# Biweekly Notice

> Briefs, arguments, decisions, and more.

URL: https://www.frixlaw.com/law-library/documents/fr%3AX97-10423

## Record

- **Collection:** Federal Register
- **Document type:** Notice
- **Published:** April 23, 1997
- **Citation:** 62 FR 19825

## Text

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 29, 1997, through April 11, 1997. The
last biweekly notice was published on April 9, 1997 (62 FR 17223).

Notice of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunith 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
Review and Directives Branch, Division of Freedom of Information and
Publications 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 May 23, 1997, 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

[[Page 19826]]

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: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina

Date of amendment request: March 17, 1997
Description of amendment request: The proposed change would revise
eight specifications for 18-month tests to delete a conditional
statement that the testing be done while the unit is shut down and to
clarify that Harris Nuclear Plant (HNP) may take credit for tests on
some components which are performed while the unit is at 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. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes permit HNP to evaluate the conditions
required to safely perform a test, but the changes do not directly
affect the functioning or operation of any plant equipment. Since no
equipment operation is involved there is no increase in the
probability or consequence of any previously identified accident.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes to the conditional statements on the
surveillance frequencies do not involve any physical alterations or
additions to plant equipment or alter the manner in which any
safety-related system performs itsfunction or is operated.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed changes to the conditional statements on the
surveillance frequency allows HNP to evaluate the conditions needed
to safely perform the required testing. There is no change in the
frequency of testing or in the testing which is required. There is
no change in the responsibility of HNP toperform tests in a safe and
responsible manner, and any changes to procedures will have to be
individually evaluated to ensure that they do 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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Mark Reinhart, Acting

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois

Date of amendment request: January 30, 1997
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 1.0, ``Definitions;'' TS 3/4.6.1,
``Primary Containment'' and associated Bases; and TS 5.4.2, ``Reactor
Coolant System Volume'' for Byron and Braidwood to support steam
generator replacement. ComEd will be replacing the original
Westinghouse D4 steam generators at Byron and Braidwood with Babcock
and Wilcox International steam generators. The replacement steam
generators increase the Reactor Coolant System volume which results in
a higher calculated peak containment pressure (Pa) value.
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.
Each of the [replacement steam generators] RSGs has a larger
[reactor coolant system] RCS side volume than the original steam
generators (OSGs). As a result of the RCS

[[Page 19827]]

volume increase, the mass and energy release during the blowdown
phase of the large break loss of coolant accident (LBLOCA) is
increased. Additionally, the heat transfer rate of the RSGs is
greater than the OSGs, and the RSGs will operate at a slightly
higher pressure than that for the OSGs. Consequently, the steam
enthalphy exiting the break during the reflood period, with the RSG,
will be greater than that for the OSG. This results in an increase
in the containment building peak pressure, Pa.
The proposed revisions to the Technical Specifications involve
the specified value of Unit 1 RCS volume and the defined value of
Unit 1 Pa. Several editorial changes are also being made
to improve clarity and consistency of the TS.
RCS volume is not an initiator for any event and an increase in
volume does not affect any operating margin or requirements.
Therefore, increasing the primary volume does not increase the
probability of any event previously analyzed.
The revised value of Pa continues to be less than the
design basis pressure for the containment building structure. The
change represents only a revision to the containment test pressure
for containment leakage testing. Such testing is only performed with
the affected unit in the shutdown condition. Therefore, the proposed
change in Pa does not involve a significant increase in
the probability of an accident previously evaluated.
All accidents in the Updated Final Safety Analysis Report
(UFSAR) were evaluated to determine the effect of an increase in
primary volume on accident consequences. The events identified that
may be impacted by an increase in primary volume are the Waste Gas
System Leak or Failure and LBLOCA. For the Waste Gas System Leak or
Failure, the activity of the decay tank is controlled to Technical
Specification limits which are unaffected by RCS volume. Therefore,
an increase in RCS volume would not increase the offsite dose.
The offsite dose calculation for the LBLOCA is unaffected by the
proposed change. The license basis offsite dose calculation is in
accordance with NRC Reg Guide 1.4 ``Assumptions Used for Evaluating
The Potential Radiological Consequences of a Loss of Coolant
Accident for Pressurized Water Reactors.'' This Regulatory Guide
states, in part, ''...a number of appropriately conservation
assumptions, based on engineering judgment and on applicable
experimental results from safety research programs conducted by the
AEC.'' These conservatisms include (but are not limited to) the
following assumptions:
Twenty five percent of the equilibrium radioactive full
power inventory is immediately available for leakage from the
primary containment.
100% of the equilibrium full power radioactive noble
gas inventory is immediately available for leakage from the primary
containment.
The primary containment should be assumed to leak at
the (maximum) leak rate specified in the technical specifications
for the first 24 hours and at 50% of this value for the remaining 29
days of the accident duration.
The design basis leakage corresponding to a peak containment
pressure of 50 psig utilized in the design basis accident analysis
is 0.10% per day of the containment free air mass. Therefore, the
offsite dose calculation was performed with a leakage of .1% per day
for day one and .05% per day for days two through 30. Isotopic
inventories are unaffected by the increase in reactor coolant
volume. Thus, the offsite dose is unaffected by the increase in the
peak containment pressure. Therefore, this proposed change to
Pa does not involve a significant increase in the
consequences of an accident previously evaluated.
The editorial changes proposed are for clarity and consistency
within the Technical Specifications and do not affect either the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change in RCS volume is a change in a plant
parameter within the ``Design Features'' section of the Technical
Specifications. Increasing the RCS volume does not create any new or
different failure modes. The existing RCS design requirements
continue to be met.
The revised value of Pa continues to be less than the
design basis pressure for the containment building structure. The
change represents only a revision to the test pressure for
containment leakage testing. Such testing is only performed with the
affected unit in the shutdown condition. Therefore, no new or
different failure modes are being introduced by modification of the
testing parameters.
The editorial changes proposed are for clarity and consistency
within the Technical Specifications and do not result in any
physical changes to the facility or how it is operated. No new or
different failure modes are being introduced by these changes.
Therefore, these proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Changing the RCS volume in the Technical Specifications does not
reduce the margin of safety. RCS volume is a design feature. The
change in RCS volume does not involve a change to any setpoint or
design requirements. An evaluation of all UFSAR accidents was
performed to determine the effect of an increase in RCS volume. This
evaluation is summarized as follows:
An evaluation of the Chemical and Volume Control System
Malfunction was performed to determine the effect of the increased
RCS volume due to the RSGs. The larger RCS volume of the RSGs
reduces the reactivity insertion for a given dilution flow rate.
Therefore, the UFSAR analyses remain bounding for Byron Unit 1 and
Braidwood Unit 1 with the RSGs and there is no reduction in the
margin of safety.
An evaluation of the Inadvertent Actuation of the Emergency Core
Cooling System During Power Operation Event was performed to
determine the effect of the increased RCS volume due to the RSGs.
For this event, the injection of borated water causes a negative
reactivity insertion, which increases DNBR. For a given Refueling
Water Storage Tank (RWST) boron concentration, the larger RCS volume
will cause a reduction in the negativity insertion rate as compared
to the current UFSAR analysis. However, negative reactivity would
still be inserted, no fuel pins would experience DNB, and there is
no reduction in the margin of safety.
An evaluation of the Small Break LOCA was performed to determine
the effect of increased RCS volume. The additional RCS volume will
cause a delay in the loop seal clearing which in turn delays the
core uncovery as compared with the UFSAR analysis. A delay in core
uncovery reduces the amount of core heatup which results in a lower
peak clad temperature (PCT) because the core decay heat would be
less than in the UFSAR analysis. The benefit is considered small,
but there is still a benefit. Therefore, the increased RCS volume
does not result in a reduction in the margin of safety.
An evaluation of the Large Break LOCA was performed to determine
the effect of increased RCS volume. For a LBLOCA, the increased RCS
volume causes the blowdown phase of the event to be longer.
Increased blowdown phase, alone, could potentially result in a
higher PCT. However, the RSGs also have less resistance to flow due
to increased primary side steam generator flow area, which results
in a higher blowdown flow compared to the OSGs. The increased
blowdown flow more than compensates for the longer blowdown phase
associated with the increased RCS volume. The net effect is a
decrease in PCT for the RSG compared to the OSG. Therefore, there is
no reduction in the margin of safety.
An evaluation of the Gas Waste System Leak or Failure was
performed to determine the effect of the increased RCS volume.
Because the activity of the decay tank is controlled within
Technical Specification limits, an increase in RCS volume would not
change the results of the event. Therefore, there is no reduction in
the margin of safety.
An evaluation was performed to determine the effect of the
increased RCS volume on the peak containment pressure following a
LBLOCA. The increased RCS volume caused the peak containment
pressure to increase to 47.8 psig. This is still below the
containment design pressure of 50.0 psig. Therefore, there is no
reduction in the margin of safety.
This proposed change involves testing requirements designed to
demonstrate adequate leakage rates are maintained. If adequate
leakage rates are maintained as outlined in the Technical
Specifications, there will be no reduction in the margin of safety.
In the event of degradation of a containment seal that results in
unacceptable leakage, plant shutdown will occur as required by
Technical Specifications and administrative requirements in
accordance with approved plant procedures. Therefore, this proposed
change does not involve a significant reduction in a margin of
safety.
The editorial changes proposed are for clarity and consistency
within the Technical Specifications and do not result in any
physical changes to the facility or how it is

