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

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URL: https://www.frixlaw.com/law-library/documents/fr%3A98-30691

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** November 18, 1998
- **Citation:** 63 FR 64106

## Text

NUCLEAR REGULATORY COMMISSION

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

I. Background

Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of
1954, as amended (the Act), to require the Commission to publish notice
of any amendments issued, or proposed to be issued, under a new
provision of section 189 of the Act. This provision grants the
Commission the authority to issue and make immediately effective any
amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from October 24, 1998, through November 5, 1998.
The last biweekly notice was published on November 4, 1998 (63 FR
59584).

[[Page 64107]]

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 December 18, 1998, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's

[[Page 64108]]

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. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: October 16, 1998.
Description of amendment request: The proposed amendments would
lower the power level 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) signals are not in effect. The bypass
setpoint (Pbypass) would be reduced from 30 percent rated
power to 25 percent rated power. The licensee also proposes to delete
the reference to turbine first stage pressure as a measure of core
thermal power in the Technical Specifications. To ensure that the trip
functions will not be inadvertently bypassed when they are required to
be operable, a requirement would be added to periodically verify that
TCV and TSV scram trip functions and the ECO-RPT trip functions are not
bypassed at greater than or equal to 25 percent of rated thermal power.
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 of
occurrence or consequences of an accident previously evaluated:
The probability of an accident previously evaluated will not
increase as a result of this change because the setpoint change does
not alter any of the initiators of an accident or cause them to
occur more frequently.
The consequences of an accident previously evaluated are not
impacted. LaSalle Units 1 and 2 each have approximately 30 percent
bypass capability. Therefore, a scram on TCV or TSV closure signals
is not needed until 30 percent core thermal power is reached, as
adequate steam bypass capacity is available. A lower
Pbypass remains conservative with respect to this
criterion.
LaSalle utilizes power and flow dependent thermal limits. The
power dependent portion of these thermal limits is dependent on the
Pbypass setpoint. These limits provide assurance that
adequate fuel thermal-mechanical margin is maintained through
adherence to the thermal limits Technical Specification
requirements.
Revised thermal limits have been determined based on the results
of GE transient analyses. Adhering to these thermal limits ensures
that the consequences of an accident or transient would not be
increased from the consequences under the approved 30 percent
setpoint. Adjustments to the thermal limits were determined through
use of the NRC-approved ODYN reactor dynamic model for the limiting
Load Rejection Without Bypass and the Feedwater Controller Failure
events.
The deletion of the reference to turbine first stage pressure
and rewording the Technical Specifications Notes does not affect
either accident initiators or plant equipment, as they are
administrative changes.
Adding the periodic verification that the bypass channels are
set correctly ensures that scrams or EOC-RPT will not be
inadvertently bypassed when Thermal Power is greater than or equal
to 25 percent of Rated Thermal Power. The statement that
specification 4.0.2 applies to the 18 month interval is needed,
since the notes are not standard surveillance requirements and the
interval is consistent with other similar instrumentation to which
4.0.2 currently applies.
Therefore, the proposed changes do not involve a significant
increase in the probability of occurrence 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 setpoint change and proposed bypass verification notes
ensure that the scrams for TSV closure and TCV fast closure, and
EOC-RPT, will be enabled above 25 percent of rated thermal power,
rather than above 30 percent of rated thermal power. This change
results in simplified reload transient analyses and does not impact
any other equipment.
No other physical modifications are being proposed by this
submittal. The only plant operational impact is that between 25
percent and 30 percent power, the plant will now scram upon a
turbine trip, which is an analyzed transient.
The remaining changes to Technical Specification wording are
administrative in nature and consistent with other Technical
Specifications.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
(3) Involve a significant reduction in the margin of safety:
LaSalle Units 1 and 2 each have approximately 30 percent bypass
capability. Therefore, a scram on TCV or TSV closure signals in not
needed until 30 percent core thermal power is reached, as adequate
steam bypass capacity is available. However, reduction of this
setpoint to 25 percent power actually aids the plant transient
response between 25 percent and 30 percent power.
The new thermal limits reflect the revised setpoint and have
been determined based on revised limiting transient analyses that
have included the new Pbypass value. If a transient were
to occur, the revised operating limits ensure that adequate margin
would be available to preclude violation of the Minimum Critical
Power Ratio (MCPR) safety limit and the fuel thermal-mechanical
limits.
All other UFSAR [Updated Final Safety Analysis Report] events
are either bounded by the analyses performed or are not impacted by
the Pbypass change.
The wording changes to the Technical Specifications do not
change the requirement for the bypass function and for maintaining
the bypass function and thus do not affect the analyses discussed
above.
The addition of the Notes periodically verifying the TCV and TSV
Closure Trip Functions are not bypassed at greater than or equal to
25 percent Rated Thermal Power ensures the trip functions will not
be inadvertently bypassed when required to be Operable.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Jacobs Memorial Library, 815
North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,
Illinois 61348-9692.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
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: July 22 and October 22, 1998.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to reflect the licensee's
planned use of fuel supplied by Westinghouse. The

[[Page 64109]]

Westinghouse fuel has different design characteristics from the fuel
currently in use. Accordingly, the following changes would need to be
made to the TS: Figure 2.1.1-1, ``Reactor Core Safety Limits--Four
Loops in Operation''; various core operating parameters specified by
Surveillance Requirements 3.2.1.2, 3.2.1.3, and 3.2.2.2; Section 4.2.1,
``Fuel Assemblies''; and Section 5.6.5, ``Core Operating Limits Report
(COLR).''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, addressing the three standards of 10 CFR 50.92(c):

First Standard

Implementation of this LAR [license amendment request] would not
involve a significant increase in the probability or consequences of
an accident previously evaluated. The revised Reactor Core Safety
Limits Figure further restricts acceptable operation. Moving an
uncertainty factor from the Improved Technical Specifications to the
Core Operating Limits Report (COLR) does not exempt this factor from
regulatory restrictions. COLR parameters are generated by NRC
approved methods with the intent of ensuring that previously
evaluated accidents remain bounding. The COLR is submitted to the
NRC upon implementation of each fuel cycle or when the document is
otherwise revised. No accident probabilities or consequences will be
impacted by this LAR.

Second Standard

Implementation of this LAR would not create the possibility of a
new or different kind of accident from any previously evaluated. The
revised Reactor Core Safety Limits Figure further restricts
acceptable operation. Moving an uncertainty factor from the Improved
Technical Specifications to the COLR does not exempt this factor
from regulatory restrictions. Since the parameter in question is not
being deleted, the possibility of a new or different kind of
accident from any previously evaluated does not exist.

Third Standard

Implementation of this LAR 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.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. Use of the ZIRLOTM
cladding material has been reviewed and approved in Reference 1 (as
listed in Chapter 2.1 of Topical Report DPC-NE-2009/DPC-NE-2009P,
Duke Power Company Westinghouse Fuel Transition Report).
ZIRLOTM cladding has been extensively used in
Westinghouse nuclear reactors. The changes proposed in this LAR are
necessary to ensure that the performance of the fission product
barriers (cladding) will not be impacted following the replacement
of one fuel design for another. No safety margin will be
significantly impacted.

The NRC staff reviewed the licensee's analysis, and agrees 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: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Project Director: Herbert N. Berkow.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: June 18, 1996. This notice supersedes
the notice published on July 31, 1996 (61 FR 40015) in its entirety.
Description of amendment request: For Beaver Valley Power Station,
Unit No. 1 (BVPS-1) only, the proposed amendment would revise Technical
Specification (TS) 4.4.5 and associated Bases; the Bases for TS 3/
4.4.6.2 would also be revised. The proposed changes are editorial in
nature and are intended to provide consistency between the TSs and
associated Bases. Index page XIX would be revised to reflect the
revision of page numbers for TS Tables 4.4-1 and 4.4-2 due to shifting
of text.
For Beaver Valley Power Station, Unit No. 2 (BVPS-2) only, the
proposed amendment would implement a voltage-based repair criteria for
steam generator tubes similar to the changes approved for BVPS-1 by
License Amendment No. 198. The proposed changes are intended to reflect
the guidance provided in NRC Generic Letter 95-05, ``Voltage-Based
Repair Criteria for Westinghouse Steam Generator Tubes Affected by
Outside Diameter Stress Corrosion Cracking.'' The proposed changes
would revise TSs 4.4.5 and 3.4.6.2 and associated Bases. TS Table 4.4-2
would be revised to reference TS 6.6 for reporting requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Tube burst criteria are inherently satisfied during normal
operating conditions due to the proximity of the tube support plate
(TSP). Test data indicates that tube burst cannot occur within the
TSP, even for tubes which have 100% throughwall electric discharge
machining notches, 0.75 inch long, provided that the TSP is adjacent
to the notched area. Since tube-to-TSP proximity precludes tube
burst during normal operating conditions, use of the criteria must
retain tube integrity characteristics which maintain a margin of
safety of 1.43 times the bounding faulted condition, main steamline
break (MSLB) pressure differential. The Regulatory Guide (RG) 1.121
criterion requiring maintenance of a safety factor of 1.43 times the
MSLB pressure differential on tube burst is satisfied by \7/8\''
diameter tubing with bobbin coil indications with signal amplitudes
less than 8.6 volts, regardless of the indicated depth measurement.
The upper voltage repair limit (VURL) will be
determined prior to each outage using the most recently approved NRC
database to determine the tube structural limit (VSL).
The structural limit is reduced by allowances for nondestructive
examination (NDE) uncertainty (VNDE) and growth
(VGR) to establish VURL. Using the Generic
Letter (GL) 95-05 NDE and growth allowances for an example, the NDE
uncertainty component of 20% and a voltage growth allowance of 30%
per full power year can be utilized to establish a
VURL of 5.7 volts. The 20% NDE uncertainty represents a
square-root-sum-of-the-squares (SRSS) combination of probe wear
uncertainty and analyst variability. The degradation growth
allowance should be an average growth rate or 30% per effective full
power year, whichever is larger.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated MSLB
outside of containment but upstream of the main steam isolation
valve (MSIV) represents the most limiting radiological condition
relative to the plugging criteria. In support of implementation of
the revised plugging limit, analyses will be performed to determine
whether the distribution of cracking indications at the tube support
plate intersections during future cycles are projected to be such
that primary-to-secondary leakage would result in postulated site
boundary and control room doses exceeding 10 CFR 100, 10 CFR 50
Appendix A, and GDC-19 [General Design Criterion-19] requirements,
respectively. A separate calculation has determined the maximum
allowable MSLB leakage limit in a faulted loop. This limit was
calculated using the technical specification reactor coolant system
(RCS) Iodine-131 activity level of 1.0 microcuries per gram dose
equivalent Iodine-131 and the recommended Iodine-131 transient
spiking values consistent with NUREG-0800. The projected MSLB
leakage rate calculation methodology prescribed in Section 2.b of GL
95-05 will be used to calculate the end-of-cycle (EOC) leakage.
Projected EOC voltage distribution will be developed using the most
recent EOC eddy current results and considering an appropriate
voltage measurement uncertainty. The log-logistic probability of

