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

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URL: https://www.frixlaw.com/law-library/documents/fr%3A99-5751

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** March 10, 1999
- **Citation:** 64 FR 11959

## Text

NUCLEAR REGULATORY COMMISSION

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

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.

[[Page 11960]]

This biweekly notice includes all notices of amendments issued, or
proposed to be issued from February 12, 1999, through February 26,
1999. The last biweekly notice was published on February 24, 1999 (FR
64 PR 9183).

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

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By April 9, 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 aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with

[[Page 11961]]

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

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: January 29, 1999.
Description of amendment request: The amendments would allow credit
for containment overpressure to assist in providing net positive
suction head (NPSH) for the emergency core cooling system pumps for a
period of greater than 8 hours. The current licensing basis recognizes
credit given only to 8 hours after a design-basis loss-of-coolant
accident and the licensee has determined this to be an unreviewed
safety question.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Does the change involve a significant increase in the
probability or consequences of an accident as previously evaluated?
The proposed amendment involves the available containment
overpressure (COP) following a design basis loss of coolant accident
(DBA-LOCA) and the resulting NPSH available to the RHR [residual
heat removal] and CS [core spray] pumps. While this change affects
the ability of these pumps to perform their required functions
following a DBA-LOCA, it does not affect the reactor recirculation
piping or the reactor coolant pressure boundary, which are the
initiators of the DBA-LOCA. Therefore, the proposed amendment does
not involve a significant increase in the probability of an accident
previously evaluated.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed for the analysis, the availability
and successful functioning of the equipment assumed to operate in
response to the analyzed event, and the set points at which these
actions are initiated. The proposed change permits limited COP to be
credited in the calculation of available NPSH for the RHR and CS
pumps following a DBA-LOCA.
The proposed change is supported by calculations, which
demonstrates that adequate COP will be available to ensure the RHR
and CS systems will be capable of performing their required safety
functions. Therefore, the proposed amendment does not involve a
significant increase in the consequences of an accident previously
evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed amendment permits limited COP to be credited in the
calculation of available NPSH for the RHR and CS pumps following a
DBA-LOCA. This amendment does not involve a physical alteration of
the plant. The proposed amendment is supported by calculations,
which demonstrate that adequate COP will be available to ensure the
RHR and CS systems will be capable of performing their required
safety functions. This amendment will not alter the manner in which
the RHR and CS systems are initiated, nor will the function demands
on the RHR or CS system be changed. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed amendment permits limited COP to be credited in the
calculation of available NPSH for the RHR and CS pumps following a
DBA-LOCA. Crediting an incremental amount of overpressure does not
result in a significant reduction in the margin of safety, because
conservative analyses demonstrate that adequate COP will be
available to ensure the RHR and CS systems will be capable of
performing their required safety functions. Therefore, the proposed
amendment does not involve a significant reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Project Director: Stuart A. Richards.

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

Date of amendment request: February 18, 1999.
Description of amendment request: The proposed amendments would
revise the joint Technical Specifications (TSs): (1) Surveillance
Requirement (SR) 3.6.16.1--This SR incorrectly characterizes the access
openings (there are five of them) to the reactor building as each
having a double-door design, when in reality there is a single door for
each opening; the proposed revision would change the wording to
correctly characterize the actual design. (2) SR 3.6.16.3--This SR
specifies that the reactor building structural integrity inspection be
performed every 40 months to 50 months and during shutdown; the
proposed revision would change this frequency to three times every 10
years coinciding with containment visual examinations required by SR
3.6.1.1. (3) Administrative Control 5.5.2--The proposed revision would
add wording to specify that containment visual examinations required by
Regulatory Guide c.3 will be conducted three times every 10 years
including during each shutdown for SR 3.6.1.1.
The proposed amendments would only revise the SRs and
Administrative Controls specified above; no physical change to any
plant design is involved.
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:

First Standard

Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Approval of this amendment will have no
significant effect on accident probabilities or consequences. The
containment and reactor building are not accident initiating systems
or structures; therefore, there will be no impact on any accident
probabilities by the approval of this amendment. The containment and
reactor buildings serve an important function to mitigate
consequences of postulated accidents previously evaluated and the
examination frequencies proposed in this amendment will not result
in a reduction in

[[Page 11962]]

their capacity to meet their intended function. Therefore, there
will be no impact on the consequences of any accident previously
evaluated.

Second Standard

Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No changes
are being made to the plant that will introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators, since the containment and
reactor building function primarily as accident mitigators.

Third Standard

Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation,
including the performance of the containment and reactor building.
These components are already capable of performing as designed, and
their functions are verified by visual examination and leakage rate
testing. The ability of the containment and reactor building to
perform their design function will not be impaired by the
implementation of this amendment at Catawba Nuclear Station.
Consequently, no safety margin will be impacted.

The NRC staff reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Project Director: Herbert N. Berkow.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida

Date of amendment request: January 27, 1999.
Description of amendment request: The proposed amendment would
provide a one-time extension of the inspection interval for the Once
Through Steam Generator (OTSG) tubes specified in the Crystal River
Unit 3 (CR-3) Improved Technical Specifications (ITS) to coincide with
the planned operating cycle. CR-3 ITS 5.6.2.10 requires the OTSG
inspection interval to be 24 calendar months for Category C-2
inspection results. However, due to a previous extended maintenance
outage, the next OTSG inspection at CR-3, which is planned for the
October 1999 refueling and maintenance outage, will be approximately 26
calendar months since the last inspection. Florida Power Corporation
indicated that the total interval between inspections would correspond
to less than 21.6 months of plant operation at a temperature of
500 deg.F or above (measured at the hot leg side of the OTSG). The
licensee stated that the conclusions reached in the operational
assessments for the OTSGs show leakage and structural integrity are
maintained by substantial margins until the end of the planned
operating cycle.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.

1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
The last Crystal River Unit 3 (CR-3) Once Through Steam
Generator (OTSG) tube surveillance was completed in August 1997.
Both standard and enhanced eddy current techniques were used to
inspect 100% of the OTSG tubes. Operational assessments performed
for CR-3 provide reasonable assurance that the OTSG performance
criteria meet the leakage and structural requirements in Draft
Regulatory Guide-1074. These performance criteria will be maintained
until the end of the planned operating cycle. These operational
assessments demonstrate that operation is acceptable for an
operating cycle length of up to 21.6 months of operating time at a
temperature of 500 deg.F or above (measured at the hot leg side).
The operational assessments concluded that the projected
cumulative leakage for the limiting OTSG would be less than 1 gallon
per minute (gpm) under the limiting accident conditions at the end
of the planned operating cycle. Thus, the accident analysis
assumptions bound the condition of the OTSGs, and structural and
leakage integrity will be maintained for the proposed operating
cycle. Therefore, the proposed one-time change does not involve a
significant increase in the probability or consequences of any
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from previously evaluated accidents?
No new failure modes or accident scenarios are created by
changing the inspection from a frequency based on calendar months,
to a one-time interval based on up to 21.6 months of operating time
at a temperature of 500'F or above (measured at the hot leg side).
Plant systems and components will not be operated in a different
manner as a result of this change. Thus, this change does not
increase the risk of a plant trip or present a challenge to any
other safety system. For all known degradation mechanisms in the CR-
3 OTSGs, the most recent operational assessments bound the
probability of tube burst and project primary-to-secondary leakage
at accident conditions for the end of Operating Cycle 11 to be less
than 1 gpm. Therefore, the proposed one-time change does not create
the possibility of a new or different kind of accident from any
previously evaluated.
3. Involve a significant reduction in a margin of safety?
Improved Technical Specification (ITS) Bases 3.4.12 contains
relevant information pertaining to the limitations on reactor
coolant system (RCS) leakage. The ITS Bases discuss the 1 gpm
primary-to-secondary leakage assumed for a main steam line break
accident, as well as for a steam generator tube rupture accident.
The evaluation provided by this license amendment request shows that
tube structural integrity is maintained, thus the required
structural margins specified in NRC Regulatory Guide 1.121 are
satisfied. The operational assessments performed show the maximum
accident leakage, assuming all these indications leak, is less than
1 gpm. Therefore, all known OTSG tube degradation mechanisms have
been assessed, and the proposed one-time change does not involve a
significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied.
Therefore, the NRC staff proposes to determine that the amendment
request involves no significant hazards consideration.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428
Attorney for licensee: R. Alexander Glenn, General Counsel, Florida
Power Corporation, MAC--A5A, P. O. Box 14042, St. Petersburg, Florida
33733-4042.
NRC Project Director: Cecil O. Thomas.