[[Page 19828]]

operated. Therefore, the changes have no effect on the margin of
safety.
Thus, this amendment request does not result in any decrease in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: For Byron, the Byron Public
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,
Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603
NRC Project Director: Robert A. Capra

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan

Date of amendment request: March 27, 1997
Description of amendment request: The proposed amendment would
alter the company name in the Facility Operating License DPR-20 and
Technical Specifications for the Palisades Plant. Specifically, the
proposed amendment would revise the name from ``Consumers Power
Company'' to ``Consumers Energy Company.''
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Since the proposed changes do not alter the technical content of
any Facility Operating License or Technical Specifications
requirements, they do not alter any feature of plant equipment,
settings, operation, or configuration.
Therefore, they cannot involve a significant increase in the
probability of an accident previously evaluated.
The proposed changes alter the company name in the Facility
Operating License and Technical Specifications to reflect the change
from ``Consumers Power Company'' to ``Consumers Energy Company''.
The proposed change will not affect any obligations. The company
will continue to own all of the same assets, will continue to serve
the same customers, and will continue to honor all existing
obligations and commitments. The proposed changes will not alter
plant operation or configuration, or its ability to respond to
accidents.
Therefore, they will not involve a significant increase in the
consequences of any accident previously evaluated.
B. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
Since the proposed changes do not alter the technical content of
any Facility Operating License or Technical Specifications
requirements, they do not alter any feature of plant equipment,
settings, operation or configuration.
Therefore, they cannot create the possibility of a new or
different kind of accident from any previously evaluated.
C. Do the proposed changes involve a significant reduction in a
margin of safety?
Since the proposed changes do not alter the technical content of
any Facility Operating License or Technical Specifications
requirements, they do not alter any feature of plant equipment,
settings, operation, or configuration.
Therefore, they cannot 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: Van Wylen Library, Hope
College, Holland, Michigan 49423
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: John N. Hannon

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: March 31, 1997 (TSC 96-10)
Description of amendment request: The proposed amendments would
modify and clarify the High Pressure Injection (HPI) System operability
requirements in Specification 3.3.1, impose additional HPI system
operability requirements for operation above 75 percent power,
incorporate the new Standard Technical Specifications format for the
HPI system, revise Specification 3.3.2 to clarify that the Reactor
Building Emergency Sump isolation valves are remote-manually operated
valves, and add new specifications and a surveillance test to address
operability requirements of the atmospheric dump valves. In addition,
corresponding Bases changes would be incorporated.
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. None of the proposed changes has any impact upon the
probability of any accident which has been evaluated in the UFSAR
[Updated Final Safety Analysis Report]. The only potential change in
operating configuration is allowing operation with the HPI [High
Pressure Injection] System pump discharge header cross-
connected. This operating mode does not affect the probability
of a LOCA [Loss-of-Coolant Accident] or of any other accident
evaluated in the UFSAR.
None of these changes have any impact upon the ability of the
HPI System to add soluble poison to the Reactor Coolant System, as
addressed by Specification 3.2. The remaining potential impact is
upon the ability to mitigate the consequences of a small break LOCA,
which is addressed below. The small break LOCA is the limiting
design basis accident with respect to HPI System operability
requirements.
The proposed changes to Specification 3.3.1 provide appropriate
actions to address any degradation in the operability of the HPI
System. The operability requirements for the HPI System are
supported by a spectrum of small break LOCA analyses based on the
approved Evaluation Model described in FTI [Framatome Technologies,
Incorporated] topical report BAW-10192P. These small break LOCA
analyses demonstrate that the acceptance criteria of 10CFR 50.46 are
not violated.
Two trains of HPI are required to mitigate a small break LOCA
above 75% FP [full power]. Operability requirements in the proposed
Technical Specifications assure that the HPI System can withstand
the worst single failure and still result in two HPI pumps injecting
through two trains. The full power small break LOCA analyses
supporting this proposed license amendment have been performed in
accordance with the approved Evaluation Model described in FTI
topical report BAW-10192P. The proposed Technical Specifications
limit operation above 75% FP with a degraded HPI System to 72 hours
before a power reduction to less than 75% FP (or a reactor shutdown)
must be initiated. The required actions depend on the HPI System
components that are inoperable. The 72 hour completion time is
consistent with the time requirements for HPI specified in NUREG-
1430.
When at or below 75% FP, one HPI train provides sufficient flow
to mitigate a small break LOCA. The 75% power level is justified by
analyses using the Evaluation Model described in FTI topical report
BAW-10192P, considering the worst case break location and size
described in LER [Licensee Event Report] 269/90-15 and Attachment 3
to this submittal. The proposed Technical Specifications require two
HPI trains to be operable at or below 75% FP. These requirements
ensure that, following the worst single failure, one train of HPI
would remain

[[Page 19829]]

