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

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## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** February 10, 1999
- **Citation:** 64 FR 6692

## 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 tendency 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 15, 1999, through January 29, 1999.
The last biweekly notice was published on January 27, 1999 (64 FR
4152).

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

[[Page 6693]]

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 March 12, 1999, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspects(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 preset 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

[[Page 6694]]

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.

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: December 7, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs) to permit a one-time only
extension of the steam generator tube inspection interval for fuel
cycle 14 and delete the requirement to have NRC staff concurrence of
the steam generator examination program. Specifically, TS 4.13A.2.a
would be revised with a footnote that states ``Examinations scheduled
for 1999 only, shall be conducted during the 2000 Refueling Outage
which will commence no later than June 3, 2000. The scheduled
examinations will be completed prior to return to service from the 2000
Refueling Outage.'' In addition, TS 4.13C.1 would be revised to state
``The proposed steam generator examination program shall be submitted
for NRC staff review at least 60 days prior to each scheduled
examination.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 59.91(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 does not involve any physical modifications
to the plant or modification in the methods of plant operation which
could increase the probability or consequences of previously
evaluated accidents. The proposed change permits an extension of the
current steam generator tube inservice inspection cycle. This
extension would allow the steam generator tube examinations to be
conducted during the 2000 refueling outage which will commence no
later than June 3, 2000. The basis for acceptance of this increase
in the technical specification limit is the ``non-operating'' steam
generator time between the last examination and the upcoming
examination. Extending the steam generator ``operating'' duration by
48 days would not significantly increase wear which might lead to
tube failure. No appreciable steam generator tube wear or
degradation is expected as a result of this extension. This change
will not affect the scope, methodology, acceptance limits and
corrective measures of the existing steam generator tube examination
program. The probability and consequences of failure of the steam
generators due to leaking or degraded tubes is not increased by the
proposed change. Additionally the proposed administrative change to
delete the requirement to receive NRC concurrence of the proposed
steam generator examinations will have no bearing on the actual
results of the steam generator examinations. Therefore, the
probability and the consequence of a design basis accident are not
being increased by the proposed change.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Plant systems and components will not be operated in a different
manner as a result of the proposed Technical Specification change.
The proposed change permits the upcoming steam generator tube
examination to be conducted during the 2000 refueling outage that
will commence no later than June 3, 2000. There are no plant
modifications or changes in methods of operation. This extension is
based upon the ``non-operating'' steam generator time between the
last examination and the upcoming examination. Extending the steam
generator ``operating'' duration by an additional 48 days would not
significantly increase wear which might lead to tube failure. The
proposed extension will not increase the probability of occurrence
of a tube rupture, increase the probability or consequences of an
accident, or create any new accident precursor. Additionally the
proposed administrative change to delete the requirement to receive
NRC concurrence of the proposed steam generator examinations will
have no bearing on the actual results of the steam generator
examinations. Therefore, the possibility of an accident of a
different type than was previously evaluated in the safety analysis
report is not created by the proposed change to the Technical
Specification.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change to Technical specification section 4.13A.2.a
will not reduce the margin of safety. This amendment involves an
extension of the current steam generator tube inservice inspection
cycle. The basis for acceptance of this increase in the technical
specification limit is the ``non-operating'' steam generator time
between the last examination and the upcoming examination. Extending
the steam generator ``operating'' duration by an additional 48 days
would not significantly increase wear which might lead to tube
failure. No appreciable steam generator tube wear or degradation is
expected as a result of this extension. Additionally the proposed
administrative change to delete the requirement to receive NRC
concurrence of the proposed steam generator examinations will have
no bearing on the actual results of the steam generator
examinations. Therefore, the accident analysis assumptions for
design basis accidents are unaffected and the margin of safety is
not decreased by the proposed Technical Specification change.

[* * *]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: S. Singh Bajwa, Director.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas

Date of amendment request: April 30, 1998.
Description of amendment request: The proposed amendment revises
the definition of quadrant power tilt to clearly allow the use of
either the incore detectors or the excore detectors for determining
quadrant power tilt.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Criterion 1--Does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to the quadrant power tilt (QPT) definition
will not alter any Safety Analysis Report (SAR) assumptions
established and implemented by the technical specifications. The
proposed change will allow the use of either the incore detectors or
the excore power range detectors for determining QPT. This change is
consistent with the improved Standard Technical Specifications (STS)
which has been previously approved by the NRC. QPT measured by
incore detectors provides a more accurate indication of reactor core
power distribution than the value determined from the excore
detectors. The accident prevention and mitigation features of the
plant are not affected by this proposed amendment.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.

[[Page 6695]]

Criterion 2--Does not create the possibility of a new or
different kind of accident from any previously evaluated.
The proposed change to the definition of QPT does not alter the
ANO-1 SAR analysis or core operating limits report (COLR). The
change will clearly permit the use of either the incore detectors or
the excore detectors for monitoring QPT. The design and physical
configuration of the plant are not affected by this change.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does not involve a significant reduction in the
margin of safety.
The proposed change to the QPT definition incorporates the
improved TS definition contained in NUREG-1430. The revised
definition allows the use of either the incore detectors or the
excore power range detectors for determination of QPT. The change
does not vary or affect any of the plant's operating parameters. The
COLR currently specifies acceptable QPT limits based upon the
measurement techniques. These limits are based upon the unique
measurement characteristics of the incore and excore power range
detectors and assure the measurement independent limit is not
violated.

Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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, NW., Washington, DC 20005-3502.
NRC Project Director: John N. Hannon.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One,
Unit No. 1, Pope County, Arkansas

Date of amendment request: August 6, 1998.
Description of amendment request: The proposed amendment revises
the minimum and the maximum concentration limits for the sodium
hydroxide tank. The proposed change also revises the minimum specified
tank volume to refer to the parameter used in the analysis with no
allowance for instrument uncertainty and deletes the maximum specified
tank volume.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Criterion 1--Does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Sodium hydroxide is not an accident initiator. It is, however, a
contributor to the mitigation of the effects of a Loss-of-Coolant-
Accident (LOCA). The proposed change in NaOH tank concentration
results in changing the expected post-LOCA reactor building sump pH.
The reduction in the lower value of sump pH, from 8.5 to 7.0, is
acceptable based on guidance contained in NUREG-0800, Standard
Review Plan, Section 6.5.2, ``Containment Spray as a Fission Product
Cleanup System Review Responsibilities,'' Revision 2, December 1988.
This guidance allows the assumption of long-term iodine retention
when the equilibrium sump pH, after mixing and dilution with the
primary coolant and ECCS injection, is above 7.0. Although the
change allows the volume of the NaOH tank to be maintained at a
lower volume, the proposed minimum volume bounds the analyses of
concern.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does not create the possibility of a new or
different kind of accident from any previously evaluated.
Sodum hydroxide is added for iodine removal and for pH
adjustment of the borated water in the reactor building sump
following a LOCA. The proposed changes in NaOH tank concentration
and volume introduce no new mode of plant operation.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--Does not involve a significant reduction in the
margin of safety.
The proposed change in NaOH tank concentration results in
changing the expected post-LOCA reactor building sump pH. This
proposed change does involve an incremental reduction in the margin
to safety since iodine retention is dependent on the pH of the sump/
spray solution. However, this reduction is not considered
significant in that the effect of the change in sump pH, from 8.5 to
7.0 has a relatively minor effect on iodine retention, as supported
by Standard Review Plan (NUREG-0800), Section 6.5.2, Revision 2,
dated December 1988. Although the change allows the volume of the
NaOH tank to be maintained at a lower volume, the proposed minimum
volume bounds the analyses of concern.
Therefore, this change does not involve a significant reduction
in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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, NW., Washington, DC 20005-3502.
NRC Project Director: John N. Hannon.