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
revise Technical Specification (TS) Table 3.7-6, ``Area Temperature
Monitoring,'' by increasing the temperature limits for the fuel
building fuel pool pump cubicles and fuel building general area. The
amendment would also change the Millstone Unit 3 licensing basis by
incorporating into the Millstone Unit 3 Final Safety Analysis Report
(FSAR) a revision to describe the full core off-load condition as a
normal evolution. In

[[Page 11963]]

addition, the amendment would increase the maximum bulk spent fuel pool
(SFP) temperature from 140 deg. F to 150 deg. F, allow the crediting of
evaporative cooling as a decay heat removal mechanism for the SFP (use
of the ONEPOOL computer code), and allow the use of Holtec's quality
assurance validated DECOR computer code as a method for predicting
decay heat loads in the SFP pool.
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 the changes 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
[satisfied]. 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 analyzed.
The proposed license amendment will permit NNECo to conduct full
core off-loads as a normal evolution through the end of plant life.
This amendment request does not affect: (1) the number of spent fuel
assemblies allowed in the spent fuel pool, (2) Spent Fuel Pool (SFP)
criticality analysis, (3) structural analysis of the spent fuel pool
or (4) radiological release scenarios.
The proposed license amendment permits the use of ORIGEN2 based
DECOR and ONEPOOL codes for the analysis of the Unit 3 SFP. The
ORIGEN2 based DECOR code more accurately predicts decay heat loads
from the spent fuel in the SFP. The ONEPOOL code credits the effect
of evaporative cooling on the SFP bulk temperature. The use of these
codes improves the accuracy of predicting SFP bulk temperatures
during normal and abnormal refueling scenarios.
The analysis of decay heat removal permits the discharge of fuel
from the reactor vessel to the SFP [to] start as early as 132 hours
(depending on cooling water temperature) after reactor shutdown at a
rate of 3 assemblies per hour. The existing accident analysis for a
dropped spent fuel bundle during refueling bounds this situation as
the analysis assumed a decay time of 100 hours after reactor
shutdown.
The increase in pool temperature from 140 deg. F to 150 deg. F
does not significantly impact the structural integrity of the fuel
handling equipment. The temperature increase does not create a new
failure of the fuel handling equipment that has not been previously
analyzed.
The increased SFP temperature results in higher ambient
temperatures in the Fuel Building. However, the duration of an
increased pool temperature event is limited. The effect on the
environmental qualification (EQ) of electrical equipment is an
increase in the Maximum Normal and Abnormal Excursion temperatures,
which are based on short duration excursions from the predicted
summer maximum temperatures. This is reflected in the proposed
Technical Specification (TS) temperature changes. The temperature
limits within TS, 3.7.14, ``Plant Systems: Area Temperature
Monitoring,'' Table 3.7-6, for the Fuel Pool Pump Cubicles and Fuel
Pool General Area increase from 110 deg. F to 119 deg. F, and from
104 deg. F to 108 deg. F respectively, based upon the revised
environmental conditions. The proposed TS changes do not involve a
significant increase in the probability or consequences of an
accident previously analyzed as the Fuel Building Ventilation System
is qualified for the increased temperature and humidity conditions.
There are no changes in the EQ of equipment.
A comprehensive review of the design of the SFP, Spent Fuel Pool
Cooling and Purification system and other associated systems,
structures and components has been completed. All systems,
structures and components are fully qualified at the higher SFP
temperature of 150 deg. F for a full core off-load as a normal
operation.
Therefore, based on the above, this change will not involve a
significant increase in the probability or consequence of an
accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The proposed license amendment will permit NNECo to conduct full
core off-loads as a normal evolution through the end of plant life.
There are no physical plant changes. The SSCs [systems, structures,
and components] supporting the SFP and Spent Fuel Pool Cooling are
fully qualified for operation at 150 deg. F. The higher Fuel Pool
Pump Cubicles and Fuel Pool General Area temperatures do not create
the possibility of a new or different kind of accident from any
previously evaluated. Thus the changes do not create the possibility
of an accident of a different type than previously evaluated.
3. Involve a significant reduction in the margin of safety.
The proposed license amendment will permit NNECo to conduct full
core off-loads as a normal evolution through the end of plant life.
The proposed changes allow a higher heat load in the SFP which
results in a higher calculated maximum temperatures than the current
analysis. In addition, several changes have been made with respect
to the analysis methods used in calculating the maximum
temperatures.
The new analysis demonstrates that the SFP cooling configuration
will maintain the SFP pool bulk temperature at or below 150 deg. F
with a single train of spent fuel pool cooling. This temperature is
above the SRP [Standard Review Plan] guidance of 140 deg. F but is
well below the 212 deg. F limit permitted for abnormal core off-
loads as defined in the Standard Review Plan (NUREG -0800). This
temperature guideline of 140 deg. F was one of the acceptance
criteria credited by the NRC staff during their review of the
adequacy of the design of the SFP Cooling System within the NRC
Safety Evaluation Report (SER) for Millstone Unit 3 (NUREG-1031) and
consequently requires prior review and approval.
A single active failure will cause the loss of one of the two
trains of spent fuel pool cooling. The complete loss of cooling to
the Spent Fuel Pool is not a creditable occurrence in that the Fuel
Pool Cooling System is designed to be able to withstand the worst
single failure and still be able to perform its intended function.
However, a loss of cooling analysis indicates that several hours are
available during a refueling, and over thirteen hours are available
during normal operations for operators to respond to the loss of
cooling prior to the Spent Fuel Pool reaching its structural design
temperature of 200 deg. F.
A comprehensive review of the design of the SFP, Spent Fuel Pool
Cooling and Purification System and other associated systems,
structures and components has been completed for qualification at
the higher pool temperature of 150 deg. F. All systems, structures
and components are fully qualified at the higher Technical
Specification Fuel Pool Pump Cubicles and Fuel Pool General Area
temperatures, and at the increased SFP temperature, and are
therefore qualified for a full core off-load as a normal operation.
The ORIGEN2 based DECOR code more accurately predicts decay heat
loads from the spent fuel in the SFP. The ONEPOOL code credits the
effect of evaporative cooling on the SFP bulk temperature. The use
of these codes improves the accuracy of predicting SFP bulk
temperatures during normal and abnormal refueling scenarios. The use
of these computer codes as a method for predicting decay heat loads
and crediting evaporative cooling as a decay heat removal mechanism
have not previously been evaluated for Unit 3, and therefore,
require[s] prior NRC review and approval.
Therefore, based on the above, this license amendment to permit
NNECo to conduct full core off-loads as a normal evolution, increase
the maximum SFP pool bulk temperature from 140 deg. F to 150 deg. F,
use the ORIGEN2 based DECOR and ONEPOOL computer codes to calculate
the decay heat load and determine the effects of evaporative cooling
respectively, and increase the TS Fuel Pool Pump Cubicles and
General Area temperatures, does not involve a significant reduction
in the margin of safety.
Thus, it is concluded that the proposed amendment does not
involve a significant reduction in the margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed amendment does not involve an SHC.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.

