98-26746. Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

[Federal Register Volume 63, Number 194 (Wednesday, October 7, 1998)]
[Notices]
[Pages 53943-53968]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-26746]

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NUCLEAR REGULATORY COMMISSION

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

I. Background

Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission
(the Commission or NRC staff) is publishing this regular biweekly
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of
1954, as amended (the Act), to require the Commission to publish notice
of any amendments issued, or proposed to be issued, under a new
provision of section 189 of the Act. This provision grants the
Commission the authority to issue and make immediately effective any
amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from September 14, 1998, through September 25,
1998. The last biweekly notice was published on September 23, 1998 (63
FR 50932).

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

The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period.

[[Page 53944]]

However, should circumstances change during the notice period such that
failure to act in a timely way would result, for example, in derating
or shutdown of the facility, the Commission may issue the license
amendment before the expiration of the 30-day notice period, provided
that its final determination is that the amendment involves no
significant hazards consideration. The final determination will
consider all public and State comments received before action is taken.
Should the Commission take this action, it will publish in the Federal
Register a notice of issuance and provide for opportunity for a hearing
after issuance. The Commission expects that the need to take this
action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By November 6, 1998, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.

[[Page 53945]]

Duke Energy Corporation, 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: October 28, 1997, as supplemented March
26, May 20, July 29, and August 13, 1998.
Description of amendment request: The proposed amendments would
revise the current Technical Specifications (CTS) of each unit to
conform with NUREG-1430, ``Standard Technical Specifications--Babcock
and Wilcox Plants.'' The Commission had previously issued a Notice of
Consideration of Issuance of Amendments published in the Federal
Register on December 5, 1997 (62 FR 64405), covering all of the
proposed Improved Technical Specification (ITS) changes that were
within the scope of NUREG-1430 for the Oconee Nuclear Station. However,
the submittals also contained proposed changes that are beyond the
scope of NUREG-1430, which were not included in the staff's December 5,
1997, notice. The following descriptions and proposed no significant
hazards analyses cover only the beyond-scope changes. Associated with
each proposed change are administrative/editorial changes such that the
new or revised requirements would fit into the format of NUREG-1430.
Some changes are ``Less Restrictive'' (meaning that the new
requirements being incorporated into the ITS are less restrictive than
the CTS requirements) and some are ``More Restrictive.'' The basis for
the no significant hazards determination is identical for all of the
more restrictive items and is presented at the end of the following
list of more restrictive beyond-scope items:
A. Certain NUREG and CTS Sections 3.1.3.5, 3.5.2.4.a, 3.5.2.5.b,
3.5.2.5.c, and 3.5.2.6, specify that they are applicable ``except
during Mode 1 physics testing.'' The exception would not be included in
the ITS and, therefore, the Mode 1 requirement would be applicable
during the tests. The proposed change is more conservative since no
exceptions would be allowed for physics tests conducted in Mode 1.
B. CTS 3.1.3.2 requires reactor coolant temperature to be greater
than the criticality values of specified heatup limitation curves. This
requirement would not be retained in the ITS. ITS 3.1.8, Limiting
Condition for Operation (LCO) Part e, would be added to provide a
restriction for loop average temperature to be greater than or equal to
520 deg.F when performing physics tests in Mode 2. ITS LCO 3.1.8 would
permit suspending the requirements of ITS LCO 3.4.2, ``RCS (reactor
coolant system) Minimum Temperature for Criticality,'' during physics
tests initiated in Mode 2. Associated Actions and a surveillance
requirement (SR) would be added to provide an appropriate required
action when outside the limit and to verify operation within the limit
periodically.