[[Page 64110]]

leakage correlation will be used to establish the MSLB leakrate used
for comparison with the faulted loop allowable limit. Therefore, as
implementation of the voltage-based repair criteria does not
adversely affect steam generator tube integrity and implementation
will be shown to result in acceptable dose consequences, the
proposed amendment does not result in any increase in the
probability or consequences of an accident previously evaluated in
the Updated Final Safety Analysis Report (UFSAR).
The proposed changes to the BVPS-1 Index, Specifications and
associated Bases and the proposed change to BVPS-2 Table 4.4-2 are
editorial in nature. Therefore, these changes do not involve an
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Implementation of the proposed steam generator tube voltage-
based repair criteria does not introduce any significant changes to
the plant design basis. Use of the voltage-based repair criteria
does not provide a mechanism which could result in an accident
outside of the region of the tube support plate elevations as no
outside diameter stress corrosion cracking (ODSCC) is occurring
outside the thickness of the tube support plates. Neither a single
or multiple tube rupture event would be expected in a steam
generator in which the plugging limit has been applied (during all
plant conditions).
Duquesne Light Company will implement a maximum primary-to-
secondary leakage rate limit of 150 gpd [gallons per day] per steam
generator to help preclude the potential for excessive leakage
during all plant conditions. The RG 1.121 criterion for establishing
operational leakage rate limits that require plant shutdown are
based upon leak-before-break considerations to detect a free span
crack before potential tube rupture during faulted plant conditions.
The 150 gpd limit provides for leakage detection and plant shutdown
in the event of the occurrence of an unexpected single crack
resulting in leakage that is associated with the longest permissible
crack length. RG 1.121 acceptance criteria for establishing
operating leakage limits are based on leak-before-break
considerations such that plant shutdown is initiated if the leakage
associated with the longest permissible crack is exceeded.
The single through-wall crack lengths that result in tube burst
at 1.43 times the MSLB pressure differential and the MSLB pressure
differential alone are approximately 0.57 inch and approximately
0.84 inch, respectively. A leak rate of 150 gpd will provide for
detection of approximately 0.41 inch long cracks at nominal leak
rates and approximately 0.62 inch long cracks at the lower 95%
confidence level leak rates. Since tube burst is precluded during
normal operation due to the proximity of the TSP to the tube and the
potential exists for the crevice to become uncovered during MSLB
conditions, the leakage from the maximum permissible crack must
preclude tube burst at MSLB conditions. Thus, the 150 gpd limit
provides for plant shutdown prior to reaching critical crack lengths
for MSLB conditions using the lower 95% leakrate data. Additionally,
this leak-before-break evaluation assumes that the entire crevice
area is uncovered during blowdown. Partial uncovery will provide
benefit to the burst capacity of the intersection. Analyses have
shown that only a small percentage of the TSPs are deflected greater
than the TSP thickness during a postulated MSLB.
As steam generator tube integrity upon implementation of the
voltage-based repair criteria continues to be maintained through
inservice inspection and primary-to-secondary leakage monitoring,
the possibility of a new or different kind of accident from any
accident previously evaluated is not created.
The proposed change to BVPS-1 Index, Specifications and
associated Bases and the proposed change to BVPS-2 Table 4.4-2 are
editorial in nature. These changes do not change the performance of
plant systems, plant configuration or method of operating the plant.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The use of the voltage-based repair criteria at BVPS-2 maintains
steam generator tube integrity commensurate with the criteria of RG
1.121. This guide describes a method acceptable to the Commission
for meeting GDCs 14, 15, 30, 31, and 32 by reducing the probability
or the consequences of steam generator tube rupture. This is
accomplished by determining the limiting conditions of degradation
of steam generator tubing, as established by inservice inspection,
for which tubes with unacceptable cracking should be repaired or
removed from service. Upon implementation of the proposed criteria,
even under the worst case conditions, the occurrence of ODSCC at the
tube support plate elevations is not expected to lead to a steam
generator tube rupture event during normal or faulted plant
conditions. The EOC distribution of crack indications at the tube
support plate elevations will be confirmed to result in acceptable
primary-to-secondary leakage during all plant conditions and that
radiological consequences remain within the licensing basis.
In addressing the combined effects of loss-of-coolant-accident
(LOCA) + safe shutdown earthquake (SSE) on the steam generator
component (as required by GDC 2), it has been determined that tube
collapse may occur in the steam generators at some plants. This is
the case as the tube support plates may become deformed as a result
of lateral loads at the wedge supports at the periphery of the plate
due to the combined effects of the LOCA rarefaction wave and SSE
loadings. Then, the resulting pressure differential on the deformed
tubes may cause some of the tubes to collapse. There are two issues
associated with steam generator tube collapse. First, the collapse
of steam generator tubing reduces the RCS flow area through the
tubes. The reduction in flow area increases the resistance to flow
of steam from the core during a LOCA which, in turn, may potentially
increase peak clad temperature. Second, there is a potential that
partial through-wall cracks in tubes could progress to complete
through-wall cracks during tube deformation or collapse.
The results of an analysis using the larger break inputs show
that the LOCA loads were found to be of insufficient magnitude to
result in steam generator tube collapse or significant deformation.
Since the leak-before-break methodology is applicable to the reactor
coolant loop piping, the probability of breaks in the primary loop
piping is sufficiently low that they need not be considered in the
structural design of the plant. The limiting LOCA event becomes the
pressurizer spray line break. Analysis results have demonstrated
that no tubes were subject to deformation or collapse. No tubes have
been excluded from application of the subject voltage-based steam
generator tube repair criteria.
Addressing RG 1.83 considerations, implementation of the
voltage-based repair criteria is supplemented by: enhanced eddy
current inspection guidelines to provide consistency in voltage
normalization, the bobbin coil inspection will include 100% of the
hot-leg TSP intersections and cold-leg intersections down to the
lowest cold-leg TSP with known ODSCC, the determination of the TSPs
having ODSCC will be based on the performance of at least 20% random
sampling of tubes inspected over their full length, and rotating
pancake coil inspection requirements for the larger indications left
inservice to characterize the principal degradation as ODSCC.
As noted previously, implementation of the tube support plate
intersection voltage-based repair criteria will decrease the number
of tubes which must be repaired. The installation of steam generator
tube plugs reduces the RCS flow margin. Thus, implementation of the
voltage-based repair criteria will maintain the margin of flow that
would otherwise be reduced in the event of increased tube plugging.
The proposed change to the BVPS-1 Index, Specifications and
associated Bases and the proposed change to BVPS-2 Table 4.4-2 are
editorial in nature. These changes will not reduce the margin of
safety because they have no impact on any safety analysis
assumptions.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin with respect to plant safety as defined in the UFSAR or any
BASES of the plant technical specifications.

