(Generic Letter 82-12). No physical modification of the plant is
involved. As such, the change does not introduce any new failure modes
or conditions that may create a new or different accident. Therefore,
plant operation in accordance with the proposed amendment will not
create the possibility of a new or different kind of accident from any
previously evaluated.
Editorial changes clarify sections 6.2.2.6.b. and 6.2.2.6.c.
without changing the intent or meaning. [* * *] Changes to sections
4.5.F.3., 4.5.F.4., 4.5 BASES, and 4.7.A.7.a. do not change the intent
or meaning of the Technical Specifications or operating procedures. All
previously performed functions are being maintained.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Does the proposed amendment involve a significant reduction in a
margin of safety?
Establishing operating personnel work hours at ``a normal 8 to 12
hour day, nominal 40-hour week'' allows normal plant operations to be
managed more effectively and does not adversely affect performance of
operating personnel. If 8 hour shifts are maintained in part or whole,
then acceptable levels of performance from operating personnel is
assured through effective control of shift turnovers and plant
activities. Overtime remains controlled by site administrative
procedures in accordance with the NRC Policy Statement on working hours
(Generic Letter 82-12). The proposed change involves no physical
modification of the plant, or alterations to any accident or transient
analysis. [* * *] Therefore, the change does not involve any
significant reduction in a margin of safety.
Editorial changes clarify sections 6.2.2.6.b. and 6.2.2.6.c.
without changing the intent or meaning. [* * *] Changes to sections
4.5.F.3., 4.5.F.4., 4.5 BASES, 4.7.A.7.a. do not change the intent or
meaning of the Technical Specifications or operating procedures.
All previously performed functions are being maintained. Therefore,
the changes do not involve any significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.

Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey

Date of amendment request: December 19, 1997.
Description of amendment request: The proposed amendment would
revise the Hope Creek Generating Station (HCGS) Technical
Specifications (TS) to incorporate changes that reflect the completion
of the Salt Drift Monitoring 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

[[Page 4322]]

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 changes, which update the Terrestrial Ecology Monitoring
Program status, are administrative in nature and in no way affect the
initial conditions, assumptions, or conclusions of the Hope Creek
Generating Station accident analyses. In addition, the proposed changes
would not affect the operation or performance of any equipment assumed
in the accident analyses. Based on the above information, we conclude
that the proposed changes would not significantly increase the
probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
As previously stated, the proposed changes are administrative in
nature and in no way impact or alter the configuration or operation of
the facilities and create no new modes of operation. PSE&G therefore
concludes that the proposed changes would not create the possibility of
a new or different kind of accident.
3. The proposed changes do not involve a significant reduction in a
margin of safety.
The changes are administrative in nature and in no way affect plant
or equipment operation or the accident analysis. PSE&G therefore
concludes that the proposed changes would not result in a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey

Date of amendment request: November 14, 1997.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) to provide surveillance
requirements for the service water accumulator vessels. Specifically,
surveillance requirements are provided for vessel level, pressure and
temperature, and discharge valve response time. The surveillance
requirements are included in TS 3/4.6.1.1 and 3/4.6.2.3, and the
applicable Bases sections are expanded to provide supporting
information.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes provide surveillance requirements for the
Service Water [SW] accumulator tank level, pressure and temperature
parameters and the discharge valve response time test. Supporting
information is included in the Bases section of the applicable
technical specifications. The SW accumulator tank and discharge valve
design has been reviewed and approved by the NRC staff as documented in
NRC Safety Evaluation Report (SER) dated June 19, 1997. The proposed
surveillance requirements do not alter the design as reviewed by the
NRC staff. The addition of tank parameter surveillance requirements to
the technical specifications does not alter the physical plant
arrangement or the installed monitoring instrumentation. The proposed
addition of tank discharge valve response time surveillance
requirements to the technical specifications does not alter the method
of performing these surveillance requirements.
Therefore the proposed changes do not increase the probability of
an accident. The surveillance requirements provide additional controls
for ensuring the SW accumulator tank and discharge valves will be
maintained within the design parameters assumed in the safety analysis.
This provides added assurance that the accumulator tanks and discharge
valves will be capable of performing their required design function
during accident conditions. There is no change to the performance
requirements of these components in preventing two phase flow
conditions and water column separation waterhammer vulnerabilities
identified in GL [Generic Letter] 96-06. Therefore, the proposed
changes do not involve an increase in the consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes provide surveillance requirements for Service
Water Accumulator tank level, pressure and temperature and discharge
valve time response. Supporting information is included in the Bases
section of the applicable technical specifications. The SW accumulator
tank and discharge valve design has been reviewed and approved by the
NRC staff as documented in NRC Safety Evaluation Report (SER) dated
June 19, 1997. The proposed surveillance requirements do not alter the
plant configuration. Installed instrumentation will be used to
accomplish the tank surveillance requirements. The current plant
installation also provides for completion of the discharge valve
response time surveillance utilizing test equipment in accordance with
plant procedures and configurations. Therefore the performance of these
surveillance requirements does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The Service Water Accumulator Vessels and discharge valves were
installed to address the Generic Letter 96-06 issues of column
separation waterhammer and two phase flow in the containment fan coil
unit (CFCU) piping during an accident involving loss of offsite power.
This design has been reviewed and approved by the NRC staff as
documented in NRC Safety Evaluation Report (SER) dated June 19, 1997.
The proposed surveillance requirements do not alter the design as
reviewed by the NRC staff. By providing added assurance that these
components are capable of performing their specified safety function as
assumed in the safety analysis, the additional surveillance
requirements assure system operability to further minimize the
possibility of waterhammer and two phase flow in the CFCU piping during
accident conditions. The proposal therefore minimizes the possibility
of a new or different kind of accident from those previously evaluated
accidents.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The additional surveillances provide added assurance that the
margin of safety assumed in the containment integrity and containment
cooling technical specification will be

[[Page 4323]]

maintained. The additional surveillance requirements further ensure
that in the event the SW accumulator vessels are out of specification
or the discharge valves do not meet their response time requirements,
corrective actions will be completed in accordance with the existing
containment integrity technical specification allowed outage time to
restore containment integrity. The surveillance requirements further
ensure that in the event the SW accumulator vessel or discharge valves
do not meet these requirements, corrective actions will be completed in
accordance with the containment cooling technical specification allowed
outage time to restore the full complement of containment fan coil
units to operability. Since the proposal maintains the margin of safety
provided in the containment integrity and containment cooling technical
specification, there is no reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: John F. Stolz.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California