[[Page 11964]]

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: February 10, 1999.
Description of amendment request: The proposed amendment would
incorporate alternative inspection requirements into Technical
Specification Surveillance Requirement 3/4.4.10, ``Structural
Integrity,'' for the reactor coolant pump flywheel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with 10
CFR 50.92 and has concluded that the revision does not involve a
Significant Hazards Consideration (SHC). The basis for this
conclusion is that the three criteria of 10 CFR 50.92(c) are not
satisfied. The proposed revision does not involve a SHC because the
revision would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
This proposed revision to the Millstone Unit No. 3 Technical
Specifications incorporates alternative reactor coolant pump
flywheel inspection requirements into Surveillance 4.4.10 based on
Topical Report WCAP-14535A. WCAP-14535A provided a technical basis
for the elimination of inspection requirements for reactor coolant
pump flywheels based on industry data. The industry data indicated
that no indications that would affect the integrity of flywheels was
[sic] revealed during 729 examinations of 217 flywheels at 57 plants
(including Millstone Unit No. 3). The NRC, during their review and
approval of the WCAP required continued inspections on a ten year
interval to protect against events and degradation that were not
anticipated and had not been considered in the WCAP analysis. The
proposed alternate inspection requirements are consistent with the
conclusions of an NRC review and generic approval of Topical Report
WCAP-14535A. Thus, it is concluded that the proposed revision does
not significantly increase the probability of an accident.
Additionally, the performance of reactor coolant pump flywheel
surveillances does not increase the consequence of an accident
previously evaluated.
Therefore, it is concluded that the proposed revision does not
involve a significant increase in the probability or consequence of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
This proposed revision to the surveillance does not change the
operation of any plant system or component during normal or accident
conditions. The proposed change incorporates alternate inspection
requirements for the reactor coolant pump flywheels that were
generically approved for use by licensees by the NRC. This change
does not include any physical changes to the plant.
Thus, this proposed revision does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
This proposed revision to the Millstone Unit No. 3 Technical
Specifications incorporates alternative reactor coolant pump
flywheel inspection requirements into Surveillance 4.4.10 that are
consistent with the conclusions of an NRC review and generic
approval of Topical Report WCAP-14535A. The current inspection
requirements of Surveillance 4.4.10 and the NRC review of WCAP-
14535A were both based on the recommendations of Regulatory Guide
1.14.
Thus, it is concluded that the proposed revision does not
involve a significant reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve an SHC.

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

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota

Date of amendment requests: February 5, 1999.
Description of amendment requests: The proposed amendments would
modify the technical specifications (TS) to incorporate Revision 3 of
the ABB Combustion Engineering, Inc.'s topical report, CEN-629-P,
``Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using
Leaktight Sleeves'', dated September 1998 (proprietary and
nonproprietary documents available). The current TS requires that steam
generator tube repair using the Combustion Engineering Inc.'s welded
sleeves shall be in accordance with the methods and criteria described
in Revision 2 of CEN-629-P and Addendum 1, Revision 1 of CEN-629-P.
Incorporation of Revision 3 of CEN-629-P would involve the following TS
changes: (1) editorial/administrative change to TS.4.12.D.3 to reflect
adoption of Revision 3 of CEN-629-P, and deletion of reference to
Addendum 1, Revision 1 of CEN-629-P since Revision 3 incorporates
Addendum 1, Revision 1 of CEN-629-P; (2) changes in sleeve installation
practices that incorporate improvements gained by prior experiences;
and (3) more restrictive change to the sleeve repair limit as specified
in TS.4.12.D.1.(f) from 31 percent of the nominal sleeve wall thickness
to 25 percent.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
evaluated.
Editorial changes have no effect on probability or consequences
of accidents previously evaluated. Changes in installation practices
incorporate improvements gained by experience in installing sleeves.
Further, the changes in the installation practices will change
neither the final configuration of installed sleeves nor the post-
installation NDE [nondestructive examination] from that which is
already approved. Accident induced steam generator tube leakage is
not [a]ffected by these changes. Post installation non-destructive
examination will be conducted using VT, UT, and ET as previously
licensed. The changes in repair limits have [led] to repair limits
that are more conservative than those which have been previously
approved. Thus, none of these changes will create the possibility of
a new or different kind of accident from any accident previously
analyzed.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident previously analyzed.
Editorial changes cannot create the possibility of a new or
different kind of accident. Changes in installation practices
incorporate improvements gained by experience in installing sleeves.
Further, changes in installation practices do not change the final
configuration of installed sleeves from that which is already
approved. The changes in repair limits have [led] to repair limits
that are more conservative than those which have been previously
approved.

[[Page 11965]]

Thus, none of these changes will create the possibility of a new or
different kind of accident from any accident previously analyzed.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
Editorial changes have no effect on the margin of safety.
Changes in installation practices incorporate improvements gained by
experience in installing sleeves. Further, changes in installation
practices do not change the final configuration of installed sleeves
from that which is already approved. The changes in repair limits
have [led] to repair limits that are more conservative than those
which have been previously approved. None of these changes will
affect the tube plugging assumptions used in the PINGP [Prairie
Island Nuclear Generating Plant] accident analyses. Thus, none of
these changes will 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 requests involve no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.

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

Date of amendment request: January 15, 1999.
Description of amendment request: The proposed changes revise
calibration requirements for the local power range monitors (LPRM).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

Operation of the FitzPatrick plant in accordance with the
proposed amendment would not involve a significant hazards
consideration as defined in 10 CFR 50.92, since it would not:
1. involve a significant increase in the probability or
consequences of an accident previously evaluated.
This change proposes to remove the listed requirement for the
method of calibration of the LPRM Signal from TS Table 4.1-2 because
the definition for Instrument Channel Calibration provides the
necessary guidance.
Other changes to the bases and adopting signal calibration
frequency units of MWD/T [Megawatt Days per Ton] vice effective full
power hours is consistent with STS [Standard Technical
Specification].
The proposed changes do not increase the probability of an
accident because the proposed surveillance requirements still ensure
that the LPRM signal is adequately calibrated. The proposed change
provides assurance that the associated Reactor Protection System
(RPS) functions are tested consistent with the analysis assumptions.
As a result, the consequences of an accident are not affected by
this change. This change will not alter assumptions relative to the
mitigation of an accident or transient event. Therefore, this change
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes will not physically alter the plant. As
such, no new or different types of equipment will be installed. The
methods governing normal plant operation and testing are consistent
with current safety analysis assumptions. Therefore, this change
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 removes specific calibration method
information in Table 4.1-2 regarding the LPRM signal which is
adequately addressed in the definition for Instrument Channel
Calibration.
Other changes to the Bases and adopting a signal calibration
Frequency units of MWD/T vice effective full power hours is
consistent with STS.
The proposed changes still provide the necessary control of
testing to ensure operability of the RPS instrumentation. The safety
analysis assumptions will still be maintained, thus no question of
safety exists. Therefore, this change does not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Project Director: S. Singh Bajwa, Director.

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

Date of amendment request: February 2, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 5.6, ``Fuel Storage, Criticality,''
to change the maximum unirradiated fuel assembly enrichment value for
new fuel storage from 4.5 to 5.0 weight percent Uranium-235 and to
allow the use of equivalent criticality control to that provided by the
current TS requirement of 2.35 mg of Boron-10 per linear inch loading
in the Integral Fuel Burnable Absorber (IFBA) pins.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(a) Fuel Assembly Drop.
There is no increase in the probability of a fuel assembly drop
accident because the mass of a fuel assembly does not increase when
the fuel enrichment is increased. This amendment affects only the
isotopic composition within the fuel pellets of a fuel assembly
without involving any changes to the outward physical
characteristics or structural integrity of the assembly.
The radiological consequences of a new fuel assembly drop
accident do not increase as a consequence of the proposed change to
new fuel enrichment. Because it has not been irradiated, there are
no significant radiological consequences associated with fresh fuel.
The radiological consequences of an irradiated fuel assembly drop
were previously evaluated and approved in the Spent Fuel license
amendment numbers 151/131 (Units 1 & 2 respectively).
(b) Misplaced Fuel Assembly in New Fuel Storage Vault or Spent
Fuel Storage Racks.
There is no increase in the probability of a misplaced fuel
assembly in the New Fuel Storage Vault or Spent Fuel Storage Racks.
The proposed change does not alter the physical structure of the New
Fuel Storage Vault or the Spent Fuel Storage Racks. All new fuel
assembly movements will continue to be made in accordance with
approved procedures.
There is no increase in the consequences of misplacing a fuel
assembly in the new fuel storage racks. The normally-dry new fuel
vault Keff is very small (approximately 0.65), as such,
there is sufficient reactivity margin to the 0.95 limit to bound any
possible misplacement. The double contingency principle does not
require consideration of a second unlikely event. Since a misplaced
bundle constitutes the first unlikely event, presence of moderator
in the normally dry