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: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts &

[[Page 64111]]

Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: October 15, 1998.
Description of amendment request: The proposed amendment would make
several changes that are administrative in nature. The changes would
(1) make editorial changes to delete obsolete material or material
adequately described elsewhere, change action statement numbers, update
the technical specification (TS) index pages, and make changes to be
consistent with the guidance of the improved standard technical
specifications (ISTS); (2) delete reporting requirements that duplicate
reporting requirements contained in 10 CFR; and (3) relocate the
requirement for meteorological monitoring instrumentation from the TS
to the Licensing Requirements Manual.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
a. This change deletes an expired Unit 1 license condition and a
Unit 2 license requirement that is not required since it is
redundant to the reporting requirements addressed in 10 CFR 50.73.
Deleting these requirements does not involve any increase in the
probability or consequences of an accident previously evaluated.
b. The reference to Specification 3.0.6 was omitted from
Specification 3.0.1 in Unit 1 Amendment 213 and Unit 2 Amendment 90
and is being added to 3.0.1 to be consistent with the Improved
Standard Technical Specifications of NUREG 1431. This does not
involve any increase in the probability or consequences of an
accident previously evaluated.
c. The Core Alteration definition has been updated to be
consistent with the regulations and ISTS. The Offsite Dose
Calculation Manual (ODCM) definition has been updated to be
consistent with the change to Administrative Control 6.9.3. The
Members of the Public definition has been changed to be consistent
with 10 CFR 20.1003. This does not involve any increase in the
probability or consequences of an accident previously evaluated.
d. Changing Table 3.3-6 Action Statement 36 to Action Statement
35 is an editorial change to eliminate redundant use of action
statement numbers. This does not involve any increase in the
probability or consequences of an accident previously evaluated.
e. The technical specification index is being revised to address
removal of the Meteorological Monitoring specification and title and
page number changes to the administrative control reporting
requirements section. The Meteorological Monitoring specification is
being relocated to the Licensing Requirements Manual (LRM).
Relocating the Meteorological Monitoring requirements is in
accordance with the guidance in the Commission's Final Policy
Statement and revisions to 10 CFR 50.36 on the content of the
technical specifications and the ISTS. The Meteorological Monitoring
requirements do not meet any of the criteria, 1 thru 4 of 10 CFR
50.36 and can, therefore, be relocated from the Technical
Specifications to the LRM. These changes do not involve any increase
in the probability or consequences of an accident previously
evaluated.
f. The exclusion area boundary is adequately described in each
unit's UFSAR [Updated Final Safety Analysis Report], therefore,
design feature 5.1 Site Location is also being modified by deleting
the description of the exclusion area boundary. This does not
involve any increase in the probability or consequences of an
accident previously evaluated.
g. The change to refer to the Unit 1 Overpressure Protection
System (OPPS) enable temperature in Specification 3.4.9.3 in lieu of
specifying 275 deg.F was evaluated and found acceptable in the
request for approval of Amendment 160. The deletion of the asterisk
in Unit 2 Specification 3.9.8.1 was justified as part of the request
for approval of Amendment 25. The inadvertent omission of the ACTION
to take in the case that the temperature of the steam generator is
precisely 50 deg.F above the cold leg temperature is being
corrected. The cases of greater than and less than 50 deg.F are
already included. These are editorial changes that do not involve
any increase in the probability or consequences of an accident
previously evaluated.
h. The administrative control reporting requirements have been
modified to incorporate various ISTS requirements. This requires
changing titles and eliminating requirements addressed elsewhere,
removing reference to deleted sections, and replacing reference to
the administrative control section reporting requirements in various
specifications with reference to 10 CFR 50.4. The 1993 NRC final
policy statement set forth the criteria for determination of those
requirements to be included in TS. The reporting requirements being
removed from the TS do not meet the criteria for inclusion in the
TS; therefore, the reporting requirements have been modified to
reflect those requirements provided in the ISTS. These are editorial
changes that do not involve any increase in the probability or
consequences of an accident previously evaluated.
i. The Technical Specification index has been modified to
address the revised pages.
These changes have been determined to be editorial and
administrative in nature, and as such, would not affect any accident
assumptions or radiological consequences of an accident. Therefore,
the proposed changes would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The editorial changes, the elimination of reporting requirements
which duplicate 10 CFR requirements and administrative improvements
to incorporate the ISTS requirements are all changes that are
administrative in nature. The proposed changes will not affect any
plant system or structure, nor will they affect any system
functional or operability requirements. Consequently, no new failure
modes are introduced as a result of the proposed changes. Therefore,
the proposed change will not create the possibility of a new or
different type of accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed amendment modifies reporting requirements and
incorporates associated editorial changes that do not impact the
UFSAR design basis or accident analyses assumptions. This change
does not introduce any new operational modes or physical
modifications to the plant; therefore, no action will occur that
will involve a significant reduction in a margin of safety. In
addition, the proposed change does not affect radiological release
limits, monitoring equipment or operating practices. 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: September 23, 1998.
Description of amendment request: The proposed amendment would
change Division III battery specific gravity acceptance criteria
outlined in River Bend Station (RBS) Technical Specifications (TS). The
change is required as a result of battery system design modifications
which are scheduled to be implemented in April 1999 during refueling
outage (RF) RF-8. During this time, the current Division III

[[Page 64112]]

battery will be replaced. The new battery, which also will have a
greater capacity rating, will be supplied with a nominal specific
gravity of 1.215 at 77 deg.F in contrast to the existing Division III
battery supplied with a nominal specific gravity of 1.210 at 77 deg.F.
Since TS Section 3.8.6, Table 3.8.6-1 values for specific gravity are
based on the manufacturer's nominal specific gravity, these values will
need to be updated.
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. This request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The system loads, voltage requirements, and inrush currents have
been calculated in accordance with IEEE Std. 485, ``IEEE Recommended
Practice for Sizing Large Lead Storage Batteries for Generating
Stations and Substations.'' To support these design requirements at
a capacity of 80%, a new battery must be installed. The nominal
specific gravity of the new battery, as provided by the manufacturer
of the battery, is 1.215 at 77 deg.F.
A review of USAR Chapter 15, including Appendix 15A, was
conducted to determine what accidents, if any, may be impacted by
the proposed change to the Division III battery specific gravity.
USAR Sections 15.2, ``Increase in Reactor Pressure;'' 15.3,
``Decrease in Reactor Coolant System Flow Rate;'' and Section 15.6,
``Decrease in Reactor Coolant Inventory'' discuss accidents that
involve the initiation of HPCS when reactor vessel level drops to
the initiation point. The function of the HPCS System is to mitigate
the consequences of an accident (i.e., to maintain reactor vessel
coolant inventory after small breaks which do not depressurize the
reactor vessel, or provide spray cooling heat transfer following
larger breaks, Ref. USAR Section 6.3.1.2.1). The function of the
Division III 125 Vdc Power System is to provide a highly reliable,
continuous, and independent source of control and motive power for
the HPCS System logic, HPCS diesel generator set control and
protection, and all Division III related control (Ref. USAR Section
8.3.2.2.1). This is a support function for the HPCS System.
USAR Section 15.5, ``Increase In Reactor Coolant Inventory,''
postulates an inadvertent HPCS actuation resulting from operator
error. The proposed changes to the Division III battery specific
gravity cannot result in an inadvertent HPCS actuation/injection.
The proposed changes to the allowable specific gravity values
provided in Technical Specification 3.8.6 are in agreement with the
manufacturer's nominal specific gravity. The revision simply ensures
that the battery has sufficient capacity to meet the energy
requirements of its critical loads. The proposed change does not
create any new internally generated missiles, nor does it affect the
High Energy Line Break Analysis or any other accident described in
Chapter 15 of the USAR. Neither the function nor the operation of
the Division III battery is impacted by the proposed change.
The replacement Division III battery will be supplied by the
manufacturer with a nominal specific gravity of 1.215 at 77 deg.F.
The battery manufacturer's rated performance is based on the
specific gravity of the battery being maintained near the nominal
specific gravity. Since the Division III design basis calculation
depends on the battery manufacturer's rated performance, battery
parameters upon which that performance is based must be monitored.
The current Technical Specification values for specific gravity are
based upon a nominal specific gravity of 1.210 at 77 deg.F. The
proposed values accurately reflect the manufacturer's nominal
specific gravity. Testing the Division III battery to the proposed
values provides assurance that the HPCS functions supported by the
125 Vdc System will not be adversely affected by the Division III
battery.
The proposed changes will not affect failure modes of existing
equipment. The proposed changes do not affect the ability of any
structures, systems or components to perform their safety functions.
Therefore, no undue risk to the health and safety of the public has
been created by the proposed changes, nor is there any change in the
radiological consequences at the site boundary.
By incorporating the correct value for battery specific gravity
verification in Table 3.8.6-1, the Technical Specifications will
accurately reflect the new design basis value for the Division III
battery specific gravity. This change allows the performance of the
Division III battery to be verified against the correct design basis
value, thus providing assurance that the Division III 125 Vdc power
system function will remain as assumed in the accident analysis.
Therefore, the proposed change cannot affect any accidents
previously evaluated (probability or consequences). Consequently,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. This request does not create the possibility of occurrence of
a new or different kind of accident from any previously evaluated.
Since a battery's capacity decreases as specific gravity
decreases below the manufacturer's nominal value, monitoring the
battery's specific gravity is one means of ensuring that the battery
will adequately supply the minimum energy required to support the
system function assumed in the accident analysis.
All safety systems will continue to function as originally
designed. The subject equipment will not function in a manner
different than described in USAR Section 8.3.2.2. The functional and
performance requirements of the Division III 125 Vdc System and its
associated interfaces have not been altered. The proposed change
simply ensures that the HPCS battery performance is verified against
the correct design basis value. This value provides assurance that
the HPCS System functions will not be adversely affected by the
capacity of the battery. Therefore, the proposed changes do not
create the possibility of occurrence of a new or different kind of
accident from any previously evaluated.
3. This request does not involve a significant reduction in a
margin of safety.
This proposed change updates the acceptance criteria of the
current specific gravity for the Division III battery. This
acceptance criteria is in accordance with manufactures
recommendations. The design and license basis for the Division III
systems and functions remain unchanged and the battery will continue
to supply the 125 Vdc loads necessary to support these functions.
This value will reflect the manufacturer's nominal specific gravity
for the Division III battery. With the system functions supported as
assumed in the accident analyses, the margin to safety remains
unchanged.
As a result, the proposed change does not involve a significant
reduction in a margin to safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92 are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW, Washington, DC 20005.
NRC Project Director: John N. Hannon.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: October 8, 1998.
Description of amendment request: The proposed amendment would
implement Boiling Water Reactor Owners Group (BWROG) Enhanced Option I-
A (EIA) Reactor Stability Long Term Solution as documented in NEDO-
32339, Revision 1, ``Reactor Stability Long-Term Solution, Enhanced
Option I-A.'' The EIA long term solution has been accepted by the NRC
in Safety Evaluation Report, ``Reactor Stability Long-Term Solution,
Enhanced Option I-A Generic Technical Specifications (TS), NEDO-32339,
Supplement 4.''
The proposed changes to the RBS TS will enable the full
implementation of the Enhanced Option I-A (EIA) long term solution to
the neutronic/thermal hydraulic instability issue. Specifically, the
proposed change deletes the limits