Entergy Operations, Inc., Et Al., Docket No. 50-416, Grand Gulf
Nuclear Station, Unit 1, Claiborne County, Mississippi and Entergy
Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458,
River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: January 12, 1999, superceding the
amendment request in the letter of September 30, 1996, for both
stations.
Description of amendment request: The proposed amendment would add
an additional required action to the Limiting Condition for Operation
(LCO) 3.9.1, ``Refueling Equipment Interlocks,'' of the Technical
Specifications for both stations. The additional action would allow an
alternative to the current action for one or more inoperable refueling
equipment interlocks. The current action is to ``suspend in-vessel fuel
movement with equipment associated with the inoperable interlock(s).''
The alternative action proposed is to (1) insert a control rod
withdrawal block, and (2) verify all control rods are fully inserted in
core cells containing one or more fuel assemblies. The proposed
amendment would also revise the Bases for the LCO 3.9.1 actions to
describe the proposed alternative actions. The previous Federal
Register notice of the amendment request in the superceded letter of
September 30, 1996, was issued on June 16, 1996, (61 FR 31178), for
Grand Gulf Nuclear Station (GGNS).
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:

I. The proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
The refueling interlocks are explicitly assumed in the GGNS
Updated Final Safety Analyses Report (UFSAR) and RBS Updated Safety
Analyses Report (USAR) analysis of the control rod removal error or
fuel loading error during refueling. This analysis evaluates the
probability and consequences of control rod withdrawal during
refueling. Criticality and, therefore, subsequent prompt reactivity
excursions are prevented during

[[Page 6696]]

the insertion of fuel, provided all required control rods are fully
inserted during the fuel insertion. The refueling interlocks
accomplish this by preventing loading fuel into the core with any
control rod withdrawn, or by preventing withdrawal of a rod from the
core during fuel loading.
When the refueling interlocks are inoperable the current method
of preventing the insertion of fuel when a control rod is withdrawn
is to prevent fuel movement. This method is currently required by
the Technical Specifications. An alternate method to ensure that
fuel is not loaded into a cell with the control rod withdrawn is to
prevent control rods from being withdrawn and verify that all
control rods required to be inserted are fully inserted. The
proposed actions will require that a control rod block be placed in
effect thereby ensuring that control rods are not subsequently
inappropriately withdrawn. Additionally, following placing the
control rod withdrawal block in effect, the proposed actions will
require that all required control rods be verified to be fully
inserted. This verification is in addition to the requirements to
periodically verify control rod position by other Technical
Specification requirements. These proposed actions will ensure that
control rods are not withdrawn and cannot be inappropriately
withdrawn because an electrical or hydraulic block to control rod
withdrawal is in place. Like the current requirements the proposed
actions will ensure that unacceptable operations are blocked (e.g.,
loading fuel into a cell with a control rod withdrawn except
following the requirements of LCO 3.10.6, ``Multiple Control Rod
Removal--Refueling,'' which is unaffected by this change).
The proposed additional acceptable Required Actions provide an
equivalent level of assurance that fuel will not be loaded into a
core cell with a control rod withdrawn as the current Required
Action or the Technical Specification Surveillance Requirement.
Therefore, the proposed change does not significantly increase the
probability or consequences of an accident previously evaluated.
II. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The change in the Technical Specification requirements does not
involve a change in plant design. The proposed requirements will
continue to ensure that fuel is not loaded into the core when a
control rod is withdrawn except following the requirements of LCO
3.10.6, ``Multiple Control Rod Removal-Refueling,'' which is
unaffected by this change.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
III. The proposed change does not involve a significant
reduction in a margin of safety.
As discussed in the Bases for the affected Technical
Specification requirements, inadvertent criticality is prevented
during the insertion of fuel provided all required control rods are
fully inserted during the fuel insertion. The refueling interlocks
function to support the refueling procedures by preventing control
rod withdrawal during fuel movement and the inadvertent loading of
fuel when a control rod is withdrawn.
The proposed change will allow the refueling interlocks to be
inoperable and fuel movement to continue only if a control rod
withdrawal block is in effect and all required control rods are
verified to be fully inserted. These proposed Required Actions
provide an equivalent level of protection as the refueling
interlocks by preventing a configuration which could lead to an
inadvertent criticality event. The refueling procedures will
continue to be supported by the proposed required actions because
control rods cannot be withdrawn and as a result fuel cannot be
inadvertently loaded when a control rod is withdrawn except
following the requirements of LCO 3.10.6, ``Multiple Control Rod
Removal--Refueling,'' which is unaffected by this change.
Therefore, the proposed changes do not cause 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: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120, for Grand Gulf
Nuclear Station, and Government Documents Department, Louisiana State
University, Baton Rouge, LA 70803, for River Bend Station.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502, for
Grand Gulf Nuclear Station, and Mark Wetterhahn, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005, for River Bend
Station.
NRC Project Director: John N. Hannon.

Florida Power and Light Company, Et Al., Docket No. 50-335, St.
Lucie Plant, Unit No. 1, St. Lucie County, Florida

Date of amendment request: November 22, 1998.
Description of amendment request: The proposed amendment would
revise the reactor thermal margin safety limit lines and flow rates
stated in the technical specifications (TS). The amendment would also
update the reference for dose conversion factors used in Dose
Equivalent Iodine-131 calculations, and administrative changes to the
criticality analysis uncertainty described in TS 5.6.1.a.1, update the
analytical methods used in determining core operating limits listed in
TS 6.9.1.11, and revise the TS bases for the steam generator pressure-
low trip setpoint.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Replacement of the St. Lucie Unit 1 steam generators in 1997
resulted in an increase in RCS [reactor coolant system] flow. The
proposed amendment would increase the values of design minimum
reactor coolant flow and the low flow trip setpoint presently stated
in the Technical Specifications (TS). These revisions are
accompanied by a corresponding change to the Thermal Margin Safety
Limit Lines of TS Figure 2.1-1. The RCS flow related revisions do
not change the probability of any previously evaluated accident, as
they do not impact any plant component, structure or system
affecting the accident initiators. The proposed changes would
continue to maintain adequate operational margin to TS limits for
RCS flow and the low-flow trip setpoint.
The proposed changes to the thyroid dose conversion factors from
TID-14844 to ICRP-30, fuel storage TS 5.6.1.a.1, the list of
analytical methods in TS 6.9.1.11, and the Bases for Steam Generator
Pressure-Low trip setting have no relevance to the accident
initiators, and thus do not affect the frequency of occurrence of
previously analyzed transients. Additionally, there are no changes
to any active plant component due to these proposed changes.
The supporting evaluation of proposed TS changes demonstrates
acceptable results for all the accidents previously analyzed, and it
is concluded that the radiological consequences would remain within
their established acceptance criteria when including the effects of
increased RCS flow, increased low flow trip setpoint, and change to
the thyroid dose conversion factors used in the determination of
dose consequences. Proposed changes to the Bases for the Steam
Generator Pressure-Low trip setpoint, fuel storage design features,
and the list of analytical methods in TS 6.9.1.11 are administrative
in nature and do not impact current safety analyses.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
This proposed amendment revises limiting flow parameters to
derive analysis benefits from increased RCS flow due to the
replacement stream generators, while assuring safe plant operation
commensurate with the proposed RCS flow and low flow

[[Page 6697]]

trip setpoint changes. These changes along with the proposed changes
to the Bases for the Steam Generator Pressure-Low trip setpoint,
dose conversion factors, the list of analytical methods in TS
6.9.1.11, and the fuel storage design features do not require
modifications to the plant configuration, systems or components
which would create new failure modes. There would be no change in
the modes of operation of the plant. The design functions of all the
safety systems remain unchanged. Therefore, operation of the
facility in accordance with the proposed amendment would not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendment revises limiting flow parameters to
derive analysis benefits from increased RCS flow due to the
replacement steam generators, while assuring safe plant operation
commensurate with the proposed design minimum RCS flow and low-flow
trip setpoint changes. FPL has evaluated the impact of the proposed
changes on available margin to the acceptance criteria for Specified
Acceptable Fuel Design Limits (SAFDL), 10 CFR 50.46(b) requirements,
primary and secondary over-pressurization, peak containment
pressure, potential radioactive releases, and existing limiting
conditions for operation. With the proposed changes to the design
minimum RCS flow, low-flow trip setpoint, and dose conversion
factors, FPL has concluded that there would be no adverse impact to
the existing safety analyses. The proposed changes to the Bases for
the Steam Generator Pressure-Low trip setpoint, the list of
analytical methods in TS 6.9.1.11, and the fuel storage design
features are administrative in nature. Therefore, operation of the
facility in accordance with the proposed amendment would 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Cecil O. Thomas.