C. CTS Table 3.5.1-1 presently requires that the operator place the
plant in hot shutdown (ITS equivalent of Mode 3) within 12 hours when
the minimum channels Operable requirement is not met. The proposed
change to the ITS would provide an equivalent requirement and add a
requirement to open all control rod drive (CRD) trip breakers within 12
hours. ITS 3.3.3 Action B, and ITS 3.3.4 Action D, would be added to
require that the unit be in Mode 3 in 12 hours with all CRD trip
breakers open or that power be removed from all CRD trip breakers when
the required action and associated completion time is not met in Mode
1, 2, or 3. For ITS 3.3.3, Action B would also apply when two or more
reactor trip modules are inoperable in Mode 1, 2, or 3. The CTS
presently requires entry into TS 3.0, which requires that the reactor
be in hot shutdown (equivalent to ITS Mode 3) in 12 hours.
D. Note c would be added to ITS Table 3.3.8-1, Post Accident
Monitoring Instrumentation, and referenced to Item No. 8, Containment
Isolation Valve Position, to specify that position indication
requirements apply only to the Containment Isolation Valves that are
electrically controlled.
E. The applicability of Table 3.5.1-1 would be expanded to require
wide range instruments to be operable in Mode 2, plus Modes 3, 4, and
5, with any control rod drive trip breaker in the closed position and
the control rod drive system capable of rod withdrawal. In addition, a
Note would define the upper limit of the applicable Modes for the
required wide range instrument channels as being 10 percent indicated
neutron power.
F. The applicability of ITS 3.3.14 would be expanded to include
Mode 4 when the steam generator is relied upon for heat removal, which
then would be consistent with the applicability of ITS LCO 3.7.5 for
the emergency feedwater (EFW) system. ITS Specifications 3.3.14 and
3.3.15 would be added to address EFW system initiation circuitry and
main steamline break and main feedwater isolation instrumentation
separately. The specification titles, LCOs, actions, and SRs would be
modified to reflect Oconee-specific terminology and design
requirements. Where appropriate, ITS-required actions would be based on
similar NUREG-required actions. EFW pump initiation circuitry operable
requirement would be changed from 250 deg.F to greater than or equal
to 246 deg.F.
G. ITS LCO 3.4.1, Departure from Nucleate Boiling Ratio (DNBR)
Limits, are specified in the core operating limits report rather than
in the LCO and SRs since they are subject to change with fuel cycle
designs. The ITS LCO 3.4.1 actions would require restoring DNBR
parameters to within limits within 2 hours or exiting the applicability
for the specification within 12 additional hours. ITS SR 3.4.1.1, SR
3.4.1.2, and SR 3.4.1.3 would require verification that each DNBR
parameter is within the limit at a 12-hour frequency. ITS SR 3.4.1.4
would require verification by measurement that total RCS flow is within
limit at an 18-month frequency. Specification 3.4.1 would ensure that
limits on RCS pressure, temperature, and flow rate are met to ensure
that the core operates within the limits assumed for the plant safety
analyses. These changes are more restrictive.
H. The NUREG allowed time to complete the SR after addition to core
flood tank (CFT) of 6 hours would be changed to 12 hours. ITS SR
3.5.1.4 would require CFT boron concentration be sampled every 31 days
or once within 12 hours after each solution volume increase greater
than or equal to 80 gallons that is not the result of addition from a
borated water source that meets CFT boron concentration requirements.
Since the CTS does not specify the time limit following addition, the
proposed ITS change is a more restrictive limit.
I. ITS 3.5.3 LCO Note 3 would be added to explicitly require that
the low pressure injection (LPI) discharge header crossover valves be
operable and capable of being opened manually when in Modes 1, 2, and
3. ITS 3.5.3 Action B would require that the LPI discharge header
crossover valves be restored to operable status within 72 hours of
being discovered incapable of being manually opened when in Modes 1, 2,
and 3. ITS 3.5.3 Action D would require LCO 3.0.3 be entered
immediately when one LPI train is inoperable in Modes 1, 2, and 3
concurrent with discovery that the LPI discharge header crossover
valves are incapable of being opened manually in Modes 1, 2, and 3.
J. ITS 3.5.3 would require the LPI system to be operable in Modes
1, 2, 3, and 4. LCO Note 1 would be added to specify that only one LPI
train is required to be operable in Mode 4. LCO Note 2 would be added
to allow an LPI train to be considered operable during