available to mitigate a small break LOCA. Operation with less than
two HPI trains operable is restricted to 72 hours before shutdown
requirements are imposed. This completion time is consistent with
the time requirements specified for an HPI System in NUREG-1430.
The additional HPI system restriction that requires the HPI pump
discharge header to be cross-connected when all three HPI pumps are
operable does not increase the consequences of a small break LOCA.
If a single failure prevents one HPI train from actuating, this
lineup results in at least two HPI pumps initially injecting through
the automatically actuating train. This increases the amount of
cooling flow initially delivered to the core as compared to the
current system configuration.
The impact of this alignment has been evaluated, considering the
potential single active failures, including the failure of any
powered component to operate and any single failure of electrical
equipment.
It has been determined that, when each of the three HPI pumps is
either running or is capable of automatic actuation upon an
Engineered Safeguards signal, cross-connection of the HPI pump
discharge header does not introduce susceptibility to any single
failure. Therefore, the potential consequences of a small break LOCA
are not increased. If fewer than three HPI pumps are either running
or are capable of automatic actuation, and the HPI pump discharge
header were cross-connected, a single failure of one pump could
cause a single pump to be aligned to both HPI trains. In this
condition, the single pump could experience runout conditions prior
to corrective operator action. However, proposed Specification 3.3.1
requires the discharge header to be isolated between the two
remaining operable HPI pumps. The proposed BASES provide guidelines
to ensure that the requirements for redundancy are properly
implemented. Therefore, the proposed specifications ensure that the
consequences of a small break LOCA are not increased by allowing the
HPI pump discharge header to be cross-connected.
In addition, proposed Specification 3.4.7 requires new
operability requirements for the main steam atmospheric dump valves.
These operability requirements do not impact the probability or
consequences of any accident. The proposed specification for the
atmospheric dump valves provides additional assurance that these
valves will be operable in the event of a small break LOCA.
In summary, the proposed Technical Specifications provide
adequate controls to assure that operability of the HPI System is
maintained in a manner consistent with the requirements of the
design basis accidents. Therefore, it is concluded that this
amendment request will not significantly increase the probability or
consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
No. Of the proposed substantive changes, only cross-connection
of the HPI pump discharge header represents any change to the way in
which the facility is normally operated. Operation with the
discharge header cross-connected is not a new configuration, as it
has always been used for HPI pump testing both at power and during
shutdown conditions. Potential failure modes have already been
considered as described earlier. No new initiating events or
potentially unanalyzed conditions have been created. Therefore, this
proposed amendment will not create the possibility of any new or
different kind of accident.
(3) Involve a significant reduction in a margin of safety.
No. The HPI restrictions associated with the proposed Technical
Specifications are supported by analyses which demonstrate that the
acceptance criteria of 10 CFR 50.46 are not violated for any small
break LOCA. These analyses were performed in accordance with the
Evaluation Model described in FTI topical report BAW-10192P.
Therefore, it is concluded that the proposed amendment request will
not result in a significant decrease in the margin of safety.
Duke has concluded, based on the above, that there are no
significant hazards considerations involved in this amendment
request.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
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: March 10, 1997
Description of amendment request: The proposed amendment would
modify Technical Specification 3.4.5, ``Steam Generators,'' and
associated Bases to allow repair of steam generator tubes by
installation of sleeves with the tungsten inert gas (TIG) welded sleeve
developed by ABB Combustion Engineering. In addition, the proposed
amendment would delete the option for using the kinetic sleeving
methodology previously approved for use at Beaver Valley, but is not
currently recommended by Framatome Technologies, Inc.
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed amendment allows the ABB Combustion Engineering
(ABB/CE) tungsten inert gas (TIG) welded tubesheet sleeves and tube
support plate sleeves to be used as an alternate steam generator
tube repair method. The sleeve configuration was designed and
analyzed in accordance with the criteria of Regulatory Guide (RG)
1.121 and Section III of the ASME [American Society of Mechanical
Engineers] Code. Fatigue and stress analyses of the sleeved tube
assemblies produce acceptable results for both types of sleeves as
documented in ABB/CE Topical Report CEN-629-P, Revision 02 and CEN-
629-P Addendum 1. Mechanical testing has shown that the structural
strength of the sleeves under normal, faulted, and upset conditions
is within the acceptable limits specified in RG 1.121. Leakage rate
testing for the tube sleeves has demonstrated that primary to
secondary leakage is not expected during any plant condition. The
consequences of leakage through the sleeved region of the tube is
fully bounded by the existing steam generator tube rupture (SGTR)
analysis included in the Updated Final Safety Analysis Report
(UFSAR).
The sleeves are designed to allow inservice inspection of the
pressure retaining portions of the sleeve and parent tube. Inservice
inspection is performed on all sleeves following installation to
ensure that each sleeve has been properly installed and is
structurally sound. Periodic inspections are performed in subsequent
refueling outages to monitor sleeve degradation on a sample basis.
The eddy current technique used for inspection will be capable of
detecting both axial and circumferential flaws. Specific guidance
for steam generator sleeve inspection is provided in the current
technical specification surveillance requirements. Tubes that
contain defects in a sleeve, which exceed the repair limit, will be
removed from service. This ensures that sleeve and tube structural
integrity is maintained.
The proposed TS change to support the installation of TIG welded
sleeves does not adversely impact any previously evaluated design
basis accident. The effect of sleeve installation on the performance
of the SG [steam generator] was analyzed for heat transfer, flow
restriction, and steam generation capacity. The sleeves reduce the
risk of primary to secondary leakage in the SG. The installation of
ABB/CE sleeves results in a hydraulic flow restriction that is
dependent on the number and types of sleeves installed. The
reduction in primary system flow rate is a small percentage of the
flow rate reduction seen from plugging one tube and is a preferable
alternative when considering core margins based on minimum reactor
coolant system flow rates. The sleeving installation will result in
a resistance to primary coolant flow through the tube for other
evaluated accidents. The results of the analyses and testing, as
well as industry operating experience, demonstrate that the sleeve
assembly is an acceptable

[[Page 19830]]

means of maintaining tube integrity. In summary, installation of
sleeves does not substantially affect the primary system flow rate
or the heat transfer capability of the steam generators.
Installation of the sleeves can be used to repair degraded tubes
by returning the condition of the tubes to their original design
basis condition for tube integrity and leak tightness during all
plant conditions. The tube bundle overall structural and leakage
integrity will be increased with the installation of the sleeves
reducing the risk of primary to secondary leakage in the SG while
maintaining acceptable reactor coolant system flow rates. Therefore,
sleeving will not increase the probability of occurrence of an
accident previously evaluated.
Removal of the kinetically welded sleeve process as an approved
SG tube repair methodology will have no effect on plant operations.
There are currently no kinetically welded sleeves installed in the
steam generators. Had there been, plant operations would have still
been bounded by the existing SGTR analysis in the UFSAR.
Therefore, these proposed changes do 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 implementation of the proposed sleeving process will not
introduce significant or adverse changes to the plant design basis.
Stress and fatigue analyses of the repair has shown the ASME Code
Section III and RG 1.121 allowable values are met. Implementation of
TIG welded sleeving maintains overall tube bundle structural and
leakage integrity at a level consistent with that of the originally
supplied tubing. Leak and mechanical testing of the sleeves support
the conclusions that the sleeve retains both structural and leakage
integrity during all conditions. Repair of a tube with a sleeve does
not provide a mechanism that would result in an accident outside of
the area affected by the sleeve.
Any hypothetical accident as a result of potential tube or
sleeve degradation in the repaired portion of the tube is bounded by
the existing SGTR analysis. The SGTR analysis accounts for the
installation of sleeves and the impact on current plugging level
analyses. The sleeve design does not affect any other component or
location of the tube outside of the immediate area repaired.
The current primary to secondary leakage limit ensures that SG
tube integrity is maintained in the event of an MSLB [main steam
line break] or LOCA [loss-of-coolant accident]. The limit will
provide for leakage detection and a plant shutdown in the event of
the occurrence of an unexpected single crack resulting in excessive
tube leakage. The leakage limit also provides for early detection
and a plant shutdown prior to a postulated crack reaching critical
crack lengths for MSLB conditions.
Inservice inspections are performed following sleeve
installation to ensure proper weld fusion has occurred to maintain
structural integrity. The post installation inspection also serves
as baseline data to be used for comparison during future
inspections. Periodic eddy current inspections monitor the pressure
retaining portions of the sleeve and parent tube for degradation.
Eddy current techniques will be employed that are sensitive to axial
and circumferential degradation.
Increasing the sample size of tubes repaired using either
sleeving process during each scheduled inservice inspection will
increase the monitoring of these tubes for any further degradation.
The improved monitoring and evaluation of the tube and the sleeves
assures tube structural integrity is maintained or the tube is
removed from service.
Corrosion testing of typical sleeve-tube configurations was
performed to evaluate local stresses, sleeve life, and resistance to
primary and secondary side corrosion. The tests were performed on
stress relieved and as-welded (non-stress relieved) sleeve-tube
joints. Using the corrosion test data in conjunction with finite
element analyses of the local stress, the stress relieved joint life
was determined to be in excess of 40 years. The ABB/CE TIG welded
sleeve operating experience in the industry has shown no sleeve
failures due to service induced degradation in sleeves that were
installed with acceptable inspection results. This experience
includes the stress relieved and as-welded sleeve configurations.
All sleeves will be stress relieved as specified in the topical
report.
Removal of the kinetically welded sleeve process as an approved
SG tube repair methodology and not completing the additional
corrosion testing necessary to establish the design life for the
kinetically welded sleeve in the presence of a crevice will not
create the possibility of a new or different type of accident from
any accident previously evaluated.
Repair of an SG tube with a kinetically welded sleeve would not
have provided a mechanism that resulted in an accident outside of
the area affected by the sleeve. Any hypothetical accident as a
result of potential tube or sleeve degradation in the repaired
portion of the tube would have been bounded by the existing SGTR
analysis. The SGTR analysis accounts for the installation of sleeves
and the impact on current plugging level analyses. The sleeve design
does not affect any other component or location of the tube outside
of the immediate area repaired. Furthermore, there are currently no
kinetically welded sleeves installed in either plant.
Therefore, the proposed changes do 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 TIG welded sleeving repair of degraded steam generator tubes
has been shown by analysis to restore the integrity of the tube
bundle to its original design basis condition. The safety factors
used in the design of the sleeves for the repair of degraded tubes
are consistent with the safety factors in the ASME Boiler and
Pressure Vessel Code Section III used in steam generator design. The
design of the ABB/CE SG sleeves has been verified by testing to
preclude leakage during normal and postulated accident conditions.
The portion of the installed sleeve assembly which represents
the reactor coolant pressure boundary can be monitored for the
initiation and progression of sleeve/tube wall degradation, thus
satisfying the requirement of RG 1.83. The portion of the SG tube
bridged by the sleeve joints is effectively removed from the
pressure boundary, and the sleeve then forms the new pressure
boundary. The sleeve enhances the safety of the plant by
reestablishing the protective boundaries of the steam generator.
Keeping the tube in service with the use of a sleeve instead of
plugging the tube and removing it from service increases the heat
transfer efficiency of the steam generator. During each scheduled
inservice inspection, each sleeve inspected and found to have
unacceptable degradation shall be removed from service.
The effect on the design transients and the accident analyses
have been revised based on the installation of sleeves equal to the
tube plugging level coincident with the minimum reactor coolant flow
rate. Evaluation of the installation of sleeves was based on the
determination that LOCA evaluations for the licensed minimum reactor
coolant flow bound the combined effect of tube plugging and sleeving
up to an equivalent of the actual plugging limit. Sleeving results
in a fractional amount of the plugging limitation of one tube and is
a preferable alternative when considering core margins based on
minimum reactor coolant system flow rates. The sleeving installation
will result in a resistance to primary coolant flow through the
tube. The primary coolant flow through the ruptured tube is reduced
by the influence of the installed sleeve; therefore, the
consequences to the public due to an SGTR event have not increased.
As SG sleeve removes an indication of a possible leak source
from the reactor coolant system (RCS) pressure boundary, eliminating
the potential of a primary-to-secondary leak. The structural
integrity of the tube is maintained by the sleeve and sleeve-to-tube
joint.
Installation of either tube sheet or tube support plate sleeves
will increase the protective boundaries of the steam generators and
will not reduce the margin of safety.
Removal of the kinetically welded sleeve process as an approved
SG tube repair methodology will not result in a reduction in the
margin of safety. There are currently no kinetically welded sleeves
installed in either plant. SG tube integrity will be maintained by
applying an alternate NRC approved repair methodology or removing
the SG tube from service by plugging.
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
amendment request involves no significant hazards consideration.