Florida Power and Light Company, Et Al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida

Date of amendment request: December 18, 1998.
Description of amendment request: The proposed amendment would
revise the St. Lucie Unit 2 Plant Technical Specifications (TS) Index
Page III; TS 1.10, Dose Equivalent I-131; TS 2.1.1.2, Linear Heat Rate;
Bases 2.1.1, Reactor Core; Bases Figure B2.1-1, Axial Power
Distributions for Thermal Margin Safety Limits; Bases 2.2.1, Reactor
Trip Setpoints (Variable Power Level-High); TS 3.1.1.1/4.1.1.1.1,
Shutdown Margin--Tavg Greater Than 200 deg.F; TS 3/4.1.1.2, Shutdown
Margin--Tavg Less Than or Equal to 200 deg.F; TS 3.1.2.2, Boration
Systems Flow Paths--Operating; TS 3.1.2.4, Charging Pumps--Operating;
TS 3.1.2.6, Boric Acid Makeup Pumps--Operating; TS 3.1.2.8, Borated
Water Sources--Operating; Bases 3/4.1.1.1 and 3/4.1.1.2, Shutdown
Margin; Bases 3/4.1.2, Boration Systems; and TS 6.9.1.11, Core
Operating Limits Report (COLR). The core operating limits for shutdown
margin will be relocated to the St. Lucie Unit 2 COLR.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the license has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:

1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendment involves changes to the dose conversion
factors used in the thyroid dose calculations and the relocation of
the SHUTDOWN MARGIN requirements for Modes 1 through 5 from TS to
the Core Operating Limits Report (COLR). Additionally, the peak
linear heat rate value corresponding to centerline melt is deleted
from the TS. The deletion of this TS remains consistent with the
requirements of 10 CFR 50.36. Bases Figure B2.1-1 is replaced with a
new figure, consistent with the input assumptions of the safety
analysis report.
The proposed amendment addresses analytical methods changes such
as the use of HERMIT code in one dimensional mode for spatial
details, the rod bow penalty calculations using L2/I
dependence discussed in CEN-289 (A)-P, CEAW methodology change for
crediting the delta-T power trip, and the methodology for core
designs containing Gadolinia-Urania burnable absorbers (CENPD-275-P,
Revision 1-P, Supplement 1-P). None of these changes is a
contributor to the initiation of previously evaluated accidents. The
changes to TS bases and the COLR methodology changes have no impact
on the accident initiators. Accordingly, the probability of an
accident previously evaluated is not significantly increased.
The proposed changes have been evaluated by Florida Power &
Light (FPL) and Asea Brown Boveri--Combustion Engineering (ABB-CE).
The safety analyses assumed bounding physics parameters, and satisfy
all the applicable acceptance criteria. Although specification
2.1.1.2 is deleted from TS, the safety analyses continue to meet the
same centerline melt acceptance criteria as before and from which
the peak linear heat rate value is derived. Additionally, the peak
linear heat rate value (corresponding to the centerline melt) does
not meet the criteria specified in 10 CFR 50.36 for safety limits.
The changes to TS bases do not affect safety analysis results.
The relocation of SHUTDOWN MARGIN requirements to COLR does not
affect analysis results or consequences as the limits remain
unchanged. Future changes to these limits will be controlled per
Generic Letter 88-16 under the provisions of 10 CFR 50.59.
The use of HERMITE code in one dimension, for space-time loss-
of-flow simulation, has been successfully applied for other ABB-CE
plants. The use of HERMITE code in this mode, for St. Lucie Unit 2,
is acceptable since there are no fundamental core and nuclear steam
supply system (NSSS) differences between St. Lucie Unit 2 and these
plants. The analyses presented in this submittal include the use of
a supplement to the gadolinia-urania core design methodology topical
report. The change in the rod bow penalty effects similar to that
approved for another ABB-CE plant is justified for St. Lucie Unit 2
based on a comparative analysis of factors influencing the rod bow.
The change in the CEAW analysis method removes unnecessary
conservatisms as compared to the previous analysis method. The
validity of results and conclusions of this evaluation are
contingent upon NRC approval of these revised methods.
The radiological does consequences for applicable safety
analyses, using the dose conversion factors from ICRP-30, Supplement
to Part 1, satisfy the acceptance criteria established to ensure
compliance with the 10 CFR 100 dose limits.
The COLR methodology changes proposed to be listed in TS are
those previously approved for CE plants with changes as described
above. The use of these methodologies remains consistent with their
applicability for safety analyses.
Therefore, the proposed changes do not significantly increase
the probability or consequences of an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendment involves changes to the Technical
Specifications for the dose conversion factors used in the thyroid
dose calculations, the deletion of TS 2.1.1.2, the replacement of
Bases Figure B2.1-1, and the relocation of SHUTDOWN MARGIN
requirements to the COLR. Additionally, there are methodology
changes related to the safety analyses reported in this submittal.
The methodology changes include the use of HERMITE code in one
dimensional mode for space-time loss-of-flow simulations, revised
rod bow DNB penalty calculations, CEAW analysis methodology change
including the use of delta-T power trip, and

[[Page 6698]]

a supplement to the methodology for core designs containing
Gadolinia-Urania burnable absorbers (CENPD-275-P Revision I-P,
Supplement I-P). None of these changes, including those of the TS
bases, will affect the plant configuration and there will be no
impact on any system performance.
Therefore, this amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed changes to the Technical Specifications have been
evaluated with respect to the safety analyses using either
previously approved methodology or methodology currently under NRC
review (CENPD-275-P, Revision I-P, Supplement
I-P). The use of HERMITE code in one-dimensional mode for spatial
details, for space-time loss-of-flow simulation, provides more accurate
data for thermal margin calculations and has been used for similar
applications at other plants. The calculations of rod bow DNB penalty
using L\2\/I dependence has been previously approved for another ABB-CE
plant and is justified for St. Lucie Unit 2 based on an analysis of
important factors influencing the rod bow. The CEAW methodology change
showed acceptable analysis results after conservatively accounting for
appropriate uncertainties.
The safety analyses performed with this methodology used
bounding physics parameters to allow flexibility for future cycles
core designs. The revised Bases Figure B2.1-1 is consistent with the
attached safety analysis report. Deleting TS 2.1.1.2 is justified
since the specified limit does not meet any of the criteria of 10
CFR 50.36, and the fuel centerline melt criteria applied to the
Specified Acceptable Fuel Design Limit (SAFDL) is not changed. The
setpoint analyses and safety analyses of all design basis accidents
meet the applicable acceptance criteria with respect to the
radiological consequences, SAFDLs, primary and secondary
overpressurization, and 10 CFR 50.46 requirements. The proposed
amendment, therefore, will not involve a significant reduction in
the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Cecil O. Thomas.

Florida Power and Light Company, Et Al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

Date of amendment request: December 16, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification 6.3, ``Unit Staff Qualifications,'' and
add specific staff qualifications for a Multi-Discipline Supervisor
position.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed changes are administrative in nature addressing
personnel qualification issues. The Multi-Discipline Supervisor
(MDS) position will be filled with personnel who are experienced in
one or more technical disciplines (maintenance, operations,
engineering, or other related technical discipline). Fundamental
working knowledge of tasks being performed will be acquired through
the MDS initial training program. The training concentrates on
developing the skills and knowledge of an MDS to safely oversee
tasks for multi-discipline work teams. Therefore, four years
experience in any related technical discipline or disciplines
combined with the MDS training program provide adequate technical
knowledge for proper job oversight. These proposed changes will not
involve a significant increase in the probability or consequences of
an accident previously evaluated because they do not affect
assumptions contained in plant safety analyses, the physical design
and/or operation of the plant, nor do they affect Technical
Specifications that preserve safety analysis assumptions. Therefore,
operation of either facility in accordance with its proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The changes being proposed are administrative in nature and do
not affect assumptions contained in plant safety analyses the
physical design and/or modes of plant operation defined in the
facility operating license, or Technical Specifications that
preserve safety analysis assumptions. These changes address
qualification requirements for the MDS position. Since the proposed
changes do not change the qualifications for those individuals
responsible for the actual licensed operation of the facility,
operation of the facility in accordance with the proposed amendments
would not create the possibility of a new or different kind of
accident from any accident previously evaluated. No new failure mode
is introduced due to the administrative changes since the proposed
changes do not involve the addition or modification of equipment nor
do they alter the design or operation of affected plant systems,
structures, or components. Therefore, operation of either facility
in accordance with its proposed amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems, structures, and components are unchanged by the proposed
amendments. The proposed changes to add the MDS position have
management and administrative controls associated with the required
qualification requirements. The St. Lucie Unit 1 and Unit 2
Technical Specifications will ensure that any individual filling the
MDS position has the requisite education, experience, and training.
The proposed changes do not alter the basis for any technical
specification that is related to the establishment of, or the
maintenance of, a nuclear safety margin. Therefore, operation of
either facility in accordance with its proposed amendment would 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Direct: Cecil O. Thomas.