[[Page 53946]]

alignment, when aligned, or when operating if capable of being manually
realigned to the LPI mode of operation. Action E would be added to
require action be initiated immediately to restore the required LPI
train to operable status and to require the reactor to be placed in
Mode 5 within 24 hours when the required LPI train cannot be restored
to OPERABLE status (provided a decay heat removal loop is available).
K. SR 3.9.4.1 would be modified to eliminate verification of a
specific decay heat removal flow rate to verification every 12 hours
that one decay heat removal loop is in operation.
L. Main feeder bus monitoring panel requirements and allowed outage
time would be added to the ITS.
M. TS Section 3.7 would be revised to include the actual trip
setpoint and/or allowable values for the loss of power sensing relays.
N. Battery performance discharge testing as related to battery
operability would be added.
O. Battery charger testing, cell-to-cell resistance measurements,
and battery discharge and overcharge conditions, surveillances would be
added to ITS Section 3.8.
P. High Pressure Injection System discharge pressure allowable
value in ITS Table 3.3.5-1 would be changed from 1500 pounds per square
inch gauge (psig) to 1590 psig.

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 for the More Restrictive Items listed above, as follows:

In accordance with the criteria set forth in 10 CFR 50.92, Duke
Energy has evaluated these proposed Technical Specification changes
and determined that they do not represent a significant hazards
consideration. The following is provided in support of this
consideration.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes provide more stringent requirements than
previously existed in the Technical Specifications. These more
stringent requirements do not result in operation that will increase
the probability of initiating an analyzed event. If anything the new
requirements may decrease the probability or consequences of an
analyzed event by incorporating the more restrictive changes. The
changes do not alter assumptions relative to mitigation of an
accident or transient event. The more restrictive requirements
continue to ensure process variables, structures, systems, and
components are maintained consistent with the safety analyses and
licensing basis. Therefore, the 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 the accidents previously evaluated?
The proposed changes provide more stringent requirements than
previously existed in the Technical Specifications. The changes do
not alter the plant configuration (no new or different type of
equipment will be installed) or make changes in the methods
governing normal plant operation. The changes do impose different
requirements. However, these changes are consistent with the
assumptions in the safety analyses and licensing basis. Therefore,
the changes do not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed changes provide more stringent requirements than
previously existed in the Technical Specifications. Adding more
restrictive requirements either increases or has no impact on the
margin of safety. The changes, by definition, provide additional
restrictions to enhance plant safety. The changes maintain
requirements within the safety analyses and licensing basis. As
such, no question of safety is involved. Therefore, the changes do
not involve a significant reduction in a margin of safety.

For the less restrictive beyond-scope items, the basis for the no
significant hazards consideration is unique for each item. The beyond-
scope item and the licensee's basis supporting its determination that
the proposed changes do not represent a significant hazards
consideration follow:
A. A proposed change to the Note for ITS SR 3.1.4.3 would provide
the additional flexibility for testing control rod drop times with
reactor coolant flow conditions other than full flow, but with at least
one reactor coolant pump (RCP) pump running. This would ensure that the
testing is bounding by restricting operation of the unit to the RCP
combination used during control rod drop testing and represents
adoption of the NUREG rather than the CTS.
Basis for proposed no significant hazards consideration
determination:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The control rods are used to support mitigation of the
consequences of an accident; however, the control rod drop time
variations are not considered the initiator of any previously
analyzed accident. As such the proposed change in the method of
performing the control rod drop time testing will not increase the
probability of any accident previously evaluated. The proposed
changes allow for testing the control rod drop times with less than
a full complement of reactor coolant pumps operating. However, the
operation of the plant is restricted to the pump combinations
providing maximum flow less than or equal to the pump flow used for
the testing. Therefore, the drop times verified during testing will
remain valid for mitigating the consequences of any accident
previously evaluated. Therefore, this change does not involve an
increase in the consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from the accidents previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will continue to ensure that the
control rods are available for insertion of reactivity in the time
frames consistent with the safety analysis. Thus, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The margin of safety provided in the acceptable control rod drop
times continues to be provided since these drop times have not been
changed. The surveillance methodology is revised to allow testing
with one, two, or three pumps operating. However, the operation of
the plant is restricted to the reactor coolant pump combinations
which maintain the margin of safety, i.e., those pump combinations
providing maximum flow less than or equal to the pump flow used for
the testing. Therefore, this change does not involve a significant
reduction in a margin of safety.

B. Required Action B.2.2 of ITS 3.3.11, 12, and 13, would be added
to provide the option of closing the main feedwater control valves
(MFCVs) and startup feedwater control valves (SFCVs) in lieu of
reducing main steam header pressure to less than 700 psig.
Applicability would be changed to Modes 1 and 2, plus Mode 3 when the
main steam header pressure is greater than 700 psig except when all
MFCVs and SFCVs are closed.
Basis for proposed no significant hazards consideration
determination:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The MSLB (main steamline break) and MFW (main feedwater)
Isolation circuitry is not an initiator of analyzed events.
Therefore, the probability of an accident is independent of the
status of the MSLB and MFW Isolation circuitry. As such the proposed
change does not involve a significant increase in the probability of
an accident previously evaluated. The proposed change eliminates the
requirement for MSLB and MFW Isolation circuitry OPERABILITY when
all the MFCVs and SFCVs are closed. When the MFCVs and SFCVs are
closed the MSLB and MFW Isolation circuitry has no safety function
since its function is to close the MFCVs and SFCVs when conditions
indicate [an] MSLB. Therefore, the change does not involve a
significant increase in the

[[Page 53947]]

consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from the accidents previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. Thus, this change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Since MSLB and MFW Isolation circuitry requirements continue to
require OPERABILITY when the reactor is in a condition that requires
their function, the proposed change does not involve a significant
reduction in a margin of safety.