[[Page 11966]]

new fuel storage racks (a second unlikely event) is not assumed in
evaluating the event.
The inadvertent misplacement of a fresh fuel assembly in the
spent fuel storage racks has the potential for exceeding the
limiting reactivity, should there be a concurrent and independent
accident condition resulting in the loss of all soluble boron.
Administrative procedures to assure the presence of soluble boron
during fuel handling operations will preclude the possibility of the
simultaneous occurrence of the two independent accident conditions.
The analyses supporting Amendments 151/131 demonstrated that 600 ppm
of soluble boron is adequate to compensate for a mis-loaded fuel
event, while plant procedures require the concentration to be
maintained at least 2300 ppm. The proposed change to allow reduced
IFBA B-10 loading does not invalidate these prior analyses since
equivalent reactivity hold down to the 2.35 mg/linear inch B-10
loading will be maintained.
(c) Introduction of Moderator to the New Fuel Vault
There is no increase in the probability of any accident
involving moderator introduction to the new fuel storage vault. The
proposed change affects only the enrichment within the fuel
assemblies. No other plant systems or components are affected by
this change.
There is no increase in the consequences of introducing a
moderator to the new fuel storage vault resulting from increased
fuel enrichment. The new fuel storage vault has been analyzed for
storage of fuel assemblies with nominal enrichments of 4.65 w/o
U235 at the fully flooded condition and 5.00 w/o
U235 at the optimum moderation condition, as described in
the attached Criticality Analysis (Attachment 2). As long as the
requirement for the number of IFBA pins versus assembly enrichment
is met, calculated Keff (including uncertainties and
biases) does not exceed 0.95 under full density conditions and does
not exceed 0.98 under optimum moderation conditions.
These analyses demonstrate that 5.0 w/o enrichment fuel storage
in the New Fuel Storage Vault complies with criticality acceptance
criteria for all moderation conditions. Therefore, based on the
conclusions of the above analyses, the proposed changes will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Do not create the possibility of a new or different kind of
accident from any accident previously analyzed.
The proposed Technical specification changes do not involve any
physical changes to the plant or any changes to the method in which
the plant is operated. No physical changes to the new fuel or spent
fuel storage racks are required, nor any changes in the process or
procedures to place fuel in the racks. The enrichment limits and
reactivity hold-down requirements ensure that the assumptions used
in the criticality analyses remain bounding. As such, these changes
do not affect the performance or qualification of safety-related
equipment. Therefore, the possibility of a new or different type of
accident than previously considered i[s] not created.
3. Do not involve a significant reduction in a margin of safety.
The new fuel storage vault has been analyzed for storage of fuel
assemblies with nominal enrichments of 4.65 w/o U235 at
the fully flooded condition and 5.00 w/o U235 at the
optimum moderation condition, as described in the attached
Criticality Analysis (Attachment 2). As long as the requirement for
the number of IFBA pins versus assembly enrichment in Equation 1 is
met, calculated Keff (including uncertainties and biases)
does not exceed 0.95 under full density conditions and does not
exceed 0.98 under optimum moderation conditions.
For the 5.00 w/o U235 enrichment requested, Equation
2, which bounds Equation 1, will be used in the Technical
Specifications related to new fuel storage.
Therefore, since the calculated values of Keff have
been shown to be below the regulatory limits (including
uncertainties and biases) and because they reflect a substantial
subcritical configuration under adverse conditions, the proposed
changes will not result in a significant reduction in the plant's
margin of safety.
Previous analyses provided in support of Amendments 151/131
demonstrate that the addition of new fuel having IFBA pins with a
loading of 2.35 mg B-10 per linear inch to the spent fuel racks does
not result in a reduction in the margin of safety. Thus, providing
for reactivity hold down for IFBA pins which is equivalent to a
nominal 2.35 mg B-10/linear inch loading in fresh fuel in the spent
fuel storage racks maintains the current margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: Elinor G. Adensam.

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

Date of amendment request: February 5, 1999.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TSs) to incorporate some of the
generic changes to the Improved Technical Specifications that have been
previously approved by the NRC. In addition, a TS has been added that
would test the Unit 1 automatic scram relay on a periodic basis.
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 the consequences of a previously evaluated event
for the following reasons:

Proposed Change One

The majority of primary containment isolation valves (PCIVs)
should be in the closed position following an accident to prevent
the release of radiation to the environment. Locked PCIVs are
verified to be in the closed position prior to being locked.
Therefore, it is unnecessary for these valves to be verified closed
under the provisions of Surveillance Requirements (SRs) 3.6.1.3.2
and 3.6.1.3.3. The fact that the valves are secured closed assures
they will be in the safe position following an accident.
Furthermore, per Plant Hatch procedure, locked valves are
periodically verified to be in their correct position. This provides
additional assurance the valves will remain in the correct position.
For these reasons, the proposed change does not involve a
significant increase in the probability or the consequences of a
previously evaluated event.

Proposed Change Two

This proposed change does not affect the function of the control
rods, the control rod drive (CRD) system, or the control rod
housing. Thus, the probability of the control rod drop accident
(CRDA) is not increased. Also, this change does not affect the
function of the rod worth minimizer (RWM). As with the present
Technical Specification, no control rods will be moved (via SRs
3.1.3.2 and 3.1.3.3) when below the low power setpoint (LPSP) to
limit interference with respect to the RWM's function in limiting
the consequences of a CRDA. Additionally, no other systems designed
to prevent or mitigate the consequences of any other transient or
accident are affected.

Proposed Change Three

This proposed change merely deletes a redundant specification in
the control rod operability section. The requirement to electrically
disarm an inoperable withdrawn control rod ensures the validity of
banked position withdrawal sequence (BPWS) is maintained, thus
ensuring the mitigation of the consequences of the CRDA. This
proposed change in no way affects the BPWS, the RWM, or the
structures of the control rods and control rod drive. Thus, the
probability, or the consequences, of a previously evaluated event
are not increased by this proposed change.

Proposed Change Four

Any physical deterioration of a station service battery that can
cause degradation of

[[Page 11967]]

battery performance will result in failure of the SR, with the
ensuing inoperable declaration of the battery. A determination that
battery performance is not degraded, or will not degrade, will
result from evaluation of the particular abnormality found while
performing the Surveillance. This is the intent of the Technical
Specification as clarified in the Bases.
Accordingly, the safety function of the station service
batteries is not compromised as a result of this proposed change.
Thus, the consequences of a previously evaluated event are not
affected by this proposed revision. The proposed revision does not
affect any system needed to prevent the occurrence of previously
analyzed events; therefore, the probability of occurrence of a
previously evaluated event is not increased.

Proposed Change Five

The purpose of the primary containment air interlock is to
provide access to the primary containment while maintaining
containment integrity. Extending the Surveillance Frequency on the
airlock to once per 24 months will not increase the likelihood of
occurrence of any previously evaluated event, since no change in the
operation or testing of any system designed for the prevention of
accidents and transients is being made.
Extending the Frequency of the airlock interlock Surveillance
does not increase the consequences of any accident or transient,
since the proposed change does not affect any system designed to
mitigate the consequences of a previously analyzed event. In fact,
the extended Frequency will challenge the airlock interlock less;
thus, the likelihood of a loss of primary containment integrity will
decrease.

Proposed Change Six

This proposed change to the Safety Function Determination
Program (SFDP) description in LCO [Limiting Condition for Operation]
3.0.6 is more restrictive than the existing version. Requiring an
SFDP evaluation upon entry into LCO 3.0.6, as stated in the Bases,
will not increase the probability of occurrence or the consequences
of a previously evaluated event, since this is purely an
administrative change to clarify the intent of LCO 3.0.6 and provide
consistency with the Bases.

Proposed Change Seven

This proposed administrative change merely relocates the review
requirements for the Offsite Dose Calculation Manual (ODCM) to
licensee controlled documents. This change does not affect any
system designed for the prevention or mitigation of previously
analyzed events or any assumptions regarding transient and accident
analyses.

Proposed Change Eight

This proposed administrative change eliminates some of the
redundant reporting requirements for safety limit violations listed
in the Technical Specifications. This change does not affect any
systems designed for the prevention or mitigation of any previously
evaluated accident or transient. Additionally, the change does not
affect any assumptions of previously evaluated accidents or
transient analyses.