GPU Nuclear Inc. Et Al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey

Date of amendment request: September 3, 1998.
Description of amendment request: The amendment would revise
Technical Specifications 3.4.A.10.e and 3.5.a.2.e to incorporate a
Condensate Storage Tank level of greater than 35 feet.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the

[[Page 6699]]

licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:

1. The proposed TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change does not alter the design or function of any
structures, systems or components and does not affect any of the
parameters or conditions that could contribute to initiation of any
accidents.
The proposed change eliminated an inconsistency between the
noted tank level and required water volume and, thereby, ensures
360,000 gallons of water are available for use. The proposed change
does not affect the volume of water required to be available, the
conditions under which it must be available nor the manner in which
it will be used. Therefore, the proposed TS change does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Eliminating an inconsistency between the noted tank level and
the required water volume does not alter the designs or function of
any structures, systems or components. The proposed tank level
requirement is within the design parameters of the tank and, as
such, does not [ ] introduce any new mechanisms which could
contribute to the creation of a new or different kind of accident
than previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed change eliminates an inconsistency between the
noted tank level and required water volume. The proposed change
ensures that an adequate makeup source is available and, in
addition, that sufficient water volume is available to support
operation of the core spray system in the event of a reactor vessel
leak. Therefore, the proposed TS 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: 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: William M. Dean.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1. (NMP1) Oswego County, New York

Date of amendment request: December 30, 1998.
Description of amendment request: The footnote of current Technical
Specification (TS) Table 3.6.14-2, Radioactive Gaseous Effluent
Monitoring Instrumentation, specifies that the requirement for the
emergency condenser system to have one operable noble gas activity
monitor per vent, is applicable during reactor power operating
conditions. Note (h) of current TS Table 4.6.14-2 specifies that the
requirement to perform a sensor check once per day of the emergency
condenser system noble gas activity monitor is applicable during
reactor power operating conditions. The proposed amendment would change
the footnote of TS Table 3.6.14-2 and note (h) of TS Table 4.6.14-2 to
extend the applicability of the channel operability and daily sensor
check surveillance requirement from during reactor power operating
conditions, to during power operation conditions and whenever the
reactor coolant temperature is greater than 212 deg.F except for
hydrostatic testing with the reactor not critical. The proposed changes
would also correct a clerical error in TS 4.6.15.d. The clerical error
cited an incorrect TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes extend the application of operability and
daily sensor check for the Emergency Condenser Vent Noble Gas
Activity Monitors to include, in addition to power operations, the
condition when reactor coolant temperature is greater than 212
deg.F, except for hydrostatic testing. These changes will make the
conditions for Emergency Condenser Vent Noble Gas Activity Monitor
operability and daily sensor check surveillance performance
consistent with the conditions for ECS [emergency cooling system]
operability as indicated in LCO [Limiting Condition for Operation]
3.1.3.a.
The proposed changes to the Emergency Condenser Vent Noble Gas
Activity Monitor operability and daily sensor check surveillance
requirements will continue to provide assurance that the intent of
the effluent monitoring requirements of 10 CFR 50 Appendix A, GDC
[General Design Criterion] 64, is satisfied and the radiological
effluents are maintained within the dose and dose rate limits
specified in 10 CFR 50 Appendix I, 10 CFR 20, and the RETS
[Radiological Effluent Technical Specifications]. The proposed
changes will not effect the capability of the ECS to mitigate the
consequences of an accident that results in a loss of feedwater or
reactor isolation from the primary heat sink and aid the Core Spray
System and Automatic Depressurization System in providing effective
core cooling following non-limiting small breaks.
The proposed changes also correct a clerical error in the
Uranium Fuel Cycle effluent monitoring SR [surveillance
requirement]. The proposed correction simply restores the SR to the
form that existed before the error was introduced. The clerical
error did not affect the ODCM [Offsite Dose Calculation Manual]
implementing procedures or plant operation. Thus, the cumulative
dose contribution from Uranium Fuel Cycle sources will continue to
be maintained within the limits of 40 CFR 190 and the RETS.
Based on the above analysis, the proposed changes do not result
in any hardware changes or physical alteration of the plant, and the
changes will have no impact on the design or function of any
structure, system or component (SSC). As such, the SSC process
variables, characteristics, and functional performance will be
maintained consistent with the event initiator and the initial
condition assumptions for the accident analyses. Moreover, the
proposed changes will not eliminate any actions or adversely affect
any SSCs required to prevent accidents or mitigate accident
conditions, nor will the changes result in the degradation of any
fission product barriers so as to increase the radiological
consequences of an accident. It is, therefore, concluded that
operation in accordance with the proposed amendment will not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The operation of Nine Mile Point Unit 1, in accordance with
the proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not result in any hardware changes or
physical alteration of the plant, and the changes do not impact the
design or function of any SSC. The proposed changes maintain the
capability of the ECS to respond to accidents, including non-
limiting small breaks, consistent with the current analyses. In
addition, the proposed changes provide continued assurance that the
radiological dose and dose rates will be maintained within limits.
The proposed changes do not alter the process variables,
characteristics, or functional performance of any SSC, do not
eliminate any requirements, and do not impose any new requirements
which could introduce new equipment failure modes or create new
credible accidents. It is, therefore, concluded that operation in
accordance with proposed amendment 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 1, in accordance with
the proposed

[[Page 6700]]

amendment, will not involve a significant reduction in a margin of
safety.
The proposed changes do not affect the capability of the ECS to
mitigate consequences of an accident that results in a loss of
feedwater or reactor isolation from the primary heat sink, or affect
the capability of the ECS to aid the Core Spray System and the
Automatic Depressurization System in providing effective core
cooling following non-limiting small breaks. Thus, there will be no
impact on the post-accident radioactive material release analyses or
a reduction in the margin to the associated 10 CFR 100 dose limits.
In addition, the proposed changes provide continued assurance that
the intent of the effluent monitoring requirements of 10 CFR 50
Appendix A, GDC 64, is satisfied and the dose and dose rates due to
the radiological effluents are maintained within the limits
specified in 10 CFR 50 Appendix I, 10 CFR 20, 40 CFR 190, and the
RETS. Moreover, the proposed changes do not eliminate any
requirements or responsibilities, nor impose new requirements or
responsibilities, or alter any physical parameters which could
reduce the margin to an acceptance limit. It is, therefore,
concluded that operation in accordance with the proposed amendment
will not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: S. Singh Bajwa, Director.

North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: December 16, 1998.
Description of amendment request: The proposed editorial and
administrative changes to the Technical Specifications would either
revise references and statements that are inaccurate or provide relief
from administrative controls which provide insignificant safety
benefit.
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 design basis accidents are not affected by the proposed
editorial and administrative changes. The proposed changes do not
change the level of programmatic controls or the procedural details
currently in place. The proposed changes do not revise the station
design, the response of the station to transients nor the manner in
which the station is operated, therefore, these changes have no
adverse affect to the safe operation of the station. The proposed
changes do not involve a significant increase in 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 previously analyzed.
The proposed changes do not alter the design assumptions,
conditions, configuration of the facility or the manner in which the
plant is operated. There are no changes to the source term,
containment isolation or radiological release assumptions used in
evaluating the radiological consequences in the Seabrook Station
UFSAR. Existing system and component redundancy is not being changed
by the proposed changes. The proposed changes have no adverse affect
on component or system interactions. The proposed changes are
editorial and administrative in nature and do not change the level
of programmatic controls and procedural details associated with the
aforementioned technical specifications. Therefore, since there are
no changes to the design assumptions, conditions, configuration of
the facility, or the manner in which the plant is operated and
surveilled, the proposed changes do not create the possibility of a
new or different kind of accident from any previously analyzed.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
There are no changes being made to the Technical Specification
safety limits or safety system settings that would adversely affect
plant safety. The changes do not affect the operation of structures,
systems or components nor do they introduce administrative changes
to plant procedures that could affect operator response during
normal, abnormal or emergency situations. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: William M. Dean.

North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: December 16, 1998.
Description of amendment request: The proposed change would
relocate Technical Specifications (TS) 3/4.7.10, ``Area Temperature
Monitoring,'' and associated TS Table 3.7-3, to the Seabrook Station
Technical Requirements Manual.
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 change does not adversely affect accident
initiators or precursors nor alter the design assumptions,
conditions, configuration of the facility or the manner in which the
plant is operated. The proposed change does not alter or prevent the
ability of structures, systems, or components (SSCs) to perform
their intended function to mitigate the consequences of an
initiating event within the acceptance limits assumed in the Updated
Final Safety Analysis Report (UFSAR). The proposed change is
administrative in nature and does not decrease the effectiveness of
programmatic controls or the procedural details of assuring
operation of the facility in a safe manner.
The provisions of TS 3/4.7.10 for area temperature monitoring of
the referenced selected areas is neither part of an initial
condition of a design basis accident or transient that either
assumes the failure of or presents a challenge to the integrity of a
fission product barrier, nor is area temperature monitoring relied
upon as a primary success path to mitigate such events. The
provisions for area temperature monitoring is not related to events
that are considered frequent or dominant contributors to plant risk.
Area temperature monitoring is not considered a design feature or an
operating restriction that is an initial condition of a design basis
accident or transient analysis, nor does it provide a function or
actuate any accident mitigation feature in order to mitigate the
consequences of a design basis accident or transient.
Relocating TS 3/4.7.10 to the Technical Requirements Manual will
still provide adequate controls for area temperature in those areas
designated in TS Table 3.7-3. The relocated requirements of TS 3/
4.7.10 to the Technical Requirements Manual will continue to be
administratively controlled in accordance with TS Section 6.0,
``Administrative Controls.''
The Seabrook Station Technical Requirements Manual is a
licensee-controlled