[[Page 19831]]

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: John F. Stolz

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: March 10, 1997
Description of amendment request: The proposed amendment would
revise Technical Specifications 3.4.5, ``Steam Generators,'' and
associated Bases to allow repair of steam generator tubes by
installation of sleeves with the Electrosleeving process developed by
Framatome Technologies, Inc. (FTI).
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The Electrosleeve configuration has been designed and analyzed
in accordance with the requirements of the ASME [American Society of
Mechanical Engineers] Code. The applied stresses and fatigue usage
for the Electrosleeve are bounded by the limits established in the
ASME Code. Minimum material property values are used for the
structural and plugging limit analysis. Mechanical testing has shown
that the structural strength of nickel Electrosleeves under normal,
upset, and faulted conditions provides margin to the acceptance
limits. These acceptance limits bound the most limiting (3 times
normal operating pressure differential) burst margin recommended by
Regulatory Guide 1.121. Leakage testing has shown that the
Electrosleeve is essentially leaktight during all plant conditions.
The Electrosleeve nominal wall thickness depth-based plugging
limit is determined using the guidance of Regulatory Guide 1.121 and
the pressure stress equation of Section III of the ASME Code. The
limiting requirement of Regulatory Guide 1.121 for the
Electrosleeve, which applies to part through wall degradation, is
the minimum acceptable wall thickness to maintain a safety factor of
three against tube failure under normal operating conditions. A
bounding set of design and transient loading input conditions was
used for the minimum wall thickness evaluation in the generic
evaluation. Evaluation of the minimum acceptable wall thickness for
normal, upset and postulated accident condition loading per the ASME
Code indicates these conditions are bounded by the design minimum
wall thickness.
Bounding tube wall degradation growth rate per cycle and
nondestructive examination uncertainty has been assumed for
determining the Electrosleeve technical specification plugging
limit. Electrosleeve wall degradation extent determined by
nondestructive examination, which would require plugging
Electrosleeved tubes, is developed using the guidance of Regulatory
Guide 1.121 and is defined in FTI Topical Report BAW-10219P,
Revision 1, to be 20% throughwall of the nominal sleeve wall
thickness.
The effect of Electrosleeving and plugging will remain below the
plugging limit assumed in the UFSAR [Updated Final Safety Analysis
Report]. The proposed change will not increase the consequences of
these accidents.
The results of the analyses and testing demonstrate that the
Electrosleeve is an acceptable means of maintaining tube integrity.
Further, per Regulatory Guide 1.83 recommendations, the
Electrosleeved tube can be monitored through periodic inspections
with present NDE [nondestructive examination] techniques. These
measures demonstrate that installation of Electrosleeves spanning
degraded areas of the tube will restore the tube to a condition
consistent with its original design basis.
Since the main steamline break post-accident primary-to-
secondary leakage is not increased by the presence of
Electrosleeves, the consequences of an accident previously evaluated
in the UFSAR are not increased. Conformance of the Electrosleeve
design with the applicable sections of the ASME Code and results of
the leakage and mechanical tests support the conclusion that
installation of Electrosleeves does not increase 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?
Electrosleeving will not adversely affect any plant component.
Stress and fatigue analysis of the repair has shown that the ASME
Code and Regulatory Guide 1.121 criteria are not exceeded.
Implementation of Electrosleeving maintains overall tube bundle
structural and leakage integrity at a level consistent with that of
the original tubing during all plant conditions. Leak and mechanical
testing of Electrosleeves support the conclusions of the
calculations that each Electrosleeve retains both structural and
leakage integrity during all conditions. Electrosleeving of tubes
does not provide a mechanism resulting in an accident outside of the
area affected by the Electrosleeves. Any accident resulting from
potential tube or Electrosleeve degradation in the repaired portion
of the tube is bounded by the existing tube rupture accident
analysis.
Implementation of Electrosleeving will reduce the potential for
primary-to-secondary leakage while not significantly impacting
available primary coolant flow area in the event of a LOCA. By
effectively isolating degraded areas of the tube through repair, the
potential for steamline break leakage is reduced. These degraded
intersections now are returned to a condition consistent with the
Design Basis. While the installation of an Electrosleeve reduces
primary coolant flow, the reduction is far below that caused by
plugging. Greater primary coolant flow area is maintained through
Electrosleeving versus plugging. Therefore, the possibility of a new
or different kind of accident from any accident previously evaluated
is not created.
3. Does the change involve a significant reduction in a margin
of safety?
The Electrosleeve repair of degraded steam generator tubes has
been shown by analysis to restore the integrity of the tube bundle
consistent with its original design basis condition. The tube/
Electrosleeve operational and faulted condition stresses are bounded
by the ASME Code requirements and the Electrosleeved tubes are
leaktight. The safety factors used in the design of Electrosleeves
for the repair of degraded tubes are consistent with the safety
factors in the ASME Code used in steam generator design. The
portions of the installed Electrosleeve assembly which represent the
reactor coolant pressure boundary can be monitored for the
initiation and progression of Electrosleeve/tube wall degradation,
thus satisfying the requirements of Regulatory Guide 1.83. The
portion of the tube bridged by the Electrosleeve is effectively
removed from the pressure boundary, and the Electrosleeve then forms
the new pressure boundary. The areas of the Electrosleeved tube
assembly which require inspection are defined in Framatome
Technologies Inc. Topical Report BAW-10219P, Revision 1.
In addition, since the installed Electrosleeve represents a
portion of the pressure boundary, a baseline inspection of these
areas is required prior to operation with Electrosleeves installed.
The effect of sleeving on the design transients and accident
analyses has been reviewed based on the installation of
Electrosleeves up to the level of steam generator tube plugging
coincident with the minimum reactor coolant flow rate and UFSAR and
has been found acceptable.
It is concluded that the proposed license amendment request does
not result in a significant reduction in the margin of safety as
defined in the UFSAR or 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 & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz

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

Date of amendment request: March 27, 1997

[[Page 19832]]