Proposed Change Nine

This change adds a footnote to Unit 1 Technical Specifications
Table 3.3.1.1-1 to ensure the auto scram relays (K14s) are tested as
part of the manual scram Functional Test. This change does not
adversely affect the ability of the reactor protection system (RPS)
to perform its safety function. In fact, the added testing
requirement enhances the ability to detect and correct problems with
the RPS. Successful testing of the K14s on a weekly basis for many
years has demonstrated that the additional testing requirements do
not impose an undue burden on the system. No other systems designed
for the prevention or mitigation of accidents are affected by this
change. Therefore, the probability, or the consequences, of a
previously evaluated event are not increased.
2. The proposed changes do not create the possibility of an
accident of a new or different kind from any previously evaluated.

Proposed Change One

Removing the SR to verify locked valves are in their ``safe''
position does not increase the likelihood of occurrence or
consequences of a new type of event, since no new modes of operation
are introduced. All plant systems will continue to be operated
within their design basis. Since the valves are verified to be in
their safe position prior to locking, and are periodically verified
to be in that position per the locked valve procedure, the valves
will be in the position assumed by accident analyses should an event
occur.

Proposed Change Two

This proposed change does not affect the function of either the
CRD system or the RWM. These systems, as well as all other systems
designed for the prevention or mitigation of accidents, will
continue to function per their design basis. Also, the BPWS will
continue to be used for control rod withdrawal. Thus, no new modes
of operation that would cause a type of failure different from any
previously analyzed are introduced.

Proposed Change Three

Deleting Required Action B.1 of Technical Specification 3.1.3
does not eliminate any Required Actions, since the subject Required
Action is redundant. Deleting the redundant specification does not
prevent any of the control rod control systems from performing their
functions per their design bases. Therefore, no new modes of
operation are introduced, and the probability of a new type event is
also not introduced by this proposed change.

Proposed Change Four

No changes to the operation, maintenance, or testing of the
batteries are proposed. The batteries will continue to operate
within their design basis. As a result, no new modes of operation
are introduced, and thus, the probability of occurrence of a new
type event is not created.

Proposed Change Five

This change is administrative in the sense that it does not
result in the airlock being operated or tested outside of its
design. The proposed revision only includes a change to the
Frequency of SR 3.6.1.2.2, which tests the interlock's ability to
prevent the two primary containment airlock doors from opening at
the same time. This change does not affect how the test is to be
performed or how the doors are operated. Therefore, the probability
of occurrence of a new type event is not increased by the proposed
change.

Proposed Change Six

This proposed administrative change to the SFDP description does
not involve the operation of any safety-related system. Furthermore,
this change does not involve accident or transient analyses; thus,
no changes to the assumptions for the analyses are made. As a
result, the probability of occurrence of a new type event is not
increased.

Proposed Change Seven

This administrative change merely relocates the review
requirements for the ODCM to licensee controlled documents. This
change does not affect any system designed for the prevention or
mitigation of previously analyzed events or any assumptions
regarding transient and accident analysis. Accordingly, the
possibility of a new type event is not created.

Proposed Change Eight

This administrative change eliminates some of the redundant
reporting requirements for safety limit violations listed in the
Technical Specifications. This change does not affect any systems
designed for the prevention or mitigation of any previously
evaluated accident or transient. Additionally, the change does not
affect any assumptions of previously evaluated accident or transient
analyses. Accordingly, the possibility of a new type event is not
created.

Proposed Change Nine

Adding a requirement to test the auto scram relays (K14s) on a
weekly basis does not create a new mode of operation for the RPS.
Also, no other safety-related systems are affected by this change,
and as a result, the possibility of occurrence of a new type
accident is not created.
3. The changes do not significantly reduce the margin of safety.

Proposed Change One

Not requiring position surveillance on PCIVs locked in position
does not reduce the margin of safety, because the valves are
verified to be in their ``safe'' position prior to locking. This
ensures the valve will remain in the ``safe'' position until it is
unlocked again. The position of these locked valves is verified
periodically by the Operations Department. Furthermore, a
``malicious'' unlocking of the valves is unlikely to take place,
since the keys to the valves are controlled by the shift supervisor
(SS). Anyone wanting to check out a key must obtain SS approval.
Also, the locked valves are periodically verified to be in their
proper

[[Page 11968]]

position whenever Operations Management deems it necessary. For
these reasons, the margin of safety is not significantly reduced.

Proposed Change Two

Moving the Technical Specification 3.1.3 Note from the Required
Action column to the Completion Time column will not affect the
safety function of the RWM system. The RWM will continue to function
through the power ranges where the control rod drop accident is of
concern. The change does not affect the safety function of the RWM
in any way. Thus, the margin of safety is not reduced.

Proposed Change Three

This proposed change only eliminates a redundant Specification.
Adherence to the requirements of the BPWS will still be maintained
during plant startups. Also, the operation of the RWM system remains
unaffected by this proposed change. For these reasons, the margin of
safety for the CRDA is not reduced.

Proposed Change Four

This proposed change clarifies that the purpose of SR 3.8.4.3 is
to determine whether a physical deterioration that could affect
battery performance exists. This is already stated in the Plant
Hatch Technical Specifications Bases; thus, the proposed revision is
merely a clarification of the Specification. Adding this
clarification does not reduce the margin of safety with respect to
battery performance, because an engineering evaluation must be
performed to document that the particular deficiency will not
prevent the battery from performing its safety function.

Proposed Change Five

This proposed change to extend the Frequency of SR 3.6.1.2.2
reduces the number of challenges to primary containment integrity.
The nature of the Surveillance is such that the primary containment
(drywell) interlock is challenged. With that challenge, the
likelihood of a primary containment breach is increased. Therefore,
reducing the Frequency of this SR actually increases the safety of
margin, since normal entry and exit procedures do not permit
challenging the interlock.

Proposed Change Six

This purely administrative change clarifies the definition of
the SFDP in LCO 3.0.6. The Technical Specifications margin of safety
is enhanced, since the new wording, together with the existing
wording in the Bases, makes it clear that the SFDP must be performed
any time LCO 3.0.6 is entered.

Proposed Change Seven

This proposed change merely allows relocation of the review and
approval functions for the ODCM revisions from the Technical
Specifications to owner-controlled documents. The purely
administrative change does not affect any Technical Specifications
required system, test, or function. Changes to the ODCM will
continue to receive the level of review necessary to ensure any
proposed changes are accurate and complete. Therefore, the margin of
safety is not reduced.

Proposed Change Eight

This purely administrative change eliminates redundant reporting
requirements with respect to a safety limit violation. The change
has no effect on any Technical Specifications required system, test,
or function, or on any other safety-related system. Accordingly, the
margin of safety is not reduced.

Proposed Change Nine

This proposed change ensures the Unit 1 auto scram relays (K14s)
are tested on a weekly basis. General Electric recognizes this as an
optimum test frequency for these scram contactors. In this respect,
the margin of safety is increased, since this change ensures the
relays will be tested at the optimum recommended Frequency. Also, at
Plant Hatch, the K14 relays and contacts have been tested at this
Frequency for many years. As a result, placing this requirement on
the relays will not pose an undue burden on the RPS. No other
safety-related systems are affected by this proposed change. For the
above reasons, this proposed change does 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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
NRC Project Director: Herbert N. Berkow.