[[Page 6701]]

document which contains certain technical requirements and is the
implementing manual for the Technical Specification Improvement
Program. Changes to these requirements are reviewed and approved in
accordance with Seabrook Station Technical Specifications, Section
6.7, and as outlined in the Technical Requirements Manual.
Specifically, changes to the Technical Requirements require a 10 CFR
50.59 safety evaluation and are reviewed and approved by the Station
Operations Review Committee (SORC) and the Nuclear Safety Audit
Review Committee (NSARC) prior to implementation.
The proposed change will not degrade the ability of systems,
structures and components important to safety to perform their
safety function. The proposed change will not change the response of
any system, structure or component important to safety as described
in the Seabrook Station Updated Final Safety Analysis Report
(UFSAR). Since the plant response to an accident will not change,
there is no change in the potential for an increase in the
consequences of an accident previously analyzed. As such, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously analyzed.
The proposed change does not alter the design assumptions,
conditions, configuration of the facility or the manner in which the
plant is operated. There are no changes to the source term,
containment isolation or radiological release assumptions used in
evaluating the radiological consequences in the Seabrook Station
UFSAR. Existing system and component redundancy is not being changed
by the proposed change. The proposed change has no adverse impact on
component or system interactions. The proposed change will not
adversely degrade the ability of systems, structures and components
important to safety to perform their safety function nor change the
response of any system, structure or component important to safety
as described in the Seabrook Station Updated Final Safety Analysis
Report (UFSAR). The proposed change is administrative in nature and
does not change the level of programmatic controls and procedural
details controls of assuring operation of the facility in a safe
manner. Therefore, since there are no changes to the design
assumptions, conditions, configuration of the facility, or the
manner in which the plant is operated and surveilled, the proposed
change does not create the possibility of a new or different kind of
accident from any previously analyzed.
Future changes to area temperature monitoring requirements will
be reviewed and approved in accordance with Seabrook Station
Technical Specifications, Section 6.7, and as outlined in the
Technical Requirements Manual. Specifically, changes to the
Technical Requirements require a 10 CFR 50.59 safety evaluation and
are reviewed and approved by the Station Operations Review Committee
(SORC) and the Nuclear Safety Audit Review Committee (NSARC) prior
to implementation.
Since the plant response to an accident will not change, there
is no change in the potential for an increase in the consequences of
an accident previously analyzed, nor can it create the possibility
of a new or different kind of accident from any previously
evaluated.
Relocation of the area temperature monitoring requirements to
the Technical Requirements Manual will not create the possibility of
a new or different kind of accident from any previously analyzed.
3. The proposed change does not involve a significant reduction
in a margin of safety.
There is no adverse impact on equipment design or operation and
there are no changes being made to the Technical Specification
required safety limits or safety system settings that would
adversely affect plant safety. The proposed change is administrative
in nature and does not change the level of programmatic controls and
procedural details associated with area temperature monitoring to
ensure that environmentally qualified equipment will not be exposed
to temperatures beyond that which they were originally qualified.
Future changes to the area temperature monitoring requirements
will be reviewed and approved in accordance with Seabrook Station
Technical Specifications, Section 6.7, and as outlined in the
Technical Requirements Manual. Specifically, changes to the
Technical Requirements require a 10 CFR 50.59 safety evaluation and
are reviewed and approved by the Station Operations Review Committee
(SORC) and the Nuclear Safety Audit Review Committee (NSARC) prior
to implementation.
Relocation of the requirements contained in TS 3/4.7.10 to the
Technical Requirements Manual does not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Exeter Public Library,
Founders Park, Exeter, NH 03833.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), Et Al., Docket No. 50-
336, Millstone Nuclear Power Station, Unit No. 2, New London
County, Connecticut

Date of amendment request: December 28, 1998.
Description of amendment request: NNECO is proposing to change
Technical Specification 2.2.1, ``Limiting Safety System Settings--
Reactor Trip Setpoints,'' and the associated Bases to reflect revised
loss of normal feedwater (LONF) analyses. An additional Technical
Specification Bases change to the floor value for the thermal margin
low pressure reactor trip is also included. This proposed change is not
related to the revised LONF analyses.
NNECO is also seeking NRC approval to incorporate changes to the
Millstone Unit No. 2 Final Safety Analysis Report (FSAR). The proposed
changes to the FSAR, except the floor value for thermal margin low
pressure reactor trip, are associated with the revised LONF analyses.
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:

In accordance with 10 CFR 50.92, NNECO has reviewed the proposed
changes and has concluded that they do not involve a Significant
Hazards Consideration (SHC). The basis for this conclusion is that
the three criteria of 10 CFR 50.92(c) are not compromised. The
proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The analysis of a loss of normal feedwater (LONF) event, as
described in the Millstone Unit No. 2 FSAR Chapters 10 and 14, has
been revised. Certain key assumptions have been changed to ensure
acceptable analysis results. An evaluation of the LONF analyses
changes, and associated Technical Specification changes will be
presented. In addition, an evaluation of an additional non LONF
analyses related Technical Specification Bases and FSAR change is
included.
LONF analyses changes. The LONF analyses, contained in FSAR
Chapters 10 and 14, have been revised using a steam generator liquid
inventory assumption, at the time of reactor trip on low steam
generator water level, that is consistent with the design of the
replacement steam generators. The revised Chapter 10 and 14 LONF
analyses also incorporate a reduction in auxiliary feedwater
delivery rates resulting from a recalculation of the Auxiliary
Feedwater (AFW) System flows. The results of revised analyses
indicate that the analytical limit for the low steam generator water
level reactor trip must be raised to 43% narrow range level from the
current 34% narrow range level. This will result in a change to the
low steam generator water level reactor trip setpoint listed in
Technical Specification 2.2.1.
The revised Chapter 14 LONF analysis will now take credit for
automatic initiation of the motor driven auxiliary feedwater (MDAFW)
pumps. The current Chapter 14 LONF analysis assumes auxiliary
feedwater flow will be initiated 10 minutes after the event. The
Chapter 10 LONF analysis assumption of

[[Page 6702]]