Description of amendment request: The proposed changes to the
Technical Specifications (TSs) would modify the limiting condition for
operation (LCO) and surveillance requirements (SR) for the ultimate
heat sink. The ultimate heat sink for Millstone Unit No. 2 is the Long
Island Sound that transfers heat from safety-related systems during
normal and accident conditions. Specifically, TS LCO 3.7.11 would be
changed to indicate that the ultimate heat sink is operable at a water
temperature of less than or equal to 75 deg.F instead of an average
value. TS SRs 4.7.11.a and .b would also delete the use of average when
verifying the water temperature and delete the reference to a specific
monitoring location, the Unit No. 2 intake structure. These proposed
changes do not change the ultimate heat sink temperature limit, which
remains at a maximum of 75 deg.F.
The TS Bases 3/4.7.11 would also be modified to reflect the above
changes and to identify the various locations that the ultimate heat
sink temperature can be measured.
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes remove the reference to a monitoring
location where the temperature of the ultimate heat sink is measured
and eliminate the use of an average ultimate heat sink temperature.
The instruments used provide information to the operators which will
permit them to ensure that the plant is operated within the design
basis of the plant. The subject instruments will provide an accurate
representation of the ultimate heat sink temperature. This role is
passive; thus, these instruments cannot initiate or mitigate any
accident.
The locations used to monitor the ultimate heat sink temperature
will be maintained in the Bases. This is a licensee controlled
document which is maintained under the requirements of 10CFR50.59.
The details being removed from the Technical Specifications are not
assumed to be an initiator of any analyzed event. Since any changes
to the relocated details will be evaluated per 10CFR50.59, any
possible increase in the probability or consequences of an accident
previously evaluated will be addressed.
The proposed changes do not revise the ultimate heat sink
temperature limit of 75 deg.F. The current analysis is based on the
ultimate heat sink temperature limit of 75 deg.F. Therefore, there
is no effect on the consequences of any accident previously
evaluated.
Thus, the license amendment request does not impact the
probability of an accident previously evaluated nor does it involve
a significant increase in the consequences of an accident previously
evaluated.
2. Created the possibility of a new or different kind of
accident from any previously evaluated.
The proposed changes remove the reference to a monitoring
location where the temperature of the ultimate heat sink is measured
and eliminate the use of an average ultimate heat sink temperature.
The instruments used provide information to the operators which will
permit them to ensure that the plant is operated within the design
basis of the plant. The subject instruments will provide an accurate
representation of the ultimate heat sink temperature. This role is
passive, thus, these instruments cannot initiate or mitigate any
accident.
The proposed changes will not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes do not
introduce any new failure modes. They will not alter assumptions
made in the safety analysis and licensing basis.
The locations used to monitor the ultimate heat sink temperature
will be maintained in the Bases. This is a licensee controlled
document which is maintained under the requirements of 10CFR50.59.
Thus, adequate control of information will be ensured.
Therefore, the changes will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes remove the reference to a monitoring
location where the temperature of the ultimate heat sink is measured
and eliminate the use of an average ultimate heat sink temperature.
They do not change the ultimate heat sink temperature limit of 75
deg.F, which is assumed by the current analysis. Therefore, there is
no effect on the consequences of any accident previously evaluated
and no significant impact on offsite doses associated with
previously evaluated accidents. Thus, there is no significant
reduction in the margin of safety for the design basis accident
analysis. The license amendment request does not result in a
reduction of the margin of safety as defined in the Bases for
Technical Specification 3.7.11. The instruments used provide
information to the operators which will permit them to ensure that
the plant is operated within the design basis of the plant. The
subject instruments will provide an accurate representation of the
ultimate heat sink temperature. The proposed changes do not alter
the way any structure, system, or component functions and do not
alter the manner in which the plant is operated. They do not have
any impact on the protective boundaries (e.g., fuel matrix and
cladding, reactor coolant system pressure boundary, and primary and
secondary containment), or on the safety limits for these
boundaries.
The locations used to monitor the ultimate heat sink temperature
will be maintained in the Bases. The Bases are a licensee controlled
document which is maintained under the requirements of 10CFR50.59.
Since any future changes to this license controlled document will be
evaluated per the requirements of 10CFR50.59, any possible reduction
(significant or insignificant) in a margin of safety will be
addressed.
Thus, the license amendment request does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270NRC Deputy Director: Phillip F. McKee

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

Date of amendment request: March 31, 1997
Description of amendment request: The proposed amendment would
modify Technical Specification Surveillance Requirement 4.7.1.2.1.b
which requires the testing of the auxiliary feedwater motor-driven and
turbine-driven pumps on recirculation flow at least once per 92 days.
The proposed amendment would also makes changes to the appropriate
Bases section.
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed changes in accordance with 10CFR
50.92 and has concluded that the changes do not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10CFR 50.92(c) are not
satisfied. The proposed changes do not involve [an] SHC because the
changes would not:

[[Page 19833]]

1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to Technical Specification Surveillance
4.7.1.2.1.b to increase the required test parameter for the motor
driven pumps from 1460 psid to 1468 psid and replacing the current
parameters for the motor driven and turbine driven pumps from
differential pressure measured in psid [pounds per square inch
differential] to total head measured in feet are consistent with
equipment design criteria and does not significantly increase the
probability of an accident previously evaluated.
The proposed changes to increase the required test parameter for
the motor driven pumps from 1460 psid to 1468 psid and replacing the
current parameters for the motor driven and turbine driven pumps
from differential pressure measured in psid to total head measured
in feet provides the necessary assurance that the pumps will
function as required in accident analyses and does not significantly
increase the consequence of an accident previously evaluated.
The moving of the reference to Specification 4.0.5 in order to
clarify that it applies to the testing of the motor driven and
turbine driven pumps and the modifications to the bases section are
administrative and do not involve a significant increase in the
probability or consequence of an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to Technical Specification Surveillance
4.7.1.2.1.b to increase the required test parameter for the motor
driven pumps from 1460 psid to 1468 psid and replacing the current
parameters for the motor driven and turbine driven pumps from
differential pressure measured in psid to total head measured in
feet does not change the operation of the auxiliary feedwater system
or any of its components during normal or accident evaluations.
The moving of the reference to Specification 4.0.5 in order to
clarify that it applies to the testing of the motor driven and
turbine driven pumps and the modifications to the bases section are
administrative and do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
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 a margin of safety.
The proposed change to Technical Specification Surveillance
4.7.1.2.1.b to increase the referenced total head of the motor
driven auxiliary feedwater pumps during surveillance testing
provides an acceptable margin between the required surveillance and
design pump performance to provide assurance that the pumps will
operate consistent with system evaluations and does not involve a
significant reduction in a margin of safety.
The change in the referenced units from differential pressure
measured in psid to total head measured in feet for the motor driven
auxiliary and turbine driven auxiliary feedwater pumps during
surveillance testing is to account for the effect of water density
on pump performance during each test and does not involve a
significant reduction in a margin of safety.
The moving of the reference to Specification 4.0.5 in order to
clarify that it applies to the testing of the motor driven and
turbine driven pumps and the modifications to the bases section are
administrative and do not involve a significant reduction in a
margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed changes do not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270NRC Deputy Director: Phillip F. McKee

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

Date of amendment request: March 31, 1997
Description of amendment request: The proposed amendment would
separate the required testing of motor-operated valve thermal overload
protection into two new surveillances.
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed change in accordance with
10CFR50.92 and has concluded that the change does not involve a
significant hazards consideration (SHC). The bases for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed change does not involve a SHC because the
change would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes to the surveillance testing of the motor-
operated valve thermal overload protection are consistent with
equipment design criteria and performing surveillance testing does
not significantly increase the probability of an accident previously
evaluated. The proposed changes to the surveillance testing provides
the necessary assurance that the motor operated valve thermal
overload protection will function as required and does not involve a
significant increase in the consequence of an accident previously
evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequence of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes to the surveillance testing of the motor-
operated valve thermal overload protection does not change the
operation of any system or system component during normal or
accident evaluations.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to the surveillance testing of the motor-
operated valve thermal overload protection are administrative in
that the changes to the surveillance only clarify that following
maintenance on the motor starter, a channel calibration is required
only on that valve. The surveillance continues to require periodic
representative sample testing.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
In conclusion, based on the information provided, it is
determined that the proposed change does not involve an SHC.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270NRC Deputy Director: Phillip F. McKee

[[Page 19834]]