[[Page 64113]]

on power and flow conditions associated with the implementation of the
guidance in General Electric Service Information Letter #380, Revision
1, ``BWR Core Thermal Hydraulic Stability'' (current TS 3.4.1, Figure
3.4.1-1 and RBS plant procedures), adds new specifications, to
establish limits for Fraction of Core Boiling Boundary (FCBB) and the
Period Based Detection System (PBDS), modifies the RPS instrumentation
specification and the description of the contents of the Core Operating
Limits Report (COLR) in current TS 5.6.5. The two new specifications
require maintaining stability control and the availability of a
stability detection system during operation in defined regions of the
power and flow operating domain.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed amendments do no involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendments allow the implementation of the Enhanced
Option I-A (EIA) long term solution to the neutronic/thermal
hydraulic instability issue. Current TS restrictions on power and
flow conditions, number of operating recirculation loops, and
operator actions implemented to reduce the probability of neutronic/
thermal hydraulic instability are eliminated and new stability
requirements consistent with NEDO-32339-A, Supplement 4, Revision 1,
are imposed.
While the proposed amendments permit operation in regions of the
power and flow operating domain postulated to be susceptible to
neutronic/thermal hydraulic instability, the implementation of the
EIA solution ensures there is not a significant increase in the
probability or consequences of an accident previously evaluated.
Operation in these regions does not increase the probability of
occurrence of initiators and precursors of other previously analyzed
accidents. The proposed amendments permit the implementation of the
features of the EIA solution which prevent neutronic/thermal
hydraulic instability. The features include pre-emptive reactor
scram upon entry into the regions of the power and flow operating
domain most susceptible to neutronic/thermal hydraulic instability--
the Exclusion Region. The EIA solution prevents neutronic/thermal
hydraulic instability during operation in regions of the power and
flow operating domain previously excluded from operation and
therefore does not significantly increase the probability of a
previously analyzed accident.
The EIA solution also requires implementation of stability
control prior to entry into a region of the power and flow operating
domain which is potentially susceptible, in the absence of stability
control, to neutronic/thermal hydraulic instability. The modified
rod block functions providing the restricted region entry alarm
(RREA), boiling boundary limits, and PBDS functions are required on
entry into the Restricted Region of the power to flow map. The
boiling boundary limits, and Period Based Detection System (PBDS)
functions are required on entry into the Monitored Region of the
power to flow map. The EIA solution prevents or allows for detection
and suppression of neutronic/thermal hydraulic instability during
operation in these regions of the power and flow operating domain.
The EIA solution includes restrictions on power and flow
conditions and actions associated with the modified APRM flow biased
scram and RREA functions. Required actions include adherence to the
boiling boundary limit stability control prior to entry and during
operation in the region of the power and flow operating domain which
is potentially susceptible to neutronic/thermal hydraulic
instability--in the absence of stability control. In addition, the
proposed amendments require operator actions based upon control room
indications generated by a new PBDS. The PBDS is designed to provide
alarm indication that conditions consistent with a significant
degradation in the stability performance of the reactor have
occurred and the potential for imminent onset of neutronic/thermal
hydraulic instability may exist. The PBDS also provides analog
indication of the highest and second highest successive period
confirmation count for all of the LPRMs monitored. This provides the
plant operators with continuous indication of reactor stability
operating conditions. The PBDS system provides indication only and
does not affect plant structures, systems, or components in any way
that could increase the probability or consequences of an accident.
Rather, the improved control room indications provide the operator
with more accurate and timely information.
The EIA solution allows for the ``Setup'' of APRM flow biased
scram and control rod block function. The EIA solution requires
adherence to certain boiling boundary limit stability controls prior
to selection by the operator of APRM flow biased scram and control
rod block function ``Setup'' setpoints. This ``Setup'' function
allows operation in a region of the power and flow operating domain
potentially susceptible to neutronic/thermal hydraulic instability
provided the additional limits of the flow control boiling boundary
(FCBB) and PBDS are met. After exiting the region requiring the
stability control to be met, the setpoints can be manually reset to
their normal values. Stability controls are required to be in place
when setpoints are ``Setup''. As a backup EIA feature, the APRM flow
biased setpoints automatically reset to their normal values above a
pre-determined flow condition. This automatic reset to the more
conservative setpoints ensures that the pre-emptive reactor scram
will prevent operation as a result of an anticipated operational
occurrence in the region most susceptible to neutronic/thermal
hydraulic instability should the operator not select the more
conservative setpoints appropriate for operation following exit from
the region requiring stability control. The FCBB, PBDS, and
automatic reset of the APRM flow biased scram and control rod block
function ``setup'' setpoints allow for the use of the ``setup''
feature and help ensure that there is not an increase in the
probability or consequences of an accident.
Operation in the regions of the power and flow operating domain
excluded by current TS 3.4.1 and Figure 3.4.1-1 can occur as a
result of anticipated operational occurrences. In the absence of
operator actions the severity of these anticipated operational
occurrences may increase due to the potential occurrence of
neutronic/thermal hydraulic instability as a result of operation in
these regions. Upon entry, as a result of an anticipated operational
occurrence, into the region most susceptible to neutronic/thermal
hydraulic instability the pre-emptive reactor scram prevents
neutronic/thermal hydraulic instability. Therefore, the consequences
of an accident do not significantly increase while operating with
stability control in place.
The required EIA features is designed to limit possible
neutronic/thermal hydraulic instabilities and to detect and suppress
further neutronic/thermal hydraulic instabilities. These features
include: a pre-emptive automatic scram, the control rod block and
alarms associated with entry into the region susceptible to
neutronic/thermal hydraulic instabilities, automatic reset of APRM
flow biased setpoints, PBDS, FCBB, and the required operator
actions, including manual reactor scram. Therefore, the proposed
amendments prevent the occurrence of neutronic/thermal hydraulic
instability during operation or as a consequence of an anticipated
operational occurrence and do not significantly increase the
consequences of any previously analyzed accident.
2. The proposed amendments do not create the possibility of a
new or different kind of accident from any previously evaluated.
The proposed amendments eliminate existing restrictions on power
and flow conditions and impose alternative restrictions which permit
the implementation of the EIA long term stability solution. The
current restrictions on the power and flow conditions do not prevent
entry into regions of the power and flow operating domain most
susceptible to neutronic/thermal hydraulic instability and therefore
the possibility of neutronic/thermal hydraulic instability exists in
the absence of operator action. The required features of the EIA
solution implement a pre-emptive scram upon entry into the region
most susceptible to neutronic/thermal hydraulic instability, without
operator action. The accessible operating domain allowed by the
proposed amendments is essentially a subset of the power and flow
operating domain currently allowed. Initial conditions are bounded
by the current initiators and precursors of accidents and
anticipated operational occurrences. Accordingly, no new accident of
initiator is present. Therefore, the proposed amendments do not
create the possibility of a new or different kind of accident from
that previously evaluated.
Concurrent with the implementation of the proposed amendments, a
modified Flow

[[Page 64114]]