automatic initiation of one MDAFW pump within 4 minutes, after the
low steam generator level AFW actuation setpoint is reached, has not
changed.
To demonstrate that one MDAFW pump delivers sufficient flow to
preclude steam generator dryout, the Chapter 10 LONF analysis will
not take credit for the operation of the steam generator atmospheric
dump valves, instead of the main steam safety valves as in the
current analysis. This new assumption yields lower predicted steam
generator pressures which result in an increase in the delivered AFW
flows.
LONF analyses related technical specification changes. The trip
setpoint and allowable value for the low steam generator water level
reactor trip will be changed to be consistent with the revised LONF
analyses. The revised analyses assume an analytical limit of 43%
narrow range level, instead of the current analytical limit of 34%
narrow range level. The calculation of the trip setpoint, which
includes instrument uncertainty, has determined that the trip
setpoint should be changed from [greater than or equal to] 36.0% to
[greater than or equal to] 48.5%.
The increase in the low steam generator level Reactor Protection
System (RPS) actuation setpoint from [greater than or equal to] 36%
to [greater than or equal to] 48.5% will result in an increase in
the probability of an RPS actuation on low steam generator water
level since the difference between the proposed setpoint and the
normal operating value of steam generator level will decrease. The
proposed actuation setpoint is below the normal operating level of
60 to 75%. Steam generator level is not expected to approach the
actuation setpoint during normal operation. An unexpected plant
event (e.g., loss of main feedwater or difficulty controlling steam
generator level at low power levels) would be necessary for steam
generator level to approach the actuation setpoint. To provide the
operators with advance notice of the steam generator low level
condition, the existing RPS low steam generator water level pretrip
alarm setpoint will be changed to provide approximately the same
margin between pretrip and trip as currently exists (5%). This will
ensure that the pretrip alarm is received prior to reaching the
actual record trip setpoint. Therefore, even though the proposed
change will decrease the margin between the normal operating steam
generator level and the RPS actuation setpoint, this change will not
significantly impact the probability of an RPS actuation on low
steam generator level during normal plant operations. In addition,
the proposed setpoint and allowable value change will ensure a
reactor trip signal is generated at, or before the analytical limit
used in the revised LONF analysis is reached. Therefore, the RPS
will continue to function as designed to mitigate the consequences
of the design basis accidents.
The basis for the steam generator level low reactor trip will be
modified to be consistent with the revised LONF analyses. The
discussion concerning available water inventory and time until
auxiliary feedwater is required will be removed. The proposed change
to the FSAR will include a discussion of the relationship between
the LONF analysis and the need to automatically initiate auxiliary
feedwater flow.
Non LONF analyses related technical specification bases and FSAR
change. This Technical Specification Bases and FSAR change is not
related to the revised LONF analyses.
The basis for the thermal margin low pressure (TMLP) reactor
trip (Technical Specification 2.2.1 Bases) will be modified. The
current basis states that the floor, or minimum value, for this trip
function is set at 1850 psia pounds per square inch absolute]. This
value will be changed to be consistent with instrument uncertainty
calculations that have determined that the floor should be increased
to 1865 psia. The increase in floor value is the result of greater
instrument uncertainties when harsh containment environment
conditions are included.
The increase in the TMLP floor (from 1850 psia to 1865 psia)
could result in an increase in the probability of an RPS actuation
on thermal margin low pressure since the difference between the
proposed floor setpoint and the normal operating value of
pressurizer pressure will decrease. However, the proposed actuation
setpoint is significantly below the normal operating pressure of
approximately 2250 psia. Pressurizer pressure is not expected to
approach the actuation setpoint during normal operation. A
significant plant event (e.g., loss of primary coolant) would be
necessary for a rapid pressure excursion to approach the actuation
setpoint. Since the setpoint change is small, it will not adversely
impact the probability of an RPS actuation on low pressurizer
pressure during normal plant operations. In addition the proposed
change to the floor value will ensure a reactor trip signal is
generated at, or before the analytical limit used in the respective
accident analyses is reached. Therefore, the RPS will continue to
function as designed to mitigate the consequences of the design
basis accidents.
Conclusion. The results of the revised LONF analyses contained
in FSAR Chapters 10 and 14 have concluded that the LONF event does
not result in the violation of the Specified Acceptable Fuel Design
Limits, that the peak pressurizer pressure does not exceed 110% of
the design pressure, that liquid primary coolant is not expelled
through the pressurizer safety valves, and that adequate cooling
water is supplied by the AFW System to prevent steam generator
dryout and allow a safe and orderly plant shutdown. By preventing
steam generator dryout, sufficient removal of decay heat from the
Reactor Coolant System (RCS) will occur, preventing excessive RCS
heatup and pressurization. This will ensure the steam generator
fatigue analysis remains valid, and excessive discharge of primary
coolant through the pressurizer safety valves does not occur.
Therefore, there will be no adverse effect on the consequences of a
LONF event. This is consistent with the acceptance criteria
contained in Standard Review Plan (SRP) 15.2.7, [``Loss of Normal
Feedwater Flow,'' Rev. 1--July 1981]. (Millstone Unit No. 2 is not
an SRP plant.)
The proposed changes do not alter the way any structure, system,
or component functions. The changes in actuation setpoints and
equipment used in the LONF analyses affect equipment important to
the mitigation of design basis accidents. These changes do not
affect any equipment that can cause a design basis accident to
occur. Therefore, the proposed changes do not affect the probability
of occurrence of a previously evaluated accident.
These proposed changes do not alter the way any structure,
system, or component functions. There will be no adverse effect on
any design basis accident previously evaluated, on any equipment
important to safety, or on the radiological consequences of any
design basis accident. Therefore, these proposed changes will not
adversely affect the consequences of a previously evaluated
accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Results of the proposed LONF analyses have demonstrated that the
Specified Acceptable Fuel Design Limits are not violated, that the
peak pressurizer and steam generator pressures do not exceed 110% of
the design pressure, that liquid primary coolant is not expelled
through the pressurizer safety valves, and that adequate cooling
water is supplied by the AFW System to prevent steam generator
dryout and allow a safe and orderly plant shutdown. Therefore, there
are no new or different types of failures of systems or equipment
important to safety which could cause a new or different type of
accident from any accident previously evaluated.
The proposed changes will 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 revised FSAR Chapter 14 analysis has concluded that the
steam generator low water level reactor trip setpoint does not
provide sufficient water inventory in the steam generators at the
time of the reactor trip such that auxiliary feedwater flow will not
be required for 10 minutes. This contradicts the current Technical
Specification Basis (Technical Specification 2.2.1) for the steam
generator low water level reactor trip setpoint. Therefore, the
revised analysis reduces the margin of safety as defined in the
Bases of the Millstone Unit No. 2 Technical Specifications. However,
with the proposed changes to increase the low steam generator water
level reactor trip setpoint and taking credit for automatic AFW
System actuation, it has been shown that operation of these systems
can mitigate the LONF event, and ensure plant response is within the
acceptance criteria. Results of the proposed LONF analyses have
demonstrated that the Specified Acceptable Fuel Design Limits are
not violated, that the peak pressurizer and

[[Page 6703]]

steam generator pressures do not exceed 110% of the design pressure,
that liquid primary coolant is not expelled through the pressurizer
safety valves, and that adequate cooling water is supplied by the
AFW System to prevent steam generator dryout and allow a safe and
orderly plant shutdown. Therefore, these proposed changes do not
involve a significant reduction in a margin of safety.
The proposed change to the floor value for the TMLP reactor trip
function is the result of a revision to the instrument loop
uncertainty and setpoint calculations. The proposed change to the
Technical Specification Basis will incorporate the RPS TMLP floor
setpoint change. This change to the TMLP floor will not adversely
affect this function. The TMLP reactor trip function will still
operate as designed. The RPS will continue to function as designed
to mitigate the consequences of design basis accidents. Therefore,
this proposed change does not involve a significant reduction in a
margin of safety.
The NRC has provided guidance concerning the application of
standards in 10 CFR 50.92 by providing certain examples (March 6,
1986, 51 FR 7751) of amendments that are considered not likely to
involve an SHC. The changes proposed herein are not enveloped by any
specific example.
As described above, this License Amendment Request does not
impact the probability of an accident previously evaluated, does not
involve a significant increase in the consequences of an accident
previously evaluated, does not create the possibility of a new or
different kind of accident from any accident previously evaluated,
and does not result in a significant reduction in a margin of
safety. Therefore, NNECO has concluded that the proposed changes do
not involve an SHC.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), Et Al., Docket No. 50-
336, Millstone Nuclear Power Station, Unit No. 2, New London
County, Connecticut

Date of amendment request: January 18, 1999.
Description of amendment request: NNECO is proposing to change
Technical Specification 3.6.1.2, ``Containment Systems--Containment
Leakage.'' The Bases for this Technical Specification and the Final
Safety Analysis Report (FSAR) will also be modified to address the
proposed changes.
The limit for secondary containment bypass leakage specified in
Technical Specification 3.6.1.2.c will be reduced from less than 0.017
La to less than 0.0072 La. This new limit is
consistent with the value of secondary containment bypass leakage used
in the revised off-site and control room dose calculations following a
design basis loss-of-coolant accident (LOCA).
Technical Specification 3.6.1.2.c will be modified by replacing
``identified in Table 3.6-1 as'' with ``that are.'' This will allow
Table 3.6-1 to be removed. The removal of this table from Technical
Specifications and the proposed wording change are consistent with the
guidance contained in Generic Letter (GL) 91-08. It is not necessary to
maintain a list of the secondary containment bypass leakage paths in
Technical Specifications. The Millstone Unit No. 2 FSAR (Section 5.3.4)
provides the necessary information to determine the secondary
containment bypass leakage paths that must be considered to ensure that
the combined leakage rate limit contained in Technical Specification
3.6.1.2.c is met.
Technical Specification 3.6.1.2 Table 3.6-1, ``Secondary
Containment Bypass Leakage Paths,'' will be removed and the phrase
``This Page Intentionally Deleted'' will be added to Page 3/4 6-5.
The Bases for Technical Specification 3.6.1.2 will be modified to
indicate that the Millstone Unit No. 2 FSAR contains a list of the
containment penetrations that have been identified as secondary
containment bypass leakage paths.
FSAR Section 5.3.4, ``Through-Line Leakage Evaluation,'' will be
changed to include the additional secondary containment bypass leakage
paths that have been identified. The criteria used to determine the
secondary containment bypass leakage paths will be modified to be
consistent with the criteria used in the evaluation that identified the
additional leakage paths.
The discussion of the use of a leakage rate of 11 cc/hr for the
control room dose calculations will be modified. The revised control
room dose calculations will assume a total secondary containment bypass
leakage rate consistent with the proposed change to Technical
Specification 3.6.1.2.
As a result of these proposed changes, the calculated off-site and
control room doses following a design basis LOCA will change. The
calculated doses are specified in FSAR Section 14.8.4, ``Radiological
Consequences of the Design Basis Accident.'' A revision to this section
of the FSAR has been submitted to the NRC by the letter dated September
28, 1998. This submittal will be revised to incorporate the proposed
total secondary containment bypass leakage rate and the associated
change to the calculated off-site and control room doses following a
design basis LOCA.
Basis for proposed no significant hazards consideration
determination: As required by 10 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration,
which is presented below:

In accordance with 10 CFR 50.92, NNECO has reviewed the proposed
changes and has concluded that they do not involve a significant
hazards consideration (SHC). The basis for this conclusion is that
the three criteria of 10 CFR 50.92(c) are not compromised. The
proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to lower the limit for secondary containment
bypass leakage, as specified in Technical Specification 3.6.1.2.c,
from [less than] 0.017 La to [less than] 0.072
La will reduce the off-site doses associated with the
design basis LOCA. The proposed change to raise the limit for
secondary containment bypass leakage from 11 cc/hr to [less than]
0.0072 La will increase the dose to the Control Room
Operators following a design basis LOCA. However, the revised off-
site and control room dose calculations, using the proposed combined
secondary containment bypass leakage limit, demonstrate that the
limits of 10 CFR 100 and 10 CFR 50, Appendix A, General Design
Criteria (GDC) 19 are met. In addition, these proposed changes will
result in the use of the same limit for secondary containment bypass
leakage when determining the radiological consequences of a design
basis LOCA.
The proposed wording change to Technical Specification
3.6.1.2.c, and the associated removal of Table 3.6-1, will not
change the requirement to verify total secondary containment bypass
leakage is within the limit assumed in the determination of the
radiological consequences of the design basis LOCA. Control of the
penetrations that have been identified as secondary containment
bypass leakage paths will be maintained by the process used to
change the Millstone Unit No. 2 FSAR. This process ensures that
appropriate changes to the FSAR are evaluated in accordance with 10
CFR 50.59 to determine if NRC approval is required prior to
implementing the change. This process also ensures that the NRC is
informed of FSAR changes via regular

[[Page 6704]]

updates to the FSAR. The removal of Table 3.6-1 from Technical
Specifications and the proposed wording change are consistent with
the guidance contained in GL 91-08.
The identification and addition of more secondary containment
bypass leakage paths to the FSAR will have no impact on the
calculated off-site and control room doses following a design basis
LOCA since the combined leakage through all secondary containment
bypass leakage paths is limited to the proposed value contained in
Technical Specification 3.6.1.2. The addition of bypass leakage
paths does not change the combined leakage limit, which is now used
in the off-site and control room dose calculations.
The Bases for Technical Specification 3.6.1.2 will be modified
to indicate that the Millstone Unit No. 2 FSAR contains a list of
the containment penetrations that have been identified as secondary
containment bypass leakage paths.
The proposed changes do not alter the way any structure, system,
or component functions. These changes do not affect any equipment
that can cause a design basis accident to occur. There will be no
adverse effect on any design basis accident previously evaluated or
on any equipment important to safety. The reduction in the allowable
secondary containment bypass leakage limit will result in a decrease
in the calculated off-site doses associated with the design basis
LOCA. The use of the proposed secondary containment bypass leakage
limit will increase the calculated doses to the Control Room
Operators following a design basis LOCA. However, the calculated
doses meet the criteria of 10 CFR 100 and GDC 19. Therefore, there
will be no significant increase in the probability or consequences
of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes will 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. Also, the response of the plant and
the operators following these accidents is essentially unaffected by
the change. The criteria used by the plant operators to terminate
containment spray following a design basis LOCA will change from
containment pressure to either time or pressure, whichever requires
longer operation. This will ensure that containment spray remains in
operation long enough to achieve the assumed iodine decontamination.
However, the operator action to terminate containment spray will
remain the same. 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 lower the Technical Specification limit
for secondary containment bypass leakage, to remove Table 3.6-1, and
to add more secondary containment bypass leakage paths to the FSAR
will have no adverse effect on equipment important to safety. The
equipment will continue to function as assumed in the design basis
accident analysis. These changes will ensure that the secondary
containment bypass leakage paths are identified and tested to verify
that the total secondary containment bypass leakage does not exceed
the Technical Specification limit. This will ensure that the
expected off-site and control room doses following a design basis
LOCA are within the limits specified in 10 CFR 100 and GDC 19.
Therefore, there will be no significant reduction in the margin of
safety as defined in the Bases for the Technical Specification
affected by these 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: 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 Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO) Et Al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut

Date of amendment request: January 18, 1999.
Description of amendment request: The proposed changes will remove
the Technical Specification related to Hydrogen Purge System from the
Millstone Unit No. 2 Technical Specifications. The proposed changes
affect Technical Specifications 3/4.6.4.3, ``Containment Systems,
Hydrogen Purge System.'' The Bases of the associated Technical
Specification will be modified to address the proposed changes. The
proposed changes will allow the licensee to downgrade the hydrogen
purge system to a non-safety-related system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

In accordance with 10 CFR 50.92, NNECO has reviewed the proposed
changes and has concluded that they do not involve a Significant
Hazards Consideration (SHC). The basis for this conclusion is that
the three criteria of 10 CFR 50.92(c) are not compromised. The
proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The Hydrogen Purge System provides a backup means to manually
control the hydrogen concentration in containment given the multiple
failure of the redundant, Seismic Category I Hydrogen Recombiner
System. The primary success path for hydr9gen control is the
Hydrogen Recombiner System. The Hydrogen Recombiner System has
redundant trains and is fully qualified to maintain hydrogen control
following a design basis accident. FSAR [Final Safety Analysis
Report] Section 14.8.3.5, ``Radiological Consequences of Purging''
is being removed from the FSAR since it is no longer required. Since
the hydrogen recombiners are fully redundant, it is not necessary to
postulate offsite doses for purge during a design basis accident.
Thus, the deletion of consequences does not represent a change in
the consequences of a design basis event. Therefore, this change
will not significantly increase the probability or consequences of
an accident previously evaluated.
Revision of Index Page VII is an administrative change. The
proposed change to Bases section 3/4.6.4 by deleting reference to
``the purge system'' is required since Technical Specification 3/
4.6.4.3 is being removed. Therefore, these changes 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 purge system is a standby purge system which is not in
service during normal operations as a hydrogen purge system (i.e.,
Charcoal Filter Heaters de-energized). Therefore, no new accident is
created either by system unavailability or actuation. The FSAR will
still address the use of the purge system as a backup to the
recombiner system, Revision of Index Page VII is an administrative
change. The proposed change to Bases section 3/4.6.4 by deleting
reference to ``the purge system'' is required since Technical
Specification 3/4.6.4.3 is being removed. Therefore, the proposed
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.

[[Page 6705]]

3. Involve a significant reduction in a margin of safety.
The margin of safety is defined in the Bases 3/4.6.4 which
states that the ``hydrogen control systems are consistent with the
recommendations of Regulatory Guide 1.7 * * * ''. Regulatory Guide
1.7 describes methods that would be acceptable in meeting the
standards for a combustible gas control system, 10 CFR 50.44,
``Standards for combustible gas control systems in light-water-
cooled power reactors.'' Regulatory Guide 1.7 acknowledges that
purging is a means of reducing the hydrogen concentration but it
should not be the primary means because of the release of
radioactivity to the environment. The regulatory guide does advise
that there be an ``installed capability for a controlled purge of
the containment atmosphere to aid in cleanup.'' Removal of the
Hydrogen Purge System Technical Specification is consistent with
Regulatory Guide 1.7. Additionally, the capability to purge is still
documented in the FSAR. Revision of Index Page VII is an
administrative change. The proposed change to Bases section 3/4.6.4
by deleting reference to ``the purge system'' is required since
Technical Specification 3/4.6.4.3 is being removed. Therefore, the
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 proposed 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 Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), Et Al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London
County, Connecticut

Date of amendment request: January 18, 1999.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) 3/4.2.2 to be in accordance with
NRC-approved Westinghouse methodologies for the heat flux hot channel
factor--FQ(Z). In addition, the proposed amendment would
make changes to the core operating limits and the analytical methods
used to determine core operating limits contained in Section 6.9.1.6.a
and b, respectively, by adding, modifying, or deleting references.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no signification hazards
consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with 10
CFR 50.92 and has concluded that the revision does not involve any
Significant Hazards Considerations (SHC). The basis for this
conclusion is that the three criteria of 10 CFR 50.92(c) are not
satisfied. The proposed Technical Specification revision does not
involve an SHC because the revision would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
To determine any potential impact, the proposed changes to the
TS are grouped into the following two categories.
(a) Changes to Technical Specification 3/4.2.2 ``Heat Flux Hot
Channel Factor--FQ(Z)''
(b) Changes that are not related to the Heat Flux Hot Channel
Factor TS, and are administrative in nature. These include defining
a new core operating limit and deleting, re-numbering, updating and
adding references to analytical methods used to determine core
operating limits in TS 6.9.1.6 ``Core Operating Limit Report
(COLR).['']
With respect to item 1.a changes related to the Heat Flux Hot
Channel Factor, FQ(Z), impact the initial conditions
assumed in the accidents analyzed for MP3 [Millstone Unit 3]. These
initial conditions are power distributions which are consistent with
reactor operation as defined in the TS. The proposed changes to the
Heat Flux Hot Channel Factor TS ensure that proper actions are taken
to maintain peaking factors within the limits assumed in the MP3
accident analysis. The proposed changes are consistent with the NRC
approved Westinghouse methodology for FQ(Z) surveillance.
Changes to the SURVEILLANCE and ACTION statements will not change
the probability of occurrence of any analyzed accidents.
Furthermore, the consequences of analyzed accidents will not change
since the power distribution assumptions will not be challenged by
reactor operation allowed by the Technical Specifications.
With respect to item 1.b the administrative changes to the
Technical Specifications do not affect existing or proposed Limiting
Conditions for Operation (LCO) or SURVEILLANCE REQUIREMENTS.
Therefore, there is no impact on the design basis accidents.
Thus it is concluded that the proposed revision does not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
(a) Proposed changes to the Heat Flux Hot Channel Factor, TS 3/
4.2.2 ensure that proper actions are taken to maintain peaking
factors within the limits assumed in the MP3 accident analysis. The
proposed changes are consistent with the NRC approved Westinghouse
methodology for FQ(Z) surveillance. Maintaining safety
analysis assumptions on power distributions cannot be an initiating
event for any design basis accidents and will not create the
possibility of a different type of accident. Therefore the changes
associated with the Heat Flux Hot Channel Factor limiting condition
for operation do not represent a new unanalyzed accident.
(b) Since the administrative changes do not affect plant
operation, the potential for an unanalyzed accident is not created.
No new failure modes are introduced.
Thus, this proposed revision does not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Involve a significant reduction on the margin of safety.
(a) The proposed changes ensure that FQ(Z), will
remain within the safety analysis assumptions. The LCO limits and
SURVEILLANCE REQUIREMENTS are not altered. Therefore, the impact on
the consequences on the protective boundaries is unchanged. Meeting
the intent of the NRC approved Westinghouse methodology for
FQ(Z), SURVEILLANCE ensures that power distributions
assumed in the accident analysis will not be challenged by reactor
operations allowed by the Technical Specifications. Therefore,
verification of no change in the margin of safety is encompassed by
meeting the power distribution limits assumed in analyzed accidents.
(b) Since the proposed changes do not affect the consequences of
any accident previously analyzed, there is no reduction in the
margin of safety.
Thus it is concluded that the proposed revision does not involve
a significant reduction in the margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve a Significant
Hazard 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: 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.