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

Date of amendment request: January 26, 1999.
Description of amendment request: The amendment would revise part
of the Inservice Inspection requirements for the Reactor Coolant Pump
flywheel from an in-place ultrasonic volumetric examination of the
areas of higher stress concentration at the bore and keyway at
approximately 3-year intervals and a surface examination of all exposed
surfaces and complete ultrasonic volumetric examination at
approximately 10 year intervals to ultrasonic examination over the
volume from the inner bore of the flywheel to the circle of one-half
the outer radius once every 10 years.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change increases the examination volume and revises
the periodicity of the ultrasonic examination required by Regulatory
Guide 1.14 regulatory position C.4.b(1) from 3-year intervals to 10-
year intervals. This change is consistent with the conclusions of
the NRC staff in the referenced safety evaluation of WCAP-14535. The
NRC staff has determined that the evaluation methodology is
appropriate and the criteria are in accordance with the design
criteria of RG 1.14. There is no change in the method of plant
operation or system design.
The proposed change revises the inspection process to eliminate
10-year surface examination of all exposed surfaces and complete
ultrasonic volumetric examination required by Regulatory Guide 1.14
Regulatory Position C.4.b(2). An ultrasonic volumetric examination
will be performed of a section of the flywheel once every 10 years.
This change is consistent with the conclusions of the NRC staff in
referenced safety evaluation of WCAP-14535. The NRC staff has
determined that the evaluation methodology is appropriate and the
criteria are in accordance with the design criteria of RG 1.14.
Based on the above, this change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change increases the examination volume and revises
the periodicity of the ultrasonic examination required by Regulatory
Guide 1.14 regulatory position C.4.b(1) from 3-year intervals to 10-
year intervals. This change is consistent with the conclusions of
the NRC staff in the referenced safety evaluation of WCAP-14535. The
only potential accident associated with this change is loss of the
flywheel. Precautionary measures taken to preclude missile formation
from Reactor Coolant Pump components assure that the pumps will not
produce missiles under any anticipated accident condition. Each
component of the primary pump motors has been analyzed for missile
generation Any fragments of the motor rotor would be contained by
the heavy stator. Effects on reactor coolant flow due to loss of
functionality of a single Reactor Coolant Pump flywheel are
enveloped by the analysis of the consequences of the Reactor Coolant
Pump locked rotor event. There is no change in the method of plant
operation or system design.
The proposed change revises the inspection process to eliminate
10-year surface examination of all exposed surfaces and complete
ultrasonic volumetric examination required by Regulatory Guide

[[Page 11969]]

1.14 Regulatory Position C.4.b(2). An ultrasonic volumetric
examination will be performed of a section of the flywheel once
every 10 years. This change is consistent with the conclusions of
the NRC staff in the referenced safety evaluation of WCAP-14535. The
only potential accident associated with this change is loss of the
flywheel. Precautionary measures taken to preclude missile formation
from Reactor Coolant Pump components assure that the pumps will not
produce missiles under any anticipated accident condition. Each
component of the primary pump motors has been analyzed for missile
generation. Any fragments of the motor rotor would be contained by
the heavy stator. Effects on reactor coolant flow due to loss of
functionality of single Reactor Coolant Pump flywheel are enveloped
by the analysis of the consequences of the Reactor Coolant Pump
locked rotor event.
Based on the above, this change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change increases the examination volume and revises
the periodicity of the ultrasonic examination required by Regulatory
Guide 1.14 Regulatory Position C.4.b(1) from 3-year intervals to 10-
year intervals. This change is consistent with the conclusions of
the NRC staff in the referenced safety evaluation of WCAP-14535. The
NRC staff used deterministic methodology to review the WCAP and came
to the conclusion that ASME margins would be maintained during the
service period and a 10-year inspection period appears reasonable.
There is no change in the method of plant operation or system
design.
The proposed change revises the inspection process to eliminate
the 10-year surface examination of all exposed surfaces and complete
ultrasonic volumetric examination required by Regulatory Position
C.4.b(2) of Regulatory Guide 1.14. An ultrasonic volumetric
examination will be performed of a section of the flywheel once
every 10 years. This change is consistent with the conclusions of
the NRC staff in the referenced safety evaluation of WCAP-14535.
Effects on reactor coolant flow due to loss of functionality of a
single Reactor Coolant Pump flywheel are enveloped by the analysis
of the consequences of the Reactor Coolant Pump locked rotor event.
Based on the above, this change does not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: John N. Hannon.

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee

Date of application for amendments: August 27, 1998 (TS 98-04).
Brief description of amendments: The proposed amendment would
change the Sequoyah (SQN) Technical Specifications (TSs) by adding a
provision to Section 5.3, ``Reactor Core,'' authorizing a limited
number of lead test assemblies (LTAs) to be installed in the core as
described in the Framatome Cogema Fuels Report BAW-2328 entitled
``Blended Uranium Lead Test Assembly Design Report.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:

A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The LTAs are identical to the other Mark-BW fuel assemblies with
the exception of minor differences internal to the fuel rods. These
differences will not adversely affect reactor neutronic or thermal-
hydraulic performance; therefore, they do not significantly increase
the probability of accidents while in the reactor.
The reload design analyses performed for SQN Unit 2 Cycle 10
accounts for any minor neutronic differences of the LTAs and
confirms any effects on the reload core to be within established
fuel design limits.
The pressure and temperature safety limits for the cycles in
which the LTAs will be in the core are the same as those for the
current operating cycle thus ensuring that the fuel will be
maintained within the same range of safety parameters that form the
basis for the FSAR [Final Safety Analysis Report] accident
evaluation. The potential effects of the LTAs on plant operation and
safety have been evaluated. This evaluation investigated both LOCA
[loss-of-coolant accident] and non-LOCA events, and concluded that
the current analyses remain bounding and that there will be no
increase in the probability of occurrence for any design basis
accident described in the FSAR.
The impact of the LTAs on key safety analysis parameters was
examined and it was concluded that there will be an insignificant
impact.
The impacts of the LTAs on the radiological consequences for all
postulated events have been evaluated. The total calculated source
term and the source term-activity of isotopes, which significantly
contribute to operator and off-site accident exposure levels, were
shown to be less than standard fuel assemblies, therefore, it will
not increase the consequences of any accident previously evaluated.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The fuel assembly design for the LTAs is identical to the
standard fuel assemblies. The main difference between the LTAs and
the production fuel is that the concentration of the U234
and U236 isotopes will be higher in the LTA fuel pellets
than that typically found in standard fuel. These isotopic
differences will not affect the chemical, mechanical or thermal
properties of the fuel pellet.
The LTAs meet the same design criteria and licensing basis
criteria as the standard fuel assemblies and were manufactured with
the same processes. The LTA skeleton is identical to the standard
skeleton, which ensures that the loadings associated with normal
operation, seismic events, LOCA events, and shipping and handling
are not affected.
Pressure and temperature safety limits will be maintained the
same as those for the current operating cycle, thus ensuring that
the fuel will be maintained within the same range of safety
parameters that form the basis for previous accident evaluations. No
new performance requirements are being imposed on any system or
component that exceed design criteria or cause the core to operate
in excess of design basis operating limits. No credible scenario has
been identified, which could jeopardize equipment that could cause
intensify or mitigate events or accident sequences. Therefore, the
LTAs will not create the possibility of accidents or equipment
malfunctions of a different type than previously evaluated while in
the reactor.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The LTAs will not adversely affect reactor neutronic or thermal-
hydraulic performance. The LOCA acceptance criteria with LTAs
installed in the core will continue to be met: peak cladding
temperature of less than or equal to 2200 deg.F, peak cladding
oxidation of less than or equal to 17 percent, average clad
oxidation of less than or equal to 1 percent, and long-term
coolability. The acceptance criteria for departure from nucleate
boiling (DNB) events with the LTAs installed in the core will also
continue to be met: 95 percent probability and 95 percent confidence
interval that DNB is not occurring during the transient. Other
acceptance criteria have also been demonstrated to remain within
acceptable limits. The total calculated source term-activity and the
source term-activity of isotopes, which significantly contribute to
operator and off-site accident exposure levels of the LTAs, was
determined to be less than that for the standard fuel assembly. All
previously evaluated events remain bounding and valid. For these
reasons, the proposed amendment does not involve a significant
reduction in a margin of safety.

[[Page 11970]]

The NRC 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: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Project Director: Cecil O. Thomas.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of application for amendments: August 14, 1998, as
supplemented by letters dated October 13, 1998, and December 23, 1998.
Brief description of amendments: The amendments revised the Dresden
Technical Specifications (TS) to reflect the use of Siemens Power
Corporation ATRIUM-9B fuel. Specifically the amendments incorporated
the following into the TS: (a) new methodologies that enhanced
operational flexibility and reduced the likelihood of future plant
derates; (b) administrative changes that eliminated the cycle-specific
implementation of ATRIUM-9B fuel and adopted Improved Standard
Technical Specification language where appropriate; and (c) changed the
Minimum Critical Power Ratio.