Control Trip Reference (FCTR) card, EIA FCTR card, and a new Period
Based Detection System (PBDS) will be installed as required by the
EIA solution. The function of the EIA FCTR card is to aid the
operator in the identification of entry into regions of the power
and flow operating domain potentially susceptible to neutronic/
thermal hydraulic instability in the absence of stability controls
and to initiate a pre-emptive scram upon entry into the regions most
susceptible to neutronic/thermal hydraulic instability. This is
accomplished by altering the existing values of setpoints of the
APRM flow biased scram and the control rod block functions generated
by the EIA FCTR card.
The design of the EIA digital FCTR card is a functional
equivalent of the original analog FCTR card. The Failure Modes and
Effects Analysis (FMEA) for the card detailed in NEDC-32339P-A
Supplement 2 found no single failure that would increase the
consequences of an accident. The EIA FCTR card maintains the
original basis for the NMS interface functions of the analog FCTR
card it replaces. The plant specific environmental conditions
(temperature, humidity, pressure, seismic, and electromagnetic
compatibility) have been confirmed to be enveloped by the
environmental qualification values for the EIA FCTR cards.
Therefore, the potential for spurious scrams or common mode failures
induced by environmental effects (e.g., electromagnetic
interference) is considered negligible. The installation of the EIA
FCTR card will therefore not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The function of the PBDS is to provide the operator with an
indication that conditions consistent with a significant degradation
in the stability performance of the reactor has occurred and the
potential for imminent onset of neutronic/thermal hydraulic
instability may exist. This is accomplished by the installation of a
new PBDS card in the Neutron Monitoring System in accordance with
NRC approved BWROG and GE design. The PBDS card takes inputs from
individual local power range monitors and provides analog indication
of the highest and second highest successive period confirmation
count, provides a Hi DR and Hi-Hi DR alarm, and INOP status
indication to the operator in the control room. These displays can
not create the possibility of a new or different kind of accident
from any accident previously evaluated. The plant specific
environmental conditions (temperature, humidity, pressure, seismic,
and electromagnetic compatibility) have been confirmed to be
enveloped by the PBDS environmental qualification values. Therefore,
the installation of the PBDS card will not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendments do not involve a significant
reduction in the margin of safety.
The proposed amendments permit the implementation of the EIA
long term solution to the stability issue. Under certain conditions,
existing BWR designs are susceptible to neutronic/thermal hydraulic
instability. GDC 10 of 10 CFR 50, Appendix A, requires that
specified acceptable fuel design limits not be exceeded during
anticipated operational occurrences. General Design Criterion (GDC)
12 of 10 CFR 50, Appendix A, requires thermal hydraulic instability
to be prevented by design or be readily and reliably detected and
suppressed. When the design of the reactor system does not prevent
the occurrence of neutronic/thermal hydraulic instability,
instability is considered an anticipated operational occurrence. The
proposed amendments and the associated design modifications provide
automatic features and operational information to the Control Room
that replace the existing BWROG Interim Corrective Actions (ICAs).
Thus the EIA solution assures compliance with GDC-10 and GDC 12 by
providing for reliable detection and suppression and by the
prevention of neutronic/thermal hydraulic instability. This
therefore precludes neutronic/thermal hydraulic instability from
becoming a credible consequence of an anticipated operational
occurrence. As a result the margins of safety are maintained.
Analyses performed by the BWROG indicate that neutronic/thermal
hydraulic instability induced power oscillations could result in
conditions exceeding the MCPR SL prior to detection and suppression
by the current design of the Neutron Monitoring System and Reactor
Protection System. To ensure compliance with GDC 12 the BWROG
developed Interim Corrective Actions (ICAs) to enhance the
capability of the operator to readily and reliably detect and
suppress neutronic/thermal hydraulic instability. The BWROG ICAs
also provided additional guidance for monitoring local power range
monitors beyond the requirements of current TS 3.4.1 to ensure
adequate margin to the onset of neutronic/thermal hydraulic
instability. Reliance on operator actions to comply with GDC 12 was
accepted on an interim basis by the NRC pending final implementation
of a long term solution to the stability issue. The modified design
of the Reactor Protection System (APRM flow biased scram) and
stability control prior to entry into a region of the power and flow
operating domain which is potentially susceptible, in the absence of
stability control, to neutronic/thermal hydraulic instability
implemented with the EIA solution prevents neutronic/thermal
hydraulic instability. In addition, significant backup protection
features, including the PBDS and specified operator actions, are
required to be implemented. As a result, the margin to the onset of
neutronic/thermal hydraulic instability provided by the existing TS
requirements and BWROG ICAs recommendations is not reduced by the
implementation of the EIA solution. The EIA solution assures
compliance with GDC 12 by the prevention of neutronic/thermal
hydraulic instability and therefore precludes neutronic/thermal
hydraulic instability from becoming a credible consequence of an
anticipated operational occurrence. The consequences of anticipated
operational occurrences will not increase and the margin to the MCPR
SL will not decrease upon implementation of the EIA solution.
Therefore, the proposed amendment does not involve a reduction in a
margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW, Washington, DC 20005.
NRC Project Director: John N. Hannon.

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

Date of amendment request: October 1, 1998.
Description of amendment request: The proposed change modifies
Technical Specification (TS) 3.3.3.7.3 and Surveillance Requirement
4.3.3.7.3 for the broad range gas detection system. A change to
Technical Specification Basis 3/4.3.3.7 has been included to support
this change. This change to the TS is necessary for the installation of
a new, more reliable broad range gas detection system.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The broad range gas detection system has no effect on the
accidents analyzed in Chapter 15 of the Final Safety Analysis
Report. It's only effect is on habitability of the control room,
which will be enhanced by installation of the new monitoring system
and this change to the Technical Specifications. Qualitative
analysis based on a quantitative risk assessment has shown that the
impact on operator incapacitation and subsequent core damage risk of
the periodic automatic background/reference spectrum check is
negligible and that the probability of malfunction of the BRGMs due
to a slowly increasing toxic chemical concentration is negligible.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of

[[Page 64115]]

accident from any accident previously evaluated?
Response: No.
The proposed Technical Specification change in itself does not
change the design or configuration of the plant. The new broad range
toxic gas monitoring system performs the same function as the old
system, but it accomplishes this function with increased
reliability.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The broad range gas detection system has no effect on a margin
of safety as defined by Section 2 of the Technical Specifications.
Its only effect is on habitability of the control room, which will
be enhanced by installation of the new monitoring system and this
change to the Technical Specifications.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400
L Street NW, Washington, DC 20005-3502.
NRC Project Director: John N. Hannon.

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: July 30, 1998 (LAR-222).
Description of amendment request: The proposed amendment will
change the Improved Technical Specifications (ITS) to add a new
Required Action for the existence of breaches in the Control Complex
Habitability Envelope (CCHE) that are in excess of allowances. A new
surveillance requirement for the performance of a periodic integrated
leak test of the CCHE boundary on a 24-month frequency would also be
added. Changes to the current Ventilation Filter Test Program (VFTP)
are proposed to adopt current standards for laboratory testing, change
acceptable values of control room emergency ventilation flow rate and
filter differential pressure, and add the Auxiliary Building
Ventilation Exhaust Filters to the VFTP. Conforming changes to the ITS
Bases are also included.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.

1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated. The Control Room Emergency Ventilation System (CREVS) and
the Control Complex Habitability Envelope (CCHE) are designed to
limit the radiation dose to the control room operating staff
following a design basis accident. Since these systems are only
effective in limiting dose following an accident, the existence of
limited breaches in the CCHE, the performance of periodic leak
tests, and changes to the Ventilation Filter Test Program (VFTP)
would not increase the probability of occurrence of any evaluated
event. The features of the CREVS and the Control Complex emergency
filters, or the CCHE have no direct function in mitigating the
offsite consequences of any evaluated accident. The Auxiliary
Building exhaust filters are not credited with reducing offsite
doses, however, if available would filter releases from the
Auxiliary Building. Adding them to the VFTP will not increase the
consequences calculated for any evaluated accident.
The proposed changes are consistent with the revised control
room operator dose calculations as presented in the Control Room
Habitability Report dated July 1998. Since all calculated doses are
within 10 CFR Part 50, Appendix A GDC 19 limits there is no
significant increase in consequences.
It is conceivable that the existence of additional breaches in
the CCHE could result in an increase in operator dose, however the
low probability of a catastrophic reactor accident, the relatively
short time allowed for breaches to be open in excess of approved
dose calculation assumptions, and the ability to close breaches
expeditiously makes the risk increase insignificant.
The changes to the ITS Bases improve information on the
operation and function of CREVS, and establish that CREVS
operability is dependent on maintaining CCHE integrity. The
inclusion of this information reinforces the importance of
maintaining the CCHE boundary, and will help to ensure the CREVS is
capable of performing its intended safety function.
The Control Room Habitability Report, dated July 1998, provided
with this LAR presents the methodology used and the results of the
operator dose calculations for the Maximum Hypothetical Accident,
toxic gas release, and other design basis accidents. The report
provides the information needed for NRC review of LAR 222, Revision
I and the associated unreviewed safety question. This evaluation
concludes that the current level of CCHE integrity provides adequate
protection for the control room operator.
Based on the foregoing, the proposed amendment does not
significantly increase the probability or consequence of an accident
previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Neither performance of periodic CCHE leak tests nor changes to
the existing VFTP can create the possibility of a new or different
kind of accident. During the period of time when CCHE breaches are
greater than the design calculation, there exists the possibility
that control room dose from an analyzed accident may be greater than
specified in General Design Criterion 19. This condition will not
however create the possibility of a new or different kind of
accident. Since CREVS and the emergency filtration units function to
provide protection following a radiological accident the changes
proposed to improve their performance cannot create a new or
different kind of accident. Changes to the Bases to provide better
information on determining CREVS and CCHE operability cannot create
the possibility of a new or different kind of accident.
3. Does not involve a significant reduction in a margin of
safety.
The proposed amendment does not involve a significant reduction
in a margin of safety. Neither performance of periodic CCHE leak
tests nor changes to the existing VFTP can create a reduction in the
margin of safety. The changes to both of these programs will result
in improved assurance that the CREVS and CCHE will perform as
expected if required for operator protection. Changes to the Bases
of the CREVS Technical Specification which clarify the conditions
necessary for operability will improve understanding of the
requirements for maintaining control room habitability, and will not
create a reduction in the margin of safety. The existence of
additional breaches in the CCHE for short periods of time does not
significantly increase the risk of control room operator exposure to
airborne radioactivity or toxic gas. There is no change in the risk
to the public since the CCHE has no direct function in mitigating
the offsite consequences of any evaluated accident. Any event that
could create these exposures has an extremely low probability of
occurrence, and while the potential for higher operator exposure
exists if additional breaches are open, the short duration allowed
would not significantly increase the risk of exposure. Therefore,
for the reason stated above the existing margin of safety would not
be reduced.