C. ITS 3.3.15 Action A.1 would be added to allow 1 hour to declare
the turbine stop valves (TSVs) inoperable prior to requiring that the
unit shut down when one or more TSV closure channels is inoperable. ITS
Specifications 3.3.14 and 3.3.15 would be added to address the
emergency feedwater system initiation circuitry and main steamline
break and main feedwater isolation instrumentation separately. The
NUREG specification combines the emergency feedwater system initiation,
main steamline isolation, and main feedwater isolation functions into
one specification. The specification titles, LCOs, actions, and SRs
would be modified to reflect Oconee-specific terminology and design
requirements. Where appropriate, ITS-required actions would be based on
similar NUREG-required actions.
Basis for proposed no significant hazards consideration
determination:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
This change establishes a 1 hour Completion Time during which
the unit may continue operation with MSLB and MFW Isolation
instrumentation inoperable. This change provides an opportunity to
repair the inoperable instrumentation channel(s) prior to declaring
the equipment supported by it inoperable. The addition of this
allowed condition with a short Completion Time does not result in
any hardware changes. The allowed condition also does not
significantly increase the probability of occurrence for initiation
of any analyzed event since the function of the equipment does not
change (and therefore any initiation scenarios are not changed).
Further, the consequences of an accident are the same during the
additional one hour time period allowed for instrument channel
restoration as it is during the time period currently allowed for
restoring TSVs to OPERABLE status. Therefore, the change does not
significantly increase the probability of occurrence of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from the accidents previously evaluated?
The change does not necessitate a physical alteration of the
plant (no new or different type of equipment will be installed) or
changes in parameters governing normal plant operation. The change
continues to ensure prompt restoration of compliance with the
limiting condition for operation, or prompt and appropriate
compensatory actions are taken. Thus, this change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Prompt and appropriate Required Actions have been determined
based on the safety analysis functions to be maintained. The allowed
condition has been determined appropriate based on a combination of
the time required to perform the action, the relative importance of
the function or parameter to be restored, and engineering judgment.
Therefore, this new allowed condition does not involve a significant
reduction in the margin of safety.

D. CTS 3.8.10 and 4.4.4.5 frequency would be changed from ``* * *
immediately prior to refueling operation'' to ``Once each refueling
outage prior to CORE ALTERATIONS or movement of irradiated fuel
assemblies within containment'' in ITS SR 3.3.16.2 for testing
frequency of the radiation monitor associated with the purge system
valve isolation and ITS SR 3.9.3.2 for testing isolation function of
the reactor building purge supply and exhaust valves.
Basis for proposed no significant hazards consideration
determination:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures or components, changes in parameters
governing normal plant operation, or methods of operation. The
isolation function of the radiation monitor associated with the
purge system valves is not assumed to be an initiator of any
analyzed event. As a result, the probability of an accident
occurring is independent of the status of testing the isolation
function of the radiation monitor associated with the purge system
valves. This change eliminates the requirement for testing of this
isolation function immediately prior to refueling operations. The
change continues to require the isolation function to be OPERABLE
and continues to ensure that this function is verified within a
reasonable interval prior to irradiated fuel assembly handling
within containment. This provides reasonable assurance the isolation
function of the radiation monitor associated with the purge system
valves remains OPERABLE. Therefore the consequence of an accident
previously evaluated are not significantly increased.
The proposed change does not involve any physical alteration of
plant systems, structures or components, changes in parameters
governing normal plant operation, or methods of operation. The
isolation function of the Reactor Building Purge supply and exhaust
valves is not assumed to be an initiator of any analyzed event. As a
result, the probability of an accident occurring is independent of
the status of testing the isolation function of the Reactor Building
Purge supply and exhaust valves. This change eliminates the
requirement for testing of the isolation function of the Reactor
Building Purge supply and exhaust valves immediately prior to
refueling operations. The change continues to require the isolation
function of the Reactor Building Purge supply and exhaust valves
train to be OPERABLE and continues to ensure that this function is
verified within a reasonable interval prior to irradiated fuel
assembly handling within containment. This continues to provide
reasonable assurance the isolation function of the Reactor Building
Purge supply and exhaust valves remains OPERABLE. Therefore the
consequence of an accident previously evaluated are not
significantly increased.
2. Does the change create the possibility of a new or different
kind of accident from the accidents previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will still require the isolation
function of the radiation monitor associated with the purge system
valves be OPERABLE. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will still require the isolation
function of the Reactor Building Purge supply and exhaust valves be
OPERABLE. Thus, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The isolation function of the radiation monitor associated with
the purge system valves is still required to be OPERABLE. This
change continues to ensure that this function is verified within a
reasonable interval prior to irradiated fuel assembly handling
within containment. Therefore the margin of safety has not been
significantly reduced.
The isolation function of the Reactor Building Purge supply and
exhaust valves is still required to be OPERABLE. This change
continues