[[Page 6706]]

NRC Project Director: William M. Dean.

Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: December 31, 1998.
Description of amendment request: The proposed amendment would
revise the technical specification (TS) reactor pressure vessel (RPV)
pressure-temperature (P-T) limit curves, delete completed RPV sample
surveillance requirements, delete requirement to withdraw a specimen at
next refueling outage, and remove the standby liquid control system
(SBLC) relief valve setpoint. Associated administrative changes are
also proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
RPV P-T curve changes. It is proposed that P-T curves be revised
to accommodate the shift in RTNDT determined using actual
surveillance program data rather than generic data provided in
Regulatory Guide [RG] 1.99 Revision 2 (Radiation Embrittlement of
Reactor Vessel Materials). The new P-T curves will increase the
margins provided in the P-T limit curves against non-ductile failure
of the RPV. Regulatory Guide 1.99 Revision 2 encourages use of plant
specific surveillance data as data becomes available.
Eliminating prescriptive requirements to remove a RPV test
specimen sample at three fourths service life will result in an
overall improvement in the RPV surveillance program since the
limited number of remaining surveillance samples will be removed at
optimum intervals. Therefore, proposed changes will neither
significantly increase the probability or the consequences of an
accident previously evaluated.
RPV surveillance requirements. Deleting completed, one time
surveillance requirements [SRs] of SR section 4.6.B and
incorporating a discussion of the results in the Bases is an
administrative change and has no effect on probability or
consequences of accidents.
SBLC relief valve setpoint testing. The testing requirements of
TS section 4.4.A.2.c are enveloped by the current testing performed
by Monticello's IST [inservice test] Program, which implements ASME
[American Society of Mechanical Engineers] Code Section XI, approved
by 10 CFR 50.55a. The IST program requires all relief valves to be
tested to their nameplate data setpoints. Any modification to a
relief valve's nameplate data is controlled by the plant's
configuration control process which would ensure the requirements of
ASME Section XI are invoked as required by TS section 3.15. The IST
program required by TS 4.15 ensures the SBLC relief valves will be
properly tested for operability. Therefore, revising section
4.4.A.2.c to remove specific setpoints does not increase the
probability or consequences of an accident.
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
RPV P-T curve change. Updated RPV P-T limit curves will not
create the possibility of a new or different kind of accident nor
alter operational standards. New limits continue a system of
operating bounds which are in place to prevent damage to reactor
vessels during normal operating conditions including hydrostatic
pressure and leakage testing, and anticipated transients. The
updated P-T curves incorporate the results of RPV surveillance
specimen testing utilizing criteria defined in RG 1.99, Revision 2.
No change is being made to the way the P-T limits provide plant
protection. No new modes of operation are involved. The changes do
not necessitate physical alteration of the plant.
RPV surveillance requirements. Deleting completed, one time
surveillance requirements of section 4.6B and incorporating a
discussion of the results in the Bases is an administrative change
and therefore has no effect on previously analyzed accidents.
SBLC Relief Valve Setpoint Testing. The testing requirements of
TS section 4.4.A.2.c are enveloped by the current testing performed
by Monticello's IST Program, which implements ASME Code Section XI,
approved by 10 CFR 50.55a. The IST program requires all relief
valves to be tested to their nameplate data setpoints. Any
modification to a relief valve's nameplate data is controlled by the
plant's configuration control process which would ensure the
requirements of ASME Section XI are invoked as required by TS
section 3.15. The IST program required by TS 4.15 ensures the SBLC
relief valves will be properly tested for operability. Therefore,
revising section 4.4.A.2.c to remove specific setpoints does not
create the possibility of a new or different kind of accident, from
any accident previously analyzed.
The proposed amendment will not involve a significant reduction
in the margin of safety.
RPV P-T curve change. The proposed RPV P-T curve changes are
designed to maintain the recommended safety factors specified in the
ASME Boiler and Pressure Vessel Code, Section III, Appendix G, and
10 CFR Part 50, Appendix G. The revised curves are based on current
NRC guidelines utilizing actual RPV surveillance program tests
results. The proposed changes shift the curves in a slightly more
conservative direction thus maintaining or increasing the previous
margins of safety.
RPV surveillance requirements. Deleting completed, one time
surveillance requirements from Section 4.6.B and incorporating a
discussion of the results in the Bases is an administrative change
and has no effect on any margin of safety.
SBLC relief valve setpoint testing. The testing requirements of
TS section 4.4.A.2.c are enveloped by the current testing performed
by Monticello's IST Program, which implements ASME Code Section XI,
approved by 10 CFR 50.55a. The IST program requires all relief
valves to be tested to their nameplate data setpoints. Any
modification to a relief valve's nameplate data is controlled by the
plant's configuration control process which would ensure the
requirements of ASME Section XI are invoked as required by TS
section 3.15. The IST program required by TS 4.15 ensures the SBLC
relief valves will be properly tested for operability. Therefore,
revising section 4.4.A.2.c to remove specific setpoints will not
reduce 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: January 4, 1999.
Description of amendment request: PECO Energy Company (PECO Energy)
is requesting Technical Specifications (TS) changes which will revise
the Administrative Section of TS pertaining to controlled access to
High Radiation Areas, and the reporting dates for the Annual
Occupational Radiation Exposure Report and the Annual Radioactive
Effluent Release Report.
The specific TS changes are as follows:
TS Section 6.12, 6.12.1, and 6.12.2 will be changed to: clarify
requirements; incorporate additional monitoring options (to allow
dosimetry and video monitoring) for entry into high radiation areas;
add the requirement that all individuals entering a high radiation area
have knowledge of the dose rates in the area; and add the requirement
that locked high radiation controls apply to each individual entering
the area.
TS Sections 6.9.1.4, 6.9.1.5(a), and 6.9.1.8 will be changed to:
support changes to the NRC reporting dates;

[[Page 6707]]

reference 10 CFR 20.2206; delete current reporting dates, and correct a
typographical error.
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 TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The changes are administrative in nature and do not impact the
operation, physical configuration, or function of plant equipment or
systems. The changes do not impact the initiators or assumptions, of
analyzed events, nor do they impact mitigation of accidents on
transient events. Therefore, these changes do not increase the
probability of occurrence of consequences of an accident previously
evaluated in the SAR [Safety Analysis Report].
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature and do not
alter plant configuration, require that new equipment be installed,
alter assumptions made about accidents previously evaluated, or
impact the operation or function of plant equipment. Therefore,
these changes do not create the possibility of a new or different
kind of accident than previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed changes are administrative in nature and do not
impact any safety assumptions, or potentially reduce any margin of
safety as described in the LGS TS basis. The proposed changes have
no impact on any safety analysis assumptions. Therefore, these
changes do not involve any reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Project Director: William M. Dean.

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

Date of amendment request: December 28, 1998.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to permit an increase in the
allowable leak rate for the main steam isolation valves (MSIVs) and to
delete the MSIV Sealing System. The main steam drain lines and the main
condenser would be utilized as an alternate MSIV leakage treatment
method.
Basis for proposed no significant hazards consideration
determination: As required by 1

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A99-3098. Public record. Not legal advice.