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

[[Page 64116]]

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: Frederick J. Hebdon.

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: September 30, 1998 (LAR-238).
Description of amendment request: The proposed amendment will
correct the reactor coolant system (RCS) leakage detection capability
of the Reactor Building atmosphere gaseous radioactivity monitor
described in the Improved Technical Specification Bases and the Final
Safety Analysis Report (FSAR). These documents currently identify that
the gaseous radioactivity monitor is capable of detecting a one gallon
per minute (gpm) RCS leak within one hour. The licensee has determined
that it would take approximately 14 hours for this instrument to detect
a one gpm RCS leak using currently accepted assumptions. The capability
of other monitors to detect a one gpm RCS leak within one hour is not
affected by this change.
The licensee cited several factors which contribute to the
difficulty in reliably detecting RCS leakage increases of one gpm
within one hour using a gaseous radioactivity monitor. These include
the relatively long half-life of Xe-133 (primary nuclide of detection),
fluctuations in background levels of radioactivity, the existence of
minor RCS leaks, improved performance of nuclear fuel, and improved
primary water chemistry control. Based on RCS radioactivity
concentrations assumed in the Environmental Report, half-lives of the
most abundant gaseous nuclides, and background radioactivity levels,
the licensee indicated a one gpm leak can conservatively be detected in
approximately 14 hours by the gaseous monitor. The licensee has
determined that this change to the licensing basis is 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.

1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
No. The function of the RM-A6 gaseous radioactivity monitor is
to detect leakage from the RCS that may develop as a result of a
flaw in a pressure boundary component. The previously identified
capability to detect a one gpm leak within one hour would have
provided an earlier warning of a small RCS leak than the actual
detection capability now identified. However, RCS loss of coolant
accidents evaluated in the FSAR cover the full spectrum of break
sizes up to and including a complete severance of the largest RCS
piping. The results of these analyses demonstrate that the
consequences of such leaks are acceptable.
No other equipment relies on the capability of the RM-A6 gaseous
monitor's ability to detect RCS leakage to perform its function.
Likewise, no accident analyses rely on RCS leak detection for
successful mitigation. Identifying the detector's actual capability
to detect an RCS leak will not increase the probability of
occurrence of an RCS leak. Detection time for an RCS leak was a
consideration in granting a partial exemption to General Design
Criterion 4. However, the capability of the RCS piping to resist
propagation of a flaw from a leak into a break was based on material
fracture analysis and material properties, not on the ability to
detect low levels of leakage.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
No. The function of the RM-A6 gaseous radioactivity monitor is
to detect RCS leakage that may develop from a flaw in a pressure
boundary component. The monitor is a passive component that provides
an indication of possible leakage for further operator evaluation.
Identifying that a longer response time is required for the monitor
to detect a small leak will not create the possibility of a new or
different kind of accident. Existing analyses for small and large
break loss of coolant accidents provide an evaluation of the full
spectrum of RCS break sizes.
3. Involve a significant reduction in a margin of safety.
No. The RM-A6 gaseous radioactivity monitor is included in plant
technical specifications as one of two containment atmosphere RCS
leak detection instruments required to be operable to satisfy a
limiting condition for operation. If the RM-A6 particulate monitor
is not operable, then the response time of the containment
atmosphere monitor will be increased. RCS piping analyses have
demonstrated that the propagation of a small primary loop leak into
a pipe break would not occur rapidly. NRC acceptance of the
applicable analyses included significant safety factors for the
propagation of flaws into pipe breaks which were based on low
probability stress combinations of normal plus safe shutdown
earthquake loads. Considering the actual detection capability of the
RM-A6 gaseous monitor and the existence of other diverse leak
detection capabilities, detection of a leak in a relatively short
period of time is anticipated. In the event an RCS leak developed
into a pipe break, current accident analyses would bound the effects
of the pipe break on and off site. Therefore, the possibility of
increased time to detect an RCS leak does not represent 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: 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: Frederick J. Hebdon.

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: October 16, 1998 (LAR-229).
Description of amendment request: The proposed amendment would
change the Crystal River Unit 3 (CR-3) Final Safety Analysis Report
(FSAR), Improved Technical Specifications (ITS) and ITS Bases to
resolve an Unreviewed Safety Question (USQ). This USQ was created by
changing the normal standby position of valves DHV-34 and DHV-35 (low
pressure injection (LPI) pump suction valves from borated water storage
tank) from normally open to normally closed. Maintaining these valves
normally closed is necessary to ensure assumptions used in fire
protection analyses remain valid. The proposed amendment would also add
new ITS surveillance requirements for verifying on a periodic basis
that the LPI system components and piping, and the building spray
suction piping, are full of water.
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?
Valves DHV-34 and DHV-35 are located in the suction lines
between the borated water storage tank (BWST) and the low pressure
injection (LPI) and building spray (BS) pumps. These valves are
maintained normally closed, and are designed to automatically open
upon receipt of a reactor coolant system (RCS) low-low pressure
signal

[[Page 64117]]

of 500 psig or a reactor building (RB) high pressure signal of 4
psig from the engineered safeguards actuation system (ESAS). The
designed full stroke time of these valves is within the assumptions
of the accident analyses performed for the specific design basis
accidents that require the LPI and/or BS systems for accident
mitigation. This is the original design basis for these valves.
Therefore, the valves are fully capable of performing their intended
safety functions while being maintained normally closed.
The failure of one of these valves to open does not impact the
mitigation of previously analyzed accidents that require the
operation of the LPI and/or BS systems, and cannot increase the
probability of these accidents occurring. No RCS or secondary system
pressure boundaries are compromised, no release paths for
radioactive materials are created, and no challenge to any safety
limit or acceptance limit are created by maintaining these valves
normally closed.
A single, active failure causing one of these valves to fail to
open upon demand would render one train of LPI and BS unavailable
for accident mitigation. However, the accident analyses have already
accounted for the possibility of only one train of LPI and BS being
available, and the consequences of previously evaluated accidents
would therefore remain unchanged.
Undetected voiding in the LPI piping and components, and BS
suction piping, is highly unlikely to occur. Based on the design and
physical layout of the LPI system and BS system, and the monitoring
of the systems performed on a periodic basis, any potential for LPI
piping and components and BS suction piping voiding will be quickly
and easily recognized and corrected. Therefore, since voiding is not
likely to occur, the consequence of previously evaluated accidents
would not be significantly increased by the proposed change.
2. Create the possibility of a new or different kind of accident
from previously evaluated accidents?
Failure of either valves DHV-34 or DHV-35 to open upon demand on
an ESAS signal will not create the possibility of a new or different
kind of accident. The LPI system and BS system are maintained in a
standby condition during normal plant operations, and automatically
actuate only after an accident has occurred to mitigate the effects
of the initiating accident. No RCS or secondary system pressure
boundaries are compromised, no release paths for radioactive
materials are created, and no challenges to any safety limit or
acceptance limit are created by maintaining these valves normally
closed. Additionally, the possibility of undetected voiding in the
LPI piping and components, and BS suction piping, is not likely to
occur by maintaining these valves normally closed. Therefore,
maintaining valves DHV-34 and DHV-35 normally closed will not be an
initiator of a new or different kind of accident from previously
evaluated accidents.
3. Involve a significant reduction in a margin of safety?
Maintaining valves DHV-34 and DHV-35 normally closed will not
create a reduction in the margin of safety. Maintaining valves DHV-
34 and DHV-35 normally closed will ensure the capability to safely
shut down the reactor under certain postulated fire scenarios, but
will result in an extremely small increase in the probability of
failure of one train of LPI and BS to perform its safety functions.
Based on use of the CR-3 Probabilistic Safety Analysis (PSA) model,
and assuming the failure of either valve DHV-34 or DHV-35 to open,
the impact on the core-damage frequency was estimated and determined
to slightly increase from 7.38 E-6 to 7.41 E-6 per year. This
increase (3 E-8 or 0.4%) is in the range considered acceptable in
Regulatory Guide 1.174, ``An Approach for Using Probabilistic Risk
Assessment in Risk-Informed Decisions on Plant-Specific Changes to
the Current Licensing Basis,'' dated July 1998.
Maintaining these valves normally closed will not result in
undetected voiding in the LPI piping and components, and BS suction
piping, as a result of performance of periodic pressure monitoring.
If voiding occurs, the Improved Technical Specifications specify the
actions required to restore the affected systems to operable status,
including correcting the external leakage creating the observed
pressure decay. Therefore, the proposed monitoring will ensure the
margin of safety is not reduced.
Based on these benefits and risks, there is no discernible
change in the risk to the public in mitigating the offsite
consequences of any evaluated accident since the failure of one
train of LPI and/or BS for any reason is bounded by the assumptions
of the accident analyses. Failure of valve DHV-34 or DHV-35 to open
upon demand results in extremely low increases in the potential for
reactor core damage. Therefore, the existing margin of safety will
not be reduced.