Date of amendment requests: September 16, 1997.
Description of amendment requests: The licensee proposes to revise
Technical Specification (TS) 3.4.13, ``RCS Operational Leakage,'' TS
5.5.2.11, ``Steam Generator (SG) Tube Surveillance Program,'' and TS
5.7.2, ``Special Reports.'' The proposed change is to allow steam
generator tube repair using ASEA Brown Boveri/Combustion Engineering
(ABB/CE) leak tight sleeving as an alternative steam generator tube
repair to plugging.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The supporting technical evaluation and safety evaluation for the
ASEA Brown Boveri/Combustion Engineering (ABB/CE) leak tight sleeves
demonstrate that the sleeve configuration will provide steam generator
(SG) tube structural and leakage integrity under normal operating and
accident conditions. The sleeve configurations have been designed and
analyzed in accordance with the requirements of the ASME Code.
Mechanical testing has shown that the sleeve and sleeve joints provide
margin above acceptance limits. Ultrasonic Testing (UT) is used to
verify the leak tightness of the weld above the tubesheet. Testing has
demonstrated the leak tightness of the hardroll joint due to the
reinforcing effect of the tubesheet. Tests have demonstrated that tube
collapse will not occur due to postulated Loss of Coolant Accident
(LOCA) loadings.
A new, more conservative, Technical Specification (TS) SG tube
leakage rate requirement is introduced by this change. Accident
analysis assumptions remain unchanged in the event that significant
leakage does occur from the sleeve joint or that the sleeve assembly
ruptures. Any leakage through the sleeve assembly is fully bounded by
the existing SG tube rupture analysis included in the San Onofre
Nuclear Generating Station (SONGS) Updated Final Safety Analysis
Report. Reactor coolant flow reduction from sleeving is addressed by a
ratio of number of tubes sleeved to equal a plugged tube. The proposed
sleeving repair process does not adversely impact any other previously
evaluated design basis accidents.
Therefore, proposed changes do not involve a significant increase
in the probability or consequences of an accident.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Installation of the sleeves does not introduce any significant
changes to the plant design basis. The use of a sleeve to span the area
of degradation of the SG tube restores the structural and leakage
integrity of the tubing to meet the original design bases. Stress and
fatigue analysis of the sleeve assembly shows that the requirements of
the ASME Code are met. Mechanical testing has demonstrated that margin
exists above the design criteria. Any hypothetical accident as a result
of any degradation in the sleeved tube would be bounded by the existing
tube rupture accident analysis.
Therefore, the operation of the facility in accordance with
proposed changes does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The use of sleeves to repair degraded SG tubing has been
demonstrated to maintain the integrity of the tube bundle commensurate
with the requirements of the ASME Code and draft Regulatory Guide (RG)
1.121 and to maintain the primary to secondary pressure boundary under
normal and postulated accident conditions. The safety factors used in
the verification of the strength of the sleeve assembly are consistent
with the safety factors in the ASME Boiler and Pressure Vessel Code
used in SG design. The operational and faulted condition stresses and
cumulative usage factors are bounded by the ASME Code requirements. The
sleeve assembly has been verified by testing to prevent both tube
pullout and significant leakage during normal and postulated accident
conditions. A test program was conducted to ensure the lower hardrolled
joint design was leak tight and capable of withstanding the design
loads. The primary coolant pressure boundary of the sleeve assembly
will be periodically inspected by Non-Destructive Examination to
identify sleeve degradation due to operation.
Installation of the sleeves will decrease the number of tubes which
must be taken out of service due to plugging. There is a small amount
of primary coolant flow reduction due to the sleeve for which the
equivalent sleeve to plug ratio is assigned based on sleeve length. The
ratio is used to assess the final equivalent plugging percentage as an
input to other safety analyses. The sleeve maintains the design basis
requirements for the SG tubing.
Therefore, operation of the facility with the proposed changes 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 requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison

[[Page 4324]]

Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California

Date of amendment requests: October 17, 1997.
Description of amendment requests: The licensee proposes to amend
the licenses for SONGS Units 2 and 3 to revise the Final Safety
Analysis Report (FSAR) to permit digital radiation monitor installation
for both trains supplying the containment purge isolation signal, and
permit digital radiation monitor installation for both trains supplying
the control room isolation signal.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change is required to permit using digital radiation
monitors as input to both trains of the Control Room Isolation Signal
(CRIS), and to both trains of the Containment Purge Isolation Signal
(CPIS). These changes will allow replacement of the remaining safety
related obsolete radiation monitor equipment to address spare parts and
equipment availability issues. The new containment airborne radiation
digital monitor will have the same basic architecture as the existing
analog system, and serves to perform the same function. In addition,
the digital radiation monitors are expected to be more reliable than
the existing equipment which is of an analog design.
Furthermore, defense-in-depth equipment is available that either
provides, or allows for, actions to mitigate the release of offsite and
Control Room doses to within existing licensing limits based on
realistic event input assumptions. Analyses show that if ``realistic''
input assumptions are utilized and reasonable operator actions are
allowed, then acceptable dose consequences result both to the general
public offsite, and to the Control Room operators.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change will permit upgrading the existing analog
radiation monitors with upgraded digital radiation monitors.
Replacement of an analog system to a predominantly digital system, uses
software algorithms to perform the required functions. A satisfactory
software verification and validation (V&V) report, including continued
software change control procedures, provides assurance that a software
common mode failure is not likely.
In addition, the design, installation, testing, maintenance, and
operation of the affected equipment will assure that no new or
different kinds of accidents will be created. The ESFAS radiation
monitors involved are portions of systems that respond to accidents.
They can not, by their actions or inactions, create a new or different
accident from any accident previously evaluated.
Therefore, the proposed change will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The CRIS and CPIS Radiation Monitor Systems provide an accident
mitigation function for offsite doses (10 CFR 100) and Control Room
doses (10 CFR 50 Appendix A, General Design Criteria 19). A change in
the margin of safety is introduced due to the possibility of a software
common mode failure in redundant equipment simultaneously affecting
equipment performing a different function.
This change is not a significant reduction in the margin of safety,
however, due to the following:
(1) A probabilistic risk analysis has determined that the
availability of the affected radiation monitors, including software,
should be better than the existing equipment based on industry data to
date,
(2) The software V&V and preoperational testing to be performed
will provide assurance of system operation, and
(3) The combined occurrence of a software common mode failure that
simultaneously causes failure of all available ESFAS radiation monitors
concurrent with a design bases accident is very unlikely.
In the unlikely event of a software common mode failure that causes
all ESFAS radiation monitors to be inoperable concurrent with a design
bases accident, analyses show that if ``realistic'' input assumptions
are utilized and reasonable operator actions are allowed, then
acceptable dose consequences result both to the general public offsite,
and to the Control Room operators.
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 requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713.
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.

Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of amendment request: December 18, 1997.
Description of amendment request: The proposed amendments would
modify or delete obsolete conditions from the Unit 1 and Unit 2
Operating Licenses. The changes are editorial or administrative in
nature.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes either remove or modify provisions in the
Plant Hatch Unit 1 and Unit 2 Operating Licenses that have been
completed or are otherwise obsolete. Certain Surveillance Requirements
(SRs) that were either added or modified at the time of Improved
Technical Specifications (ITS) implementation were listed in the
Operating Licenses with a schedule for performance. With the exception
of Unit 1 SR 3.8.1.18, all SRs are deleted from the Operating Licenses,
because they have since been performed according to schedule, and will
henceforth be

[[Page 4325]]

performed in accordance with the Technical Specifications.
A requirement for submittal of the Unit 1 inservice inspection plan
for the recirculation and residual heat removal systems' piping is
deleted due to completion of the activity.
Two exemptions granted at Unit 2 startup are deleted due to
completion of the required activities associated with the exemptions.
These were seismic qualification demonstration for the Unit 2 reactor
protection system power supply and completion of the long-term BWR
[boiling water reactor] Owner's Group Mark I containment program.
A requirement to conduct the Unit 2 Initial Test Program according
to the requirements in Chapter 14 of the Final Safety Analysis Report
without major changes is deleted due to completion of the activity. A
condition relating to environmental protection is deleted from the Unit
2 Operating License, since it was superseded by the Environmental
Protection Plan (Nonradiological), Appendix B to the Operating
Licenses. Attachment 2, Items To Be Completed Prior To Opening Main
Steam Isolation Valves, is deleted due to completion of the activities.
The proposed changes discussed above are strictly administrative/
editorial and do not affect the operation or function of any plant
system, component, or structure. Therefore, the proposed changes do not
increase the probability of occurrence or the consequences of a
previously evaluated accident.
2. The proposed changes do not create the possibility of a new and
different type of accident from any previously evaluated.
The proposed administrative/editorial changes do not alter the
operation of any plant system or equipment and do not introduce a new
mode of operation. Thus, the proposed changes cannot create a new
accident initiating mechanism. Therefore, the proposed changes do not
create the possibility of a new and different type of accident from any
previously evaluated.
3. The proposed changes do not involve a significant reduction in
the margin of safety.
Since the proposed changes are strictly administrative/editorial
and do not involve any physical or procedural changes to the plant, the
margin of safety, as defined in the bases for any Technical
Specification is not affected 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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
NRC Project Director: Herbert N. Berkow.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: December 17, 1997.
Description of amendment request: The proposed amendment would
extend the surveillance interval of the containment spray system nozzle
air flow test from five years to ten years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. Operation of the facility in accordance with the proposed
amendment does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change does not result in any hardware changes. The
Containment Spray system trains or nozzles are not assumed to be the
initiators of any analyzed events. Extending the surveillance interval
for performing the Containment Spray system nozzle air flow test from
five to ten years does not represent a significant increase in the
probability of an accident. The Containment Spray system nozzles are
not precursors to any accident analyses.
The Containment Spray system trains and nozzles function to
mitigate the consequences of an analyzed event by providing spray flow
to containment during an accident. The proposed change still provides
assurance that the Containment Spray system nozzles will be maintained
operable due to the passive nature of the design, the materials of
construction, and the low-stress non-wetted environment. The extension
of the surveillance interval does not significantly increase the
probability or consequences of an accident since the nozzle will still
be OPERABLE between surveillance tests.
B. Operation of the facility in accordance with the proposed
amendment does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change does not necessitate a physical alteration of
the plant or changes in parameters governing normal plant operation. No
new or different types of equipment will be installed. The proposed
change will still ensure Containment Spray system nozzle OPERABILITY is
adequately maintained.
C. Operation of the facility in accordance with the proposed
amendment does not involve a significant reduction in a margin of
safety.
The increased interval between the Containment Spray system nozzle
air flow test is acceptable due to the passive design of the nozzles
and industry operating experience as detailed in NURG-1366. The
increased interval is considered acceptable for maintaining nozzle
OPERABILITY. The Containment Spray system, including the nozzles, will
continue to provide their required safety function with the increase
from five to ten years between inspections.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: John Hannon.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: December 31, 1997.
Description of amendment request: The proposed amendment would
revise Technical Specifications 2.1 (Safety Limits), 2.2 (Limiting
Safety System Settings), and 3/4.2.5 (Departure from Nucleate Boiling
Parameters) by including alternate operating criteria to allow
continued plant operation with a reduced measured reactor coolant
system flow rate, if necessary.
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