Date of issuance: February 16, 1999
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 171; 166.
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: 63 FR 48258 (September
9, 1998) and 63 FR 59588 (November 4, 1998). The October 13 and
December 23, 1998 submittals provided additional clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of application for amendments: October 16, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) to lower the power level (from 30 percent
to 25 percent rated thermal power) below which the turbine control
valve (TCV) and turbine stop valve (TSV) closure scram signals and the
end-of-cycle recirculation pump trip (EOC-RPT) signal are not in
effect. The amendments also (1) delete from TSs the reference to
turbine first stage pressure as a measure of rated thermal power, and
(2) add a requirement to periodically verify that TCV and TSV scram
trip functions and the EOC-RPT trip functions are not bypassed at
greater than or equal to 25 percent rated thermal power.
Date of issuance: February 12, 1999.
Effective date: For Unit 1--Immediately, to be implemented within
90 days; for Unit 2--immediately, to be implemented prior to startup of
L2C8.
Amendment Nos.: 130; 114.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 54108). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.

Commonwealth Edison Company, Docket No. 50-374, LaSalle County Station,
Unit 2, LaSalle County, Illinois

Date of amendment request: November 9, 1998.
Brief description of amendment: The amendment revised Technical
Specification 3/4.3.2, ``Isolation Actuation Instrumentation'' to add/
revise various isolation setpoints for leak detection instrumentation.
These changes are necessary due to modifications to the reactor water
cleanup (RWCU) system to restore ``hot'' suction to the RWCU pumps and
due to a re-evaluation of the high energy line break analysis. In
addition, the amendment eliminated isolation actuation trip functions
for the residual heat removal system steam condensing mode and shutdown
cooling mode.
Date of issuance: February 16, 1999.
Effective date: February 16, 1999.
Amendment No.: 115.
Facility Operating License No. NPF-18: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69335).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois

[[Page 11971]]

Valley Community College, Oglesby, Illinois 61348-9692.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan

Date of application for amendment: June 20, 1997 (NRC-97-0037), as
supplemented July 2, 1997 (NRC-97-0066), and March 10 (NRC-98-0036) and
April 9, 1998 (NRC-98-0083).
Brief description of amendment: The amendment revises the technical
specifications by relocating surveillance requirement 4.4.1.1.2 for
setting the reactor recirculation system motor-generator set scoop tube
stops to the updated final safety analysis report (UFSAR), with
modifications.
Date of issuance: February 8, 1999.
Effective date: February 8, 1999, with full implementation within
90 days. Implementation of this amendment shall include the relocation
of surveillance requirement 4.4.1.1.2 from the technical specifications
to the UFSAR as described in the licensee's application dated June 20,
1997, as supplemented on July 2, 1997, and March 10 and April 9, 1998,
and evaluated in the staff's safety evaluation dated February 8, 1999.
Amendment No.: 130.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 16, 1997 (62 FR
38134)
The July 2, 1997, and March 10 and April 9, 1998, submittals
provided additional clarifying information within the scope of the
original Federal Register notice and did not change the staff's initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.

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

Date of application for amendment: June 30, 1998, as supplemented
by letter dated November 23, 1998.
Brief description of amendment: The amendment authorizes the
licensee to modify the plant to correct a design deficiency with the
plant protection system (PPS). This deficiency could have rendered the
system vulnerable to a single failure (i.e., failure of a DC buss) with
one channel in bypass. The proposed modification would ensure the
required redundancy and independence for the PPS such that no single
failure results in a loss of the protection function with a channel in
indefinite bypass, and removal from service of any component or channel
does not result in a loss of the minimum redundancy required by the
Technical Specifications.
Date of issuance: February 17, 1999.
Effective date: This license amendment is effective as of its date
of issuance to be implemented within six months following the
facility's restart from refueling outage 2R14.
Amendment No.: 201
Facility Operating License No. NPF-6: Amendment revised the license
to authorize a modification to the plant protection system.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66593).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: June 29, 1998, as supplemented by letter
dated January 12, 1999.
Brief description of amendment: The amendment changes the Appendix
A TSs by modifying TS 3.7.6.1, ``Control Room Emergency Air Filtration
System'' in Modes 1-4, TS 3.7.6.2, ``Control Room Emergency Air
Filtration System'' in Modes 5 and 6, TS 3.7.6.3, ``Control Room Air
Temperature'' in Modes 1-4, TS 3.7.6.4, ``Control Room Air
Temperature,'' in Modes 5 and 6, TS 3.7.6.5, ``Control Room Isolation
and Pressurization,'' and its associated basis. This amendment also
modifies TS Tables 3.3-6 and 4.3-3 for the Control Room Intake
Monitors.
Date of issuance: February 17, 1999.
Effective date: This license amendment is effective as of its date
of issuance, to be implemented within 60 days.
Amendment No.: 149.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56247).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: May 28, 1996, as supplemented by
letter dated October 27, 1998.
Brief description of amendment: This amendment increases the test
interval for reactor protection system instrumentation and anticipatory
reactor trip system instrumentation.
Date of issuance: February 22, 1999.
Effective date: February 22, 1999.
Amendment No.: 230.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 31, 1996 (61 FR
40031). The supplemental information provided did not impact the
proposed no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 22, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Library, Government Documents Collection, 2801 West
Bancroft Avenue, Toledo, OH 43606

FirstEnergy Nuclear Operating Company, Docket No. 50-440 Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: September 8, 1997, as
supplemented by submittal dated October 27, 1998.
Brief description of amendment: This amendment revised Technical
Specification 5.2.2.e, ``Organization--Unit Staff,'' by removing the
reference to the NRC Policy Statement on working hours and
incorporating a requirement for administrative procedures necessary to
ensure that the working hours of unit staff who perform safety-related
functions are limited and controlled.
Date of issuance: February 22, 1999.
Effective date: February 22, 1999.
Amendment No.: 98.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 19, 1997 (62
FR

[[Page 11972]]

61847) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 22, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: September 3, 1998, as
supplemented by submittals dated December 3, and December 9, 1998 and
January 12, and January 26, 1999.
Brief description of amendment: This amendment revised Technical
Specification 3.8.1, ``AC Sources--Operating,'' by extending the
emergency diesel generator (EDG) Completion Time from 72 hours to 14
days for the Division 1 and 2 EDG and allows performance of the EDG 24-
hour test run in Modes 1 and 2. The amendment also establishes
Technical Specification 5.5.13.1, ``Configuration Risk Management
Program,'' an administrative program that assesses risk based on plant
status.
Date of issuance: February 24, 1999.
Effective date: February 24, 1999.
Amendment No.: 99.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56261)
The supplemental information provided clarifying information that
did not change the initial no significant hazards consideration
determination or alter the scope of the proposed action.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.

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

Date of application for amendment: January 22, 1999.
Brief description of amendment: The amendment revises Technical
Specification Surveillance Requirement (SR) 3.8.1.7 to better match
plant conditions during diesel generator (DG) testing by clarifying
which voltage and frequency limits are applicable during the transient
and steady state portions of the DG start. A Notice of Enforcement
Discretion (NOED) related to SR 3.8.1.7 was issued verbally on January
20, 1999. The NOED is documented in a letter dated January 22, 1999.
Date of issuance: February 17, 1999.
Effective date: February 17, 1999, to be implemented within 30
days.
Amendment No.: 225.
Facility Operating License No. DPR-49: Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration (NSHC): Yes (64 FR 4902 dated
February 1, 1999). The notice provided an opportunity to submit
comments on the Commission's proposed NSHC determination. No comments
have been received. The notice also provided for an opportunity to
request a hearing by March 3, 1999, but indicated that if the
Commission makes a final NSHC determination, any such hearing would
take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final NSHC determination are contained in a
Safety Evaluation dated February 17, 1999.
Attorney for Licensee: Al Gutterman; Morgan, Lewis & Bockius, 1800
M Street NW, Washington, D.C. 20036-5869.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, IA 52401

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: October 22, 1998.
Brief description of amendment: The amendment revises Technical
Specifications 3.3.2.1, ``Instrumentation--Engineered Safety Feature
Actuation System Instrumentation''; 3.4.9.3, ``Reactor Coolant System--
Overpressure Protection Systems''; and 3.5.3, ``Emergency Core Cooling
Systems--ECCS Subsystems--Tavg < 300 [degrees] F.'' The amendment
allows Millstone Unit No. 2 to prevent an automatic start of any high-
pressure safety injection (HPSI) pump when the shutdown cooling system
(SDCS) is in operation (Mode 4 and below). An inadvertent start of an
HPSI pump could result in overpressurization of the SDCS.
Date of issuance: February 10, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 227.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66600)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 10, 1999.
No significant hazards consideration comments received: No.
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.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California

Date of application for amendments: March 10, 1997, as supplemented
by letters dated May 20, 1997; March 13, August 28, and October 22,
1998; and January 29 and February 2, 1999.
Brief description of amendments: The amendments revised the
combined Technical Specifications (TS) for the Diablo Canyon Power
Plant (DCPP) Unit Nos. 1 and 2 that changed TS 3/4.4.5 and its
associated Bases to allow the implementation of steam generator (SG)
tube alternate repair criteria for axial indications in the
Westinghouse explosive tube expansion (WEXTEX) region below the top of
the tubesheet and below the bottom of the WEXTEX transition that may
exceed the current TS depth-based plugging limit.
Date of issuance: February 19, 1999.
Effective date: February 19, 1999, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 1--129; Unit 2--127.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 19, 1997 (62
FR 61843). The March 13, August 28, and October 22, 1998; and January
29 and February 2, 1999, supplemental letters provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated February 19, 1999. No

[[Page 11973]]

significant hazards consideration comments received: No.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.

Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of application for amendment: June 16, 1998.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) by moving certain administrative requirements from
the TSs to the Final Safety Analysis Report.
Date of issuance: February 25, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 188.
Facility Operating License No. DPR-64: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40560).
No significant hazards consideration comments received: No.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 25, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.

PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit
2, Luzerne County, Pennsylvania

Date of application for amendment: August 4, 1998, as supplemented
by letters dated December 16, 1998, and January 12 and 28, 1999.
Brief description of amendment: This amendment would modify the
Susquehanna Steam Electric Station, Unit 2 Technical Specifications to
replace figures 2.1.1.2-1 and 2.1.1.2-2, and associated footnotes, with
single value minimum critical power ratio Safety Limits of Section
2.1.1.2; remove references from Section 5.6.5 which do not directly
support the generation of Core Operating Limits; remove references from
Section 5.6.5 which were previously included to address the application
of the ANFB-10 correlation to ATRIUM-10 fuel; include Siemiens Power
Corporation ANFB-10 topical report in Section 5.6.5; and to change the
Bases to reflect the inclusion of the ANFB-10 critical power
correlation.
Date of issuance: February 17, 1999.
Effective date: As of date of issuance, to be implemented in 30
days.
Amendment No.: 154.
Facility Operating License No. NPF-22. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48262). The December 16, 1998, and January 12, and 28, 1999, letters
provided clarifying information that did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 17, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama

Date of amendments request: October 12, 1998.
Brief Description of amendments: The amendments revise Technical
Specification Section 6, ``Administrative Controls,'' to recognize the
additional management positions associated with the steam generator
replacement project. The new positions would provide the ability to
approve procedures regarding this project, which may affect nuclear
safety.
Date of issuance: February 19, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1--141 and Unit 2--133.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64122). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

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

Date of application for amendments: June 19, 1998, as supplemented
by letters dated December 4, 1998, and January 13, 1999.
Brief description of amendments: The proposed changes would modify
the technical specifications (TS) to (1) reduce the minimum RCS cold
leg temperature (Tc); (2) convert the specified reactor coolant system
(RCS) flow from mass units (lbm/hr) to volumetric units (gpm); and (3)
eliminate the maximum RCS flow rate limit from the TS.
Date of issuance: February 12, 1999.
Effective date: February 12, 1999, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2--149; Unit 3--141.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48266). The supplemental letters dated December 4, 1998, and January
13, 1999, provided additional clarifying information, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.

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

Date of application for amendments: November 23, 1998, as
supplemented by letter dated January 13, 1999.
Brief description of amendments: The amendments revised the
technical specifications (TS) to (1) reinstate the log power reactor
trip at or above 4E-5% RATED THERMAL POWER (RTP); (2) reinstate reactor
trips for Reactor Coolant Flow--Low (RCS flow), the Local Power
Density--High (LPD), and the Departure from Nucleate Boiling Ratio--Low
(DNBR); (3) remove the word ``automatically'' from notes (a) and (d) of
Table 3.3.1-1 to clarify that the

[[Page 11974]]

manual enable of the trip is permissible; and (4) clarify that the
setpoints on Table 3.3.1-1 are set relative to logarithmic power.
Date of issuance: February 12, 1999.
Effective date: February 12, 1999, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2--150; Unit 3--142.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998
(63 FR 71973). The January 13, 1999, supplemental information
provided additional clarifying information and did not change the
staff's initial no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.

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

Date of amendment request: November 23, 1998.
Brief description of amendments: Relocates descriptive design
information from Technical Specification (TS) Section 3.7.1.1, Table
3.7-2, regarding orifice sizes for main steam line Code safety valves,
to the Bases section for this TS.
Date of issuance: February 24, 1999.
Effective date: This license amendment is effective as of its date
of issuance, and shall be implemented within 30 days of issuance.
Amendment Nos.: Unit 1--Amendment No. 103; Unit 2--Amendment No.
90.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71974).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: December 10, 1998.
Brief description of amendment: The amendment corrects an error in
the technical specifications by changing to the use of ``hydrogen
balance air'' rather than the incorrect ``hydrogen balance nitrogen''
for calibration of the Augmented Offgass System hydrogen monitors.
Date of Issuance: February 12, 1999.
Effective date: February 12, 1999, to be implemented within 30
days.
Amendment No.: 166.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71975).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: August 20, 1997, as
supplemented on September 18, 1997, and October 31, 1997.
Brief description of amendment: The amendment makes administrative
changes to the Technical Specifications to add and revise reference to
NRC-approved methodologies which will be used to generate the cycle-
specific thermal operating limits in the Vermont Yankee Core Operating
Limits Report.
Date of Issuance: February 23, 1999.
Effective date: February 23, 1999, to be implemented within 30
days.
Amendment No.: 167.
Facility Operating License No. DPR-28: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 25, 1998 (63 FR
14489).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: December 10, 1996, as
supplemented on January 22, 1999.
Brief description of amendment: The amendment makes changes to the
Technical Specifications regarding fire protection requirements as
recommended by NRC Generic Letters 86-10 and 88-12. This includes
relocating certain fire protection requirements to the Vermont Yankee
Fire Protection Plan, Technical Requirements Manual, and Final Safety
Analysis Report.
Date of Issuance: February 24, 1999.
Effective date: February 24, 1999, to be implemented within 30
days.
Amendment No.: 168.
Facility Operating License No. DPR-28. Amendment revised the
Technical Specifications and Facility Operating License.
Date of initial notice in Federal Register: February 26, 1997 (62
FR 8801).
The January 22, 1999, supplement did not change the original
proposed no significant hazards consideration determination, or expand
the scope of the amendment request as initially noticed.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated February 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: November 4, 1998.
Brief Description of amendments: These amendments revise the
Technical Specifications (TS) to change Emergency Diesel Generator
start and load time testing requirements in TS 4.6.A.1.b. The TS Basis
Section 3.16 is also revised to reflect the basis for the new TS
requirements. The TS changes are in a conservative direction, and are
being made to bring the TS and the Updated Final Safety Analysis Report
into conformance with each other.
Date of issuance: March 1, 1999.
Effective date: March 1, 1999.
Amendment Nos.: 218 and 218.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.

[[Page 11975]]

Date of initial notice in Federal Register: January 27, 1999 (64 FR
4161).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: March 4, 1998, as supplemented
September 21, 1998.
Brief description of amendment: This amendment revises the
Technical Specifications to provide a one-hour limiting condition for
operation that will permit a safety injection pump to be used for the
addition of make-up fluid to safety injection accumulators during power
operation.
Date of issuance: February 23, 1999.
Effective date: February 23, 1999.
Amendment No.: 143.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 8, 1998 (63 FR
17237).
The September 21, 1998, supplement provided clarifying information
that did not change the initial no significant hazards determination or
alter the scope of the proposed action.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

Dated at Rockville, Maryland, this 3rd day of March 1999.

For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-5751 Filed 3-9-99; 8:45 am]
BILLING CODE 7590-01-P

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A99-5751. Public record. Not legal advice.