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: Frederick J. Hebdon.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: October 15, 1998.
Description of amendment request: The proposed amendment request
would revise the TMI-1 Updated Final Safety Analysis Report (UFSAR)
Chapter 14 postulated accident analysis radiological dose consequences
resulting from application of revised atmospheric dispersion factors
(X/Q) at the Technical Specification Section 5.1.1 defined exclusion
area boundry (EAB) and low population zone (LPZ).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed amendment has no effect on
structures, systems or components. More extensive and recent
meteorological data have been utilized for atmospheric dispersion
factor (X/Q) determination for both EAB and LPZ. An evaluation of
the design basis accidents with revised EAB and LPZ X/Q values
results in increases in UFSAR Chapter 14 EAB and LPZ dose
consequences which remain well within the guidelines of 10 CFR Part
100.
Therefore, this activity does not involve a significant increase
in the probability of occurrence or the consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. The proposed
amendment has no impact on any plant structures, systems or
components. The proposed change revises the atmospheric dispersion
factors for EAB and LPZ used in the existing UFSAR Chapter 14
accident analyses, based on more extensive meteorological data.
These changes only effect the postulated dose consequences of
currently analyzed accidents. Therefore, this activity does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed amendment has no impact on structures, systems
or components. The proposed revisions to the EAB and LPZ X/Q values
are based on recent more extensive meteorological data and
Regulatory Guide 1. 145 methods. The increased X/Q values provide a
more accurate assessment of meteorological conditions which result
in postulated dose consequences which remain well within the
guidelines of 10 CFR Part 100. Therefore, this activity 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.

[[Page 64118]]

Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pitman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: October 19, 1998.
Description of amendment request: The proposed Technical
Specification change request would add operability and surveillance
requirements for the remote shutdown system similar to those in NUREG-
1430, ``Standard Technical Specifications--Babcock and Wilcox Plants''
Section 3.3.18 entitled ``Remote Shutdown System''.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed amendment adds operability and
surveillance requirements for the existing TMI-1 remote shutdown
system similar to those contained in NRC NUREG-1430, ``Standard
Technical Specifications--Babcock & Wilcox Plants''. The addition of
these requirements to Technical Specifications provides further
assurance of remote shutdown system operability in the event that
operators must place and maintain the unit in a safe shutdown
condition from outside the control room. The function and operation
of the remote shutdown system has not changed. Therefore, this
activity has no affect on the probability of occurrence or
consequences of an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. The proposed
amendment has no impact on any plant structures, systems or
components. The function and operation of the remote shutdown system
has not changed. Therefore, this activity does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed amendment provides additional assurance of
remote shutdown system operability. The function and operation of
the remote shutdown system has not changed. Therefore, this activity
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: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: October 19, 1998.
Description of amendment request: The proposed change to the TMI-1
Technical Specification would revise the limit on reactor coolant
system activity to a maximum allowable of 1.0 microcurie/gram dose
equivalent I-131. The proposed revision provides an allowable reactor
coolant system specific activity limit base on once-through steam
generator (OTSG) inspection results performed each refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated. The proposed amendment has no effect on
structures, systems or components. The existing steam line break
criteria are maintained. This change only accounts for radiological
consequences resulting from a revised maximum allowable reactor
coolant system (RCS) specific activity limit of 1.0 iCi/gm.
The new radiological consequences of the revised MSLB accident,
which also incorporate more conservative values for atmospheric
dispersion, are below 10 CFR 100 limits and 10 CFR 50, Appendix A,
GDC-19 limits for the control room. The use of revised atmospheric
dispersion factors for other TMI-1 accident analysis is addressed in
a separate license amendment request submittal. Therefore, this
activity does not involve a significant increase in the probability
of occurrence or the consequences of an accident previously
evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated. The proposed
amendment has no impact on any plant structures, systems or
components. OTSG tube structural integrity is maintained. Therefore,
this activity does not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed amendment has no impact on structures, systems
or components. OTSG tube structural integrity is maintained. The
existing TMI-1 Technical Specification Section 3.1.4.1 Bases state
that the limitations on the specific activity of the primary coolant
ensure that the resulting 2-hour doses at the site boundary will be
well within the Part 100 limit following associated design basis
accidents postulated in conjunction with an assumed steady state
primary-to-secondary steam generator tube leakage of 1.0 gpm. This
margin of safety is preserved since resulting does consequences
incorporating more conservative values for atmospheric dispersion
remain well within the Part 100 limit. Therefore, this activity 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: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, Pitman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit
1, DeWitt County, Illinois

Date of amendment request: October 23, 1998.
Description of amendment request: The proposed amendment would
allow implementation of a feedwater leakage control system to address
leakage through the primary containment feedwater penetration isolation
valves.

[[Page 64119]]

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 implements a method of providing a
qualified sealing system for the primary containment feedwater
penetration isolation valves. This water sealing function, i.e., the
FWLCS, constitutes a new operating mode of the Residual Heat Removal
(RHR) system. The FWLCS introduces new piping that constitutes an
extension of the reactor coolant system (RCS); however, such piping
is designed to the same requirements as other RCS piping and as such
introduces no significant increase in the probability of any
accident previously evaluated. Notwithstanding, a postulated line
break in any of the new FWLCS piping would not, by itself, introduce
any new effects or consequences not already bounded by postulated
line-break or LOCA events previously evaluated in the USAR. Since
the proposed change does not affect any parameters or conditions
that contribute to the initiation of any accidents previously
evaluated, the proposed change cannot increase the probability of
any accident previously evaluated.
The proposed change potentially affects the leak-tight integrity
of the primary containment designed to mitigate the consequences of
a loss-of-coolant accident (LOCA). Once the FWLCS mode has been
initiated and a water seal for the seating surfaces of the primary
containment feedwater penetration isolation valves has been
established (within one hour after the accident), post-LOCA primary
containment atmosphere will be prohibited from leaking through the
feedwater penetrations and thus bypassing the secondary containment.
Calculations of post-accident DBA LOCA doses affected by this
change use accepted ICRP 30 dose conversion factors and take credit
for suppression pool scrubbing. Suppression pool scrubbing is
effective in reducing iodine release but has no assumed effect on
the removal of noble gases. Since the methodology and assumptions
for scrubbing are acceptable to the NRC per the guidance in SRP
Section 6.5.5 and the values for decontamination factors are
conservative, considerable margin is preserved within the analysis.
However, these calculations show increases in some of the previously
evaluated post-accident doses when compared with dose calculations
performed as part of the initial plant licensing basis. Although
some of the newly calculated post-accident doses are larger than
those that were previously approved, the increases remain small
enough to be within the acceptance limits given in 10 CFR 50,
Appendix A, GDC 19 and in 10 CFR 100.11.
Since all of the newly calculated post-accident doses resulting
from the proposed addition of a water sealing system for the
feedwater primary containment penetration isolation valves are below
the 10 CFR 50, Appendix A, GDC 19 and 10 CFR 100.11 acceptance
limits, IP has concluded that the proposed change does not result in
a significant increase in the consequences of an accident previously
evaluated.
2. The proposed change institutes a new operating mode of the
RHR system (the FWLCS mode). When this mode is established, it will
reduce primary containment atmosphere leakage to the environment in
the event of a LOCA. Flow diverted from the RHR system to the FWLCS
has been evaluated, and has been determined to have no adverse
impact on the capability of the RHR system to perform its intended
safety functions. Further, the additional piping added for the FWLCS
is designed to appropriate requirements for the RCS, thus ensuring
that RCS integrity is maintained per design. Sufficient isolation
between the RCS and the RHR low-pressure piping will also be
maintained per the FWLCS design. Thus, no safety functions are
altered or impacted as a result of this change. Installing,
operating, or testing the components that support the FWLCS mode has
no influence on, nor does it contribute to the possibility of a new
or different kind of accident or malfunction from those previously
analyzed. Because the USAR analysis already assumes leakage through
the feedwater primary containment penetrations following a design
basis LOCA, and the subject change does not affect the type of
accident(s) that are postulated to occur, the proposed change does
not present the possibility of an accident of a different type.
Additionally, the change in dose analysis methodology does not
create an accident or malfunction of a different type since it only
involves the analysis of the effects of accidents or malfunctions
previously evaluated in the USAR.
Based on the above, IP has concluded that the proposed change
will not create the possibility of a new or different kind of
accident not previously evaluated.
3. The margin of safety impacted by the proposed change involves
the dose consequences of postulated accidents which are directly
related to the primary containment leakage rate, specifically those
consequences associated with dose attributable to leakage through
the feedwater lines which are secondary containment bypass leakage
paths.
Although considerable conservatisms were included in the
reanalysis, this reanalysis identified some dose values that
increased above the previously licensed values as well as some dose
values that decreased below the previously licensed values. However,
all of the radiation dose consequences resulting from the proposed
change will continue to be below the 10 CFR 50, Appendix A, GDC 19
and 10 CFR 100.11 acceptance criteria.
Except for providing a method of sealing the feedwater primary
containment penetration isolation valves (and therefore the method
of performing periodic leakage testing of these components) no other
change in the method of primary containment leakage testing or
secondary containment bypass leakage path testing is being proposed.
All other primary and secondary containment bypass leakage testing
will continue to be performed in accordance with existing Technical
Specification requirements. Adequate programs are in place to ensure
that proper maintenance and repairs are performed during the service
life of the primary containment, systems and components penetrating
the primary containment, and for all secondary containment bypass
leakage paths.
As a result, IP has concluded that the proposed change will not
result in a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, IL 61727.
Attorney for licensee: Leah Manning Stetzner, Vice President,
General Counsel, and Corporate Secretary, 500 South 27th Street,
Decatur, IL 62525.
NRC Project Director: Stuart A. Richards.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska

Date of amendment request: September 28, 1998.
Description of amendment request: The proposed amendment request
would resolve an unreviewed safety question (USQ) and amend the
operating license to allow manual override capability for the
containment isolation actuation signal to reactor coolant system
letdown isolation valves.
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 modification does not change the probability of any
accident previously evaluated since it does not change any mode of
normal operation. Neither the accident signal (CIAS) nor the
override feature is an initiator of an analyzed event. The
consequences of an accident are also not changed significantly due
to the fact that design and administrative controls ensure that
previous accident analyses are bounding. The associated isolation
valves will operate as they have in the past in response to an
accident signal. There is no single failure that would prevent the
letdown isolation function from occurring. The CIAS override feature
can only be used if operators have verified that an UHE is the event
which has taken place and safety functions are being met.

[[Page 64120]]

This ensures that no significant fuel failures will occur due to the
event and the consequences of overriding CIAS will not adversely
impact radiological conditions in the auxiliary building.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed modification does not create any failure mode which
could impact the operation of the RCS or associated systems in a
manner that would create a new or different kind of accident. With
respect to the letdown isolation function, the plant will operate as
it previously has and will respond the same way, automatically, to
an accident signal. No new accidents have been identified.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The procedural restrictions associated with the use of the CIAS
override feature will ensure that existing analyses addressing the
consequences of an UHE will be bounding and that safety functions
will be maintained as defined in EOPs. The radiological consequences
of letdown restoration in the auxiliary building will be similar to
normal operating conditions and will be bounded by that assumed in
the EEQ analysis. RCS inventory and pressure control will be
maintained within the established procedural limits.
Letdown restoration capability already exists after ESF reset.
The modification permits letdown restoration to occur earlier than
it would previously have been possible.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, NW, Washington, DC 20005-3502.
NRC Project Director: William H. Bateman.

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

Date of amendment request: October 15, 1998.
Description of amendment request: The proposed Technical
Specification (TS) changes involve revising TS Section 3/4.10 to
include a new Special Test Exception allowing the reactor to be
considered in operational condition (OPCON) 4 (cold shutdown) during
inservice leak or hydrostatic testing with a reactor coolant water
temperature greater than 200 deg.F and less than or equal to 212 deg.F.
This is an exception to certain OPCON 3 (hot shutdown) requirements,
including primary containment. The proposed TS changes will permit
unrestricted access to the primary containment for the performance of
required inspections.
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 Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS changes do not make any physical alterations or
modifications to plant systems or equipment. The proposed TS changes
will permit the performance of inservice leak or hydrostatic
testing, with the reactor in OPERATIONAL CONDITION (OPCON) 4 (COLD
SHUTDOWN) and the average reactor coolant temperature greater than
200 deg.F and less than or equal to 212 deg.F. The probability of a
leak in the reactor coolant pressure boundary during inservice leak
or hydrostatic testing is not increased by considering the reactor
in OPCON 4 with reactor coolant temperatures greater than 200 deg.F
and less than or equal to 212 deg.F. The inservice leak and
hydrostatic testing is performed water solid or near water solid.
The stored energy in the reactor core will be very low and the
potential for failed fuel and a subsequent increase in reactor
coolant activity above TS limits is minimal. In addition, Secondary
Containment will be operable and capable of handling airborne
radioactivity from leaks that could occur during the performance of
inservice leak or hydrostatic testing. Requiring the Secondary
Containment to be operable will ensure that potential airborne
radioactivity from leaks will be filtered through the Standby Gas
Treatment System (SGTS), thereby limiting any radioactivity releases
to the environment.
In the event of a large primary system leak, the reactor vessel
would rapidly depressurize allowing the low pressure Emergency Core
Cooling System (ECCS) subsystems to operate. The capability of the
systems that are required for OPCON 4 would be adequate to keep the
core flooded under this condition. Small system leaks would be
detected by leakage inspections before significant inventory loss
has occurred. This is an integral part of the hydrostatic testing
program.
Therefore, the proposed TS changes will not significantly
increase the probability or consequences of an accident previously
evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes do not make any physical alterations or
modifications to plant systems or equipment. The proposed TS changes
do not adversely impact the operation of any plant equipment.
Allowing the reactor to be considered in OPCON 4 during hydrostatic
or inservice leak testing, with a reactor coolant temperature
greater than 200 deg.F and less than or equal to 212 deg.F, is an
exception to certain OPCON 3 (HOT SHUTDOWN) requirements, including
primary containment integrity. The hydrostatic or inservice testing
is performed water solid, or near water solid. The stored energy in
the reactor core will be very low and the potential for failed fuel
and a subsequent increase in coolant activity above TS limits is
minimal. In addition, the Secondary Containment will be operable and
capable of handling airborne radioactivity from leaks that could
occur during the performance of hydrostatic or inservice leakage
testing.
The inservice leak or hydrostatic test conditions remain
unchanged. The potential for a system leak remains unchanged since
the reactor coolant system is designed for temperatures exceeding
500 deg.F with similar pressures. There are no alterations of any
plant systems or components that cope with the spectrum of
accidents.
Therefore, the proposed TS changes will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The proposed TS changes do not make any physical alterations or
modifications to plant systems or equipment. The proposed changes
will permit the performance of inservice leak and hydrostatic
testing with a reactor coolant temperature greater than 200 deg.F
and less than or equal to 212 deg.F and the reactor in OPCON 4.
Since the reactor vessel head will be in place, Secondary
Containment integrity will be maintained, and all systems required
in OPCON 4 will be operable in accordance with the applicable TS
requirements. The proposed TS changes will not have any significant
impact on any design basis accident or safety limit. The hydrostatic
or inservice leak testing is performed water solid, or near water
solid. The stored energy in the reactor core is very low and the
potential for failed fuel and a subsequent increase in coolant
activity would be minimal. In the event of a large primary system
leak, the reactor pressure vessel would rapidly depressurize and the
low pressure ECCS subsystems would function as designed to maintain
adequate reactor core coverage. This would ensure that the fuel
would not exceed peak clad temperature limits.
Also, requiring Secondary Containment integrity will assure that
potential airborne radioactive material can be filtered through the
SGTS. This will assure that any offsite doses remain well within the
limits of 10 CFR 100 guidelines. Small system leaks would be
detected by inspections before significant inventory loss could
occur.
Therefore, this proposed TS change will not involve a
significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 64121]]

satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101.
NRC Project Director: Robert A. Capra.

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

Date of amendment request: October 19, 1998.
Description of amendment request: The proposed amendment would
eliminate restrictions imposed by Technical Specification (TS) 3.0.4
for the Filtration, Recirculation and Ventilation System (FRVS) during
fuel movement and core alteration activities. Specifically, TS Limiting
Conditions for Operation (LCOs) 3.6.5.3.1 and 3.6.5.3.2 would each be
revised to add a note stating that the provisions of TS 3.0.4 are not
applicable for initiation of handling of irradiated fuel in the
secondary containment and core alterations provided that the plant is
in Operational Condition 5, with reactor water level equal to or
greater than 22 feet 2 inches.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS change does not involve any physical changes to
plant structures, systems or components (SSC). FRVS will continue to
function as designed. FRVS is an Engineered Safety Feature (ESF)
designed to mitigate th

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3A98-30691. Public record. Not legal advice.