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

Date of amendment request: March 4, 1997
Description of amendment request: The amendments would modify the
Emergency Core Cooling System (ECCS) surveillance test acceptance
criteria in Technical Specification 3/4.5.2 for the Centrifugal
Charging (CH) and the Safety Injection (SI) pumps. The changes to the
specified flow values would account for system alignments that effect
the suction pressure to the pumps. In the recirculation mode, increased
flow occurs when the CH and SI pumps take suction from the discharge of
the Residual Heat Removal pumps.
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 evaluations performed by Westinghouse determined
that, with the proposed changes, the subject pumps remain operable
and the safety analyses criteria remain valid.
Previous conclusions under LCR [License Change Request] 91-03
evaluating the consequences of the LOCA [loss-of-coolant-accident]
considered in the Salem Units 1 & 2 licensing basis remain
unchanged. With respect to the LOCA, the Peak Cladding Temperature
(PCT) continues to conform to the 10CFR50.46 guidelines of less than
2200*F. Evaluation of LOCA mass and energy releases previously found
acceptable remain valid. Decreasing the acceptance window to
accommodate the potential of an increase to pump runout flow,
assures that the current limits on pump runout flows continue to be
met. This change ensures pump integrity is maintained during the
accident. The reduction of the flow by throttling valves to
compensate for the potential suction boost remains within the
current analyses and therefore more conservative values are being
proposed. Additionally, the proposed change balances the pump flows
more appropriately by differentiating between the hot and cold leg
alignments. Flow to the reactor core is unaffected by the very
slight reduction in the upper flow limits. Since the design
limitations continue to be met and the integrity of the reactor
coolant system pressure boundary is not challenged, offsite dose
assumptions and calculations remain valid. Further, the ECCS is
post-accident mitigation system and probability of an accident is
not increased by this proposed change. Lastly, the correction of
double use of the word ``the'' in Salem Unit 1 Technical
Specification section 4.5.2.h.1.a is of editorial nature.
Based on the above information, the proposed changes do not
increase the risk or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. No new single failures are initiated. The proposed
changes will therefore not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change addresses suction boost by changing the
Technical Specification surveillance acceptance criteria. The
typographical correction is of editorial nature.
3. The proposed change does not involve a significant reduction
in a margin of safety. The evaluation of LOCA accident analysis
previously performed by Westinghouse continues to be met and
verifies that, with the proposed changes to the TS, plant operations
will be maintained within the bounds of safe, analyzed conditions as
defined in the UFSAR [Updated Final Safety Analysis Report] and that
conclusions presented in the UFSAR remain valid. The peak cladding
temperatures (PTC) remains unchanged as no effective differences in
the operating parameters have occurred. The typographical correction
is of editorial nature. The proposed changes will therefore 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: Salem Free Public library, 112
West Broadway, Salem, NJ 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz

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

Date of amendments request: March 7, 1997
Description of amendments request: The proposed amendments would
allow operability testing for the containment isolation valves listed
in Table 3.6-1 of the Technical Specifications during a defueled
status. These proposed changes are technically consistent with the
requirements of NUREG-1431, Revision 1, ``Westinghouse Standard
Technical Specifications,'' issued on April 7, 1995.
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.]
The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the FSAR [Final Safety Analysis Report]. The proposed changes have
no impact on the probability of an accident. The containment
isolation valves will continue to require operability testing.
Allowing the testing to be performed when the unit is in a defueled
status will have no impact on any accidents previously evaluated.
The net effect of these changes is not significant and, as a result,
does not involve a significant increase in the consequences of an
accident previously evaluated.
[2. Create the possibility of a new or different kind of
accident from any accident previously evaluated.]
The proposed changes to the Technical Specifications do not
increase the possibility of a new or different kind of accident than
any accident already evaluated in the FSAR. No new limiting single
failure or accident scenario has been created or identified due to
the proposed changes. Safety-related systems will continue to
perform as designed. The proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
[3. Involve a significant reduction in a margin of safety.]
The proposed changes do not involve a significant reduction in
the margin of safety. Although, as a result of these proposed
changes, the containment isolation valves could be tested for
operability while the unit is in a defueled state, there is no
impact in the accident analyses. These proposed changes are
technically consistent with the requirements of NUREG-1431, Revision
1 which has already received the requisite review and approval of
the NRC staff. Thus 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: Herbert N. Berkow

[[Page 19835]]

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: August 22, 1996, as supplemented on
March 28, 1997 (TS 96-02)
Description of amendment request: The proposed changes would revise
Section 3.6.5 of the Sequoyah Technical Specifications (TS) and
associated Bases to lower the minimum TS ice basket weight of 1,155
pounds to 1,071 pounds. This would reduce the overall weight of ice
required in the ice condenser from 2,245,320 pounds to 2,082,024
pounds. The TVA license amendment request also proposed to extend the
chemical analysis surveillance interval for the ice condenser ice bed
from 12 months to 18 months based on the provisions of Generic Letter
93-05.
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of Sequoyah Nuclear Plant (SQN) in accordance with the
proposed amendment will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
TVA proposes to modify the SQN Unit 1 and Unit 2 TSs [Technical
Specifications] to revise Surveillance Requirement (SR) 4.6.5.1.d to
lower SQN's minimum TS basket weight from 1,155 pounds (lbs) to
1,071 lbs, thus lowering the overall ice condenser weight from
2,245,320 lbs to 2,082,024 lbs.
The ice condenser system is provided to absorb thermal energy
release following a loss-of-coolant accident (LOCA) or high energy
line break (HELB) and to limit the peak pressure inside containment.
The current containment analysis for SQN is based on a minimum of
993 lbs of ice per basket evenly distributed throughout the ice
condenser at the end of an 18-month refueling cycle. The revised
containment analysis shows that for the predicted sublimation rate
of 15 percent for 18 months, an average basket weight of 922 lbs at
the end of the 18-month period would ensure containment design
pressure is not exceeded.
Based on TVA's evaluation and the revised containment analysis,
TVA considers the reduction of ice weight to be acceptable for
satisfying the safety function of the ice condenser for an 18-month
ice weighing interval. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
TVA is also proposing to extend the surveillance interval as it
pertains to the ice bed chemical analysis. Based on test results,
both at SQN and the industry, the average boron concentration and pH
changes are minimal; therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Elimination of the temperature at which the pH of the ice bed is
determined is an administrative change. Future testing will be
accomplished in accordance with American Society for Testing and
Materials Standards recommendations. Therefore, this change cannot
increase the probability of an accident and the consequences of an
accident will not increase.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
TVA's request to lower the TS limit for ice weight at the start
of the surveillance interval will not result in a new or different
kind of accident from that previously analyzed in SQN's Final Safety
Analysis
Report. SQN's ice condenser serves to limit the peak pressure
inside containment following a LOCA. TVA has evaluated the revised
containment pressure analysis for SQN (Enclosure 4, Westinghouse
WCAP-12455, Revision 1) and determined that sufficient ice would be
present at all times to keep the peak containment pressure below
SQN's containment design pressure of 12 pounds per square inch gage
(psig). Therefore, this change would not result in a new or
different kind of accident from any previously analyzed.
The proposed reduced testing frequency of the chemical
composition of the ice bed does not change the manner in which the
plant is operated. Additionally, the ice condenser is a passive
system that reacts to an accident, but does not support plant
operation on a daily basis. The reduced testing frequency of the ice
bed chemical composition does not generate any new accident
precursors; therefore, the possibility of a new or different kind of
accident from any previously analyzed is not created.
Elimination of the temperature at which the pH of the ice bed is
determined is an administrative change. This change cannot create
the possibility of a new or different kind of accident.
3. Involve a significant reduction in a margin of safety.
The ice condenser system is provided to absorb thermal energy
release following a LOCA and to limit the peak pressure inside
containment. The current ice condenser analysis for SQN is based on
a minimum of 993 lbs of ice per basket. The revised containment
analysis changes the minimum ice weight assumed in the analysis to
922 lbs per basket.
The revised containment analysis shows that using an average
basket weight of 1,071 lbs and a sublimation allowance of 15
percent, all bays would have an average basket weight of 922 lbs at
the end of the 18-month surveillance interval. The revised analysis
utilizes new mass and energy releases (refer to Westinghouse WCAP-
10325-P-A), which substantially delays ice-bed meltout and limits
the initial containment peak pressure to approximately 7.15 psig
during the blowdown phase. The ice-bed meltout delay allows the
second containment pressure peak, which is driven mainly by the
decay heat, to be limited to approximately 11.45 psig, which is
below the containment design pressure of 12 psig.
Based on TVA's evaluation and the revised containment analysis,
TVA considers the reduction of the average basket weight to be
acceptable for satisfying the safety function of the ice condenser
for the current 18-month interval. Therefore, the proposed change
does not involve a significant reduction in the margin of safety.
The proposal to extend the surveillance from 12 to 18 months
does not change the boron concentration or pH requirements.
Experience at Duke Power Company, as stated in NUREG-1366, indicates
that these parameters do not change appreciably when verified every
9 months. SQN has a similar experience with a 12-month interval.
Since the boron concentration and the post-LOCA pH requirements
remain essentially the same, there is no reduction in the margin of
safety.
Elimination of the temperature at which the pH of the ice bed is
determined is an administrative change. Future testing will be
accomplished in accordance with ASTM recommendations. The difference
between the pH values determined at the current TS specified
temperature and the temperature currently recommended by the ASTM
standards is insignificant. Therefore, there is no reduction in the
margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon

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

Date of amendment request: September 12, 1996
Description of amendment request: The proposed change to the
Technical Specifications is administrative in nature in that it would
add the NRC standard fire protection license condition to each unit's
Operating License and relocate the fire protection requirements from
the Technical Specifications to the Updated Final Safety Analysis
Report (UFSAR).
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:

[[Page 19836]]

Specifically, operation of Surry Power Station with the proposed
amendment will not:
1. Involve a significant increase in either the probability of
occurrence or consequences of any accident or equipment malfunction
scenario that is important to safety and which has been previously
evaluated in the UFSAR. The requirements of the Fire Protection
Program have not been changed by theproposed amendment. Relocation
of the Fire Protection Program requirements into the UFSAR and
station procedures does not decrease any portion of the program. The
same fire protection requirements exist as before the change.
2. Create the possibility of a new or different type of accident
than those previously evaluated in the safety analysis report. The
requirements of the Fire Protection Program have not been changed by
the proposed amendment. This is an administrative change to relocate
the Fire Protection Program requirements from the Technical
Specifications to the UFSAR and station procedures. Consequently,
the possibility of a new or different kind of accident from any
accident previously evaluated has not been created.
3. Involve a significant reduction in a margin of safety.
Implementation of the Fire Protection Program requirements is
assured by the UFSAR and station procedures. Since the rogram is
being retained intact, there is no reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219
NRC Project Director: Mark Reinhart, Acting

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin

Date of amendment request: January 16, 1997
Description of amendment request: The proposed amendments (Point
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request
(TSCR) 191) would revise the minimum boron concentration required in
the refueling water storage tank(s)(RWST), boric acid storage tanks
(BAST), and safety injection (SI) accumulators during normal operation;
the minimum boron concentration of primary coolant during refueling
conditions; and the minimum boron concentration in the reactor when
positive reactivity could be added and/or boron dilution could occur
and containment integrity is not intact. These changes are necessary to
accommodate the planned extension of the operating cycle from 12 months
to 18 months. The licensee proposes to change TS 15.3.2, ``Chemical and
Volume Control System,'' TS 15.3.3, ``Safety Injection and Residual
Heat Removal Systems,'' TS 15.3.6, ``Containment System,'' TS 15.3.8,
``Refueling,'' and associated Bases.
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of this facility under the proposed Technical
Specifications will not create a significant increase in the
probability or consequences of an accident previously evaluated.
The probabilities of accidents previously evaluated are based on
the probability of initiating events for these accidents. Initiating
events for accidents previously evaluated are described in the PBNP
FSAR [final safety analysis report].
In effect, the proposed changes will result in: (1) higher boron
concentrations of primary coolant during refueling, and (2) higher
boron inventories in the RWSTs, BASTs, and SI accumulators. These
changes do not require hardware changes or changes to the operation
of accident-mitigating equipment. These changes relate to the
performance capability of particular accident mitigation systems;
equipment that is not postulated to cause accidents. Therefore,
these proposed changes do not cause an increase in the probabilities
of any accidents previously evaluated.
The consequences of accidents previously evaluated in the PBNP
FSAR are determined by the results of analyses that are based on
initial conditions of the plant, the type of accident, transient
response of the plant, and the operation and failure of equipment
and systems.
In effect, the proposed changes will result in: (1) higher boron
concentrations of primary coolant during refueling, and (2) higher
boron inventories in the RWSTs, BASTs, and SI accumulators. These
increased boron concentrations do not increase the probability that
engineered safety features equipment will fail, nor do these changes
affect the capability of this equipment to operate as required for
the accidents previously evaluated in the PBNP FSAR. These changes
do not require hardware changes or changes to the operation of
accident-mitigating equipment.
The consequential effects of a lower containment spray pH will
not affect the capability of the containment spray to remove
elemental iodine during design basis LOCA [loss-of-coolant accident]
accidents. Also, the consequential reduction in containment sump
water pH will not affect the fluid's capability to retain elemental
iodine, nor will it adversely increase the potential corrosion rates
for materials inside containment if the sump water is sprayed into
containment during the recirculation phase of a LOCA.
Another consequence of injecting a higher concentration boric
acid solution into the core during a LOCA may be an abbreviated
onset to boron precipitation in the post-LOCA core. An incremental
change in the boron injection concentration would not have
significant effect on the postulated onset, but each core reload
safety evaluation will continue to verify that the existing
emergency operating procedures accommodate the potential for boron
precipitation.
Therefore, this proposed license amendment does not affect the
consequences of any accident previously evaluated in the PBNP FSAR,
because the factors that are used to determine the consequences of
accidents are not changed.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any previously evaluated.
New or different kinds of accidents can only be created by new
or different accident initiators or sequences. New and different
types of accidents (different from those that were originally
analyzed for Point Beach) have been evaluated and incorporated into
the licensing basis for PBNP. Examples of different accidents that
have been incorporated into the PBNP licensing basis include
anticipated transients without scram and station blackout.
The changes proposed by this TSCR do not create any new or
different accident initiators or sequences because these changes to
minimum boron concentrations will not cause failures of equipment or
accident sequences different than the accidents previously analyzed.
No new equipment interfaces are created, and no new materials or
fluids are introduced. The incremental increase in boron
concentrations will not create a failure mechanism not previously
known and evaluated. Therefore, these proposed TS changes do not
create the possibility of an accident of a different type than any
previously evaluated in the PBNP FSAR.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The margins of safety for Point Beach are based on the design
and operation of the reactor and containment and the safety systems
that provide their protection. Plant safety margins are established
through Limiting Conditions for Operation, Limiting Safety System
Settings and Safety Limits specified in the Technical
Specifications. The proposed Technical Specification changes to
refueling water storage tank (RWST), SI accumulator, and BAST boron
inventory requirements have all been evaluated to preserve the
shutdown capability described in the associated bases (boration from
just critical, hot zero or full power, peak xenon with control rods
at the

[[Page 19837]]

insertion limit, to xenon-free cold shutdown with the highest worth
control rod assembly fully withdrawn). Similarly, the proposed TS
change to the minimum boron concentration of the primary coolant
system for refueling operations have been evaluated to preserve the
subcriticality margin described in the associated TS bases (i.e., 5%
[delta] k/k in the cold condition with all rods inserted).
Because there are no changes to any of these margins, the
proposed license amendment does not involve a reduction in any
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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John N. Hannon

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin

Date of amendment request: January 21, 1997
Description of amendment request: The proposed amendments (Point
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request
195) would revise TS Section 15.6.11, ``Radiation Protection Program,''
to update all references to 10 CFR Part 20, ``Standards for Protection
Against Radiation,'' to restore consistency between 10 CFR Part 20
regulations and the PBNP TS.
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendments are administrative in nature, providing
consistency between the Point Beach licenses and Commission
regulations. The amendments do not affect the operation or
maintenance of any PBNP structure[,] system or component. In
addition, the regulations and proposed changes provide more
conservative determinations of high radiation areas, thereby
potentially resulting in lower personnel radiation exposures during
normal operation and post accident. The consequences of an accident
related to personnel radiation exposures may be reduced.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed amendments are administrative only and do not
affect the operation or maintenance of any structure[,] system or
component at Point Beach Nuclear Plant. No new systems or components
are introduced. Therefore, no new accident initiators or sequences
result from any previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create a significant reduction in a
margin of safety.
The proposed amendments are administrative and reflect
regulatory requirements that are more conservative than those
presently reflected in the PBNP Technical Specifications. These more
conservative requirements result in more conservative designation of
high radiation areas thereby providing additional margins of safety
related to the control of radiation exposures to personnel. No
structure[,] system or component at PBNP at PBNP is changed[,]
thereby maintaining the margins of safety for the operation of the
Point Beach Nuclear Plant.
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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John N. Hannon