[[Page 4326]]

consideration, which is presented below:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident previously
evaluated?
The affected Reactor Protection System functions will continue to
provide their current safety function under alternate operating
criteria for reduced measured Reactor Coolant System flow conditions.
The OT Delta-T [Overtemperature Delta-T], OP Delta-T [Overpower Delta-
T], and f(Delta-I) [a function of the indicated difference between top
and bottom detectors of the power-range neutron ion chambers] safety-
analysis reactor trip setpoints have been recalculated to appropriately
reflect the reduced flow conditions. In doing so, the difference, or
margins, between the nominal and maximum values of the reference trip
setpoints (i.e., K1, and K4 for the OT Delta-T and the OP Delta-T
setpoints, respectively) have been maintained so that the Total
Allowance remains unchanged and, therefore, the instrument accuracy
uncertainties are unaffected.
Furthermore, implementation of the provisions for reduced measure
Reactor Coolant System flow under alternate operating criteria for the
South Texas Project Technical Specifications does not increase the
probability or consequences of an accident previously evaluated in the
UFSAR [Updated Final Safety Analysis Report]. This change cannot
directly initiate an accident. The consequences of accidents previously
evaluated in the UFSAR are unaffected by this proposed change because
no change to any equipment response or accident mitigation scenario has
resulted. There are no additional challenges to fission product barrier
integrity. Therefore, the probability of an accident previously
evaluated has not been increased.
(2) Does the proposed license amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
No new failure mechanisms or accident scenarios or limiting single
failures are introduced as a result of this proposed change. Operation
of the plant will be consistent with that previously modeled. All of
the accident analyses previously evaluated in the UFSAR for South Texas
Project Units 1 and 2 have been evaluated to support alternate
operating condition with a 3 percent reduction in the minimum measured
Reactor Coolant System flow. The new nominal Reactor Coolant System
operating conditions supported by these evaluations have been
determined. Revised Core Thermal Safety Limits have been established
and will be incorporated into the Technical Specifications for the 3
percent Reactor Coolant System measured flow reduction; and, the OT
Delta-T and OP Delta-T setpoints are re-calculated based on the new
Safety Analysis Limits, appropriate for the reduced flow operation.
These reactor protection system functions affected by the change in
operating conditions will, therefore, continue to provide an
appropriate response equivalent to current safety analysis modeling.
The proposed Technical Specification amendment does not challenge the
performance or integrity of safety-related systems. The possibility of
a new or different kind of accident, therefore, is not created.
(3) Does the proposed amendment involve a significant reduction in
a margin of safety?
The modification will have no effect on the availability,
operability, or performance of the South Texas Project safety-related
systems and components. This is based on: the evaluation performed of
all accidents previously evaluated in the UFSAR for operation of South
Texas Project Units 1 and 2 at reduced Reactor Coolant System flow
conditions; establishment of revised Core Thermal Safety Limits that
are reflected in the proposed Technical Specification applicable for
the 3 percent Reactor Coolant System flow reduction; and, the
appropriately re-calculated OT Delta-T and OP Delta-T setpoints, also
applicable for these reduced flow conditions. Allowing provision for
these alternate operating criteria does not prevent inspections or
surveillance required by the Technical Specifications. The margin of
safety associated with the acceptance criteria for any accident is
unchanged, and therefore, the proposed modification will not reduce the
margin of safety as defined in the Bases of the South Texas Project
Technical Specifications.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: John N. Hannon.

The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power
Plant, Unit 1, Lake County, Ohio

Date of amendment request: December 23, 1997.
Description of amendment request: The license amendment request
proposes changes to technical specification surveillances to remove the
requirements related to accelerated testing of the standby emergency
diesel generators, consistent with the recommendations in NRC Generic
Letter 94-01, ``Removal of Accelerated Testing and Special Reporting
Requirements for Emergency Diesel Generators.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes do not significantly increase the probability
of occurrence of a previously evaluated accident because the standby
diesel generators (including the High Pressure Core Spray [HPCS] diesel
generator) are not initiators of previously evaluated accidents. The
standby diesel generators mitigate the consequences of previously
evaluated accidents involving a loss of offsite power. The Perry
Nuclear Power Plant (PNPP) program developed to meet the Maintenance
Rule (10 CFR 50.65) will continue to ensure the diesel generators
perform their function when called upon. The change to the surveillance
frequency does not affect the design of the diesel generators, the
operational characteristics of the diesel generators, the interfaces
between the diesel generators and other plant systems, the function, or
the reliability of the diesel generators. Thus, the diesel generators
will be capable of performing their accident mitigation function, there
is no impact to the radiological consequences of any accident analysis,
and the probability and consequences of previously evaluated accidents
are not increased by this activity.
2. The proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed activity involves a change to the frequency for
specific technical specification surveillance requirements. No physical
or

[[Page 4327]]

operational changes to the diesel generators or supporting systems are
made by this activity. Since the proposed changes do not involve a
change to the plant design or operation and thus no new system
interactions are created by this change, these changes do not produce
any parameters or conditions that could contribute to the initiation of
accidents different from those already evaluated in the Updated Safety
Analysis Report. The proposed changes only address the methods used to
ensure diesel generator reliability. Thus, the proposed amendment does
not create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The proposed changes involve the methods used to ensure diesel
generator performance and reliability. No changes, other than to
frequency, are made to Technical Specification Surveillance
Requirements 3.8.1.2 and 3.8.1.3. The NRC, in Generic Letter 94-01, has
acknowledged the acceptability of the use of the Maintenance Rule
program for the diesel generators to ensure diesel generator
performance in lieu of accelerated testing. These proposed changes do
not involve a change to the plant design or operation, and thus do not
affect the design of the diesel generators, the operational
characteristics of the diesel generator, the interfaces between the
diesel generators and other plant systems, or the function or
reliability of the diesel generators. Because the diesel generator
performance and reliability will continue to be ensured by the diesel
generator program to meet the Maintenance Rule, the proposed changes do
not result in a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Richard P. Savio.

Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: December 23, 1997.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 4.4.5, ``Reactor Coolant
System--Steam Generators--Surveillance Requirements (SRs).'' SR 4.4.5.8
would be modified to provide flexibility in the scheduling of steam
generator inspections during refueling outages.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Davis-Besse Nuclear Power Station has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power Station,
Unit No. 1, in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no change is being made to any
accident initiator. No previously analyzed accident scenario is
changed, and initiating conditions and assumptions remain as previously
analyzed. The proposed change to Technical Specification (TS)
Surveillance Requirement (SR) 4.4.5.8, to allow performance of required
visual inspections of the secured internal auxiliary feedwater header,
header to shroud attachment welds, and the external header thermal
sleeves during the third period of the ten-year Inservice Inspection
Interval, does not affect any Updated Safety Analysis Report (USAR)
accident initiators. These inspections will continue to take place at a
prescribed time interval scheduled similar to American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI
components. Therefore, it can be concluded that the proposed change
does not involve a significant increase in the probability of an
accident previously evaluated.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed change does not
affect accident conditions or assumptions used in evaluating the
radiological consequences of an accident. The proposed change does not
alter the source term, containment isolation or allowable radiological
releases.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
change does not alter the way the plant is operated, and no new or
different failure modes have been defined for any plant system or
component important to safety, nor has any limiting single failure been
identified as a result of the proposed changes.
These inspections were established to ensure that there are no new
failure mechanisms resulting from these components. These inspections
will continue to take place in the third period of each inservice
inspection interval. No new or different types of failures or accident
initiators are introduced by the proposed changes.
3. Not involve a significant reduction in a margin of safety
because visual inspections will be performed on a prescribed frequency
that is consistent with the schedules established for ASME Code
components in accordance with ASME Code Section XI.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.

Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: December 23, 1997.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 1.0, ``Definitions,'' to
clarify the meaning of core alteration; would relocate TS Section 3/
4.9.5, ``Refueling Operations--Communications,'' and the associated
bases to the Technical Requirements Manual; and would add TS Section
3.0.6 and the associated bases to address the return to service of
inoperable equipment.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards

[[Page 4328]]

consideration, which is presented below:
The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the
proposed changes and determined that a significant hazards
consideration does not exist because operation of the Davis-Besse
Nuclear Power Station, Unit Number 1, in accordance with these changes
would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because the probability of previously
analyzed accidents is not affected by the criteria in the core
alteration definition (Technical Specification (TS) 1.12). Nor do these
changes, the proposed relocation of the refueling communications TS 3/
4.9.5 and Bases to the DBNPS Updated Safety Analysis Report (USAR)
Technical Requirements Manual (TRM), or the proposed addition of new TS
3.0.6 and Bases regarding return to service of inoperable equipment,
affect any accident initiator, or assumption made in any safety
analysis. The proposed changes are administrative in nature and are
consistent with NUREG-1430, Revision 1, ``Standard Technical
Specifications, Babcock and Wilcox Plants,'' dated April 1995, as
modified by a pending NUREG-1430 change approved by the NRC, Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler Number 165.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
affect accident conditions or assumptions used in evaluating the
radiological consequences of an accident. The proposed changes do not
significantly alter the source term, containment isolation, or
allowable radiological releases.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because the proposed
changes do not change the way the plant is operated. No new or
different types of failures or accident initiators are introduced by
the proposed changes.
3. Not involve a significant reduction in a margin of safety
because no inputs into the calculation of any Technical Specification
Safety Limit, Limiting Safety System Settings, Technical Specification
Limiting Condition for Operation, or other previously defined margins
for any structure, system, or component important to safety are being
affected 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 Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH.
Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &
Brockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Acting Project Director: Richard P. Savio.

Yankee Atomic Electric Company, Docket No. 50-029, Yankee Nuclear Power
Station, Franklin County, Massachusetts

Date of amendment request: December 18, 1997.
Description of amendment request: By letter dated May 15, 1997, the
licensee submitted a License Termination Plan. The NRC previously
published a notice dated August 14, 1997, in the Federal Register (62
FR 43559) advising of receipt of the Plan. The proposed request is for
a license amendment approving the Plan for the Yankee Nuclear Power
Station.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. Accident analyses are
included in the approved Decommissioning Plan and incorporated into the
FSAR. All decommissioning and fuel storage activities described in the
License Termination Plan are consistent with those in the approved
Decommissioning Plan. No systems, structures, or components that could
initiate or be required to mitigate the consequences of an accident are
affected by the proposed change in any way not previously evaluated in
the approved Decommissioning Plan. Therefore, the proposed change is
administrative in nature and does not involve an increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated. Accident analyses are included
in the approved Decommissioning Plan and are incorporated into the
FSAR. All decommissioning and fuel storage activities described in the
License Termination Plan are consistent with those in the approved
Decommissioning Plan. The proposed change does not affect plant
systems, structures, or components in any way not previously evaluated
in the approved Decommissioning Plan, and no new or different failure
modes will be created. Therefore, the proposed change is administrative
in nature and does not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety. Approval
of the License Termination Plan by license amendment is administrative
in nature since all decommissioning and fuel storage activities
described in the License Termination Plan are consistent with those in
the approved Decommissioning Plan. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Greenfield Community College,
1 College Dr

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A98-1904. Public record. Not legal advice.