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin

Date of amendment request: January 24, 1997
Description of amendment request: The proposed amendments (Point
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request
(TSCR) 193) would revise TS 15.5.4, ``Fuel Storage,'' to increase fuel
assembly enrichment limits to 5.0 w/o U-235 while maintaining Keff in
the storage pools (spent fuel pool and new fuel storage racks) less
than 0.95.
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 this facility under the proposed Technical
Specifications will not create a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes do not involve a change to structures,
systems, or components which would affect the probability or
consequences of an accident previously evaluated in the PBNP Final
Safety Analysis Report (FSAR). The only relevant concern with
respect to increasing enrichment limits in the spent fuel pool and
new fuel storage racks is one of criticality. The proposed changes
use the same criticality limit used in the current Technical
Specifications. Therefore, margin to safe operation of Units 1 and 2
is maintained. The probability and consequences of an accident
previously evaluated are dependent on this criticality limit.
Because the limit will not change, the probability and consequences
of those accidents previously evaluated will not change.
2. Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve a change to plant design.
The proposed increase in spent fuel pool and new fuel storage racks
fuel assembly enrichment limits maintains the margin to safe
operation of Units 1 and 2 because the criticality limit for the
spent fuel pool and new fuel storage racks will not change. These
changes do not affect any of the parameters or conditions that
contribute to the initiation of any accidents. Because the
criticality limit remains the same, these changes have no effect on
plant operation, design, or initiation of any accidents. Therefore,
the proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety.
The proposed changes maintain the margin to safe operation of
Units 1 and 2. The margin of safety is based on the criticality
limit of the spent fuel pool and the new fuel storage racks. Because
this limit will not change, the margin of safety will not be
affected. Therefore, the proposed changes will not create 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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241

[[Page 19838]]

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John N. Hannon

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin

Date of amendment request: February 12, 1997, as supplemented on
March 11, 1997
Description of amendment request: The proposed amendments (Point
Beach Nuclear Plant (PBNP) Technical Specifications (TS) Change Request
196) would relocate turbine overspeed protection specifications,
limiting conditions for operation, surveillance requirements, and
associated bases from TS Section 15.3.4, ``Steam and Power Conversion
System,'' and Section 15.4.1, ``Operational Safety Review,'' to the
Final Safety Analysis Report (FSAR) in accordance with Generic Letter
95-10, ``Relocation of Selected Technical Specifications Requirements
Related to Instrumentation.''
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 Point Beach Nuclear Plant in accordance with the
proposed amendments will not result in a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments administratively relocate turbine
overspeed protection Specifications to the Point Beach Final Safety
Analysis Report (FSAR). The Specifications will be transferred
verbatim, except for the turbine load limit with the crossover steam
dump system inoperable, which has already been evaluated under 10
CFR 50.59 and will be conservatively reduced. In addition, the
regulatory requirements of 10 CFR 50.55a, ``Codes and Standards, ''
will still apply to the relocated Specifications. Therefore,
operation of Point Beach Nuclear Plant in accordance with the
proposed amendments cannot create an increase in the probability or
consequences of an accident previously evaluated.
2. Operation of Point Beach Nuclear Plant in accordance with the
proposed amendments will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed amendments administratively relocate Specifications
to the FSAR and in one case result in a more conservative operating
limit. Therefore, operation of Point Beach Nuclear Plant in
accordance with the proposed amendments cannot create a new or
different kind of accident from any accident previously evaluated.
3. Operation of Point Beach Nuclear Plant in accordance with the
proposed amendments will not create a significant reduction in a
margin of safety.
The proposed changes are administrative in nature. There is no
physical change to the facility, its systems, or its operation,
except for the conservative reduction of the turbine load limit with
the crossover steam dump system inoperable which has already been
justified via 10 CFR 50.59. Therefore, operation of PBNP in
accordance with the proposed amendments cannot 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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: John N. Hannon

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas

Date of amendment request: February 17, 1997; supersedes March 24,
1995, as supplemented by letter dated August 16, 1995, amendment
request.
Description of amendment request: This amendment request proposes
to revise Technical Specification 1.7, ``Containment Integrity,''
Technical Specification 3/4.6.1, ``Containment Integrity,'' and
Technical Specification 3/4.6.3, ``Containment Isolation Valves.''
These proposed changes would relocate Technical Specification Table
3.6-1, ``Containment Isolation Valves,'' to the Wolf Creek Generating
Station (WCGS) procedures. This proposed change is in accordance with
the guidance provided in Generic Letter 91-08, ``Removal of Component
Lists from Technical Specifications,'' dated May 6, 1991. In addition,
this request proposes that the August 16, 1996, supplemental submittal
that provided an additional footnote allowing an increased outage time
for certain component cooling water system valves be withdrawn. The
determination that the additional footnote is not required supersedes
the staff's proposed no significant hazards consideration determination
evaluation for the requested changes that was published on September
27, 1995 (60 FR 49949).
Basis for proposed no significant Hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes simplify the technical specifications, meet
the regulatory requirements for control of containment isolation,
and are consistent with the guidelines of GL 91-08. The procedural
details of Technical Specification Table 3.6-1 have not been
changed, but only relocated to a different controlling document. The
proposed changes are administrative in nature, should result in
improved administrative practices, and do not affect plant
operations.
The probability of occurrence of a previously evaluated accident
is not increased because this change does not introduce any new
potential accident initiating conditions. The consequences of an
accident previously evaluated is not increased because the ability
of containment to restrict the release of any fission product
radioactivity to the environment will not be degraded by this
change.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature, do not result
in physical alterations or changes to the operation of the plant,
and cause no change in the method by which any safety-related system
performs its function. Therefore, this proposed change will not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The administrative change to relocate Technical Specification
Table 3.6-1 to appropriate plant procedures does not alter the basic
regulatory requirements for containment isolation and will not
adversely affect containment isolation capability for Coordinator
credible accident scenarios. Adequate control of the content of the
table is assured by existing plant procedures.
The proposed relocation of Technical Specification Table 3.6-1
does not alter current technical specification requirements for
containment isolation valve operability. The LCO and Surveillance
Requirements would be retained in the revised technical
specifications. Therefore, the proposed change will not affect the
meaning, application, and function of the current technical
specification requirements for the valves in Table 3.6-1.
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 19839]]

Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: William H. Bateman

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas

Date of amendment request: March 18, 1997
Description of amendment request: This license amendment request
revises Technical Specification Surveillance Requirement 4.5.2.c to
clarify when a containment entry visual inspection is required. This
proposed change to reduce the visual inspection requirement to at least
once daily is in accordance with the guidance provided in Generic
Letter 93-05, ``Line-Item Technical Specifications Improvements to
Reduce Surveillance Requirements for Testing During Power Operation.''
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.
Implementing the proposed change could potentially increase the
chances of loose debris (trash, rags, clothing, etc.) being left in
containment for some period of time greater than would be allowed
under current surveillance requirements. However, the proposed
change also clarifies that the visual inspection must be performed
at least once daily. Therefore, the period of time that debris could
be left uncontrolled inside containment would still be less than 24
hours. Based on work controls placed on material entry/exit into
containment and personnel training on housekeeping controls inside
containment, and the results of past surveillances, it is unlikely
that a significant amount of debris would be left uncontrolled
inside containment for this period of time. Also, based on
containment sump design, relatively small amounts of debris would
not be sufficient to cause a significant amount of blockage of the
sump screens.
The probability of occurrence of a previously evaluated accident
is not increased because the reduced frequency of the visual
inspection does not cause a significant impact on the possibility of
containment sump screen blockage. Therefore containment sump
operability is not affected by the proposed change. In addition, the
proposed change will not result in any changes to the design or
operation of any plant systems or components.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change decreases the frequency of performing a
visual inspection for loose debris in containment, but does not
result in a change to the design or operation of any plant system or
component. The purpose of the inspection is to ensure that there is
no loose debris, left in containment following a containment entry,
that could potentially block the containment sump screens during
LOCA conditions. Delaying this inspection until the last containment
entry each day will not result in a significant amount of debris
being left in containment, based on housekeeping practices
controlling the entry/removal of trash and material into/from
containment; training of employees to increase awareness of material
control in radiologically-controlled areas; and retaining the
requirement to perform a visual inspection at least once per day
when containment entries are made (during periods when containment
integrity is established), thereby ensuring that trash and debris
can be identified and removed on a daily basis (on days containment
entries are made).
Based on the above, this proposed change will not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The purpose of performing a visual inspection of areas affected
by a containment entry is to ensure any debris or trash generated by
the activity during the containment entry is identified and removed
from containment. This ensures that no trash or debris is left in
containment that could be transported to and block the containment
sump screens during LOCA conditions. Based on current material
control and housekeeping practices imposed on containment entry/
exit, and past inspection results, reducing the surveillance
requirement to a once per day basis, on days containment entries are
made, would not result in a significant amount of trash or debris
being left in containment following completion of the entry, and any
such material would be identified and removed prior to the end of
the day. The amount of trash or debris that could be left in
containment for a 24 hour period would be significantly less than
the amount that would be required to cause sump screen blockage
sufficient to affect sump performance. Therefore, the proposed
change will not result in a significant reduction in the

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Source: Frix Law Library, https://www.frixlaw.com/law-library/documents/fr%3AX97-10423. Public record. Not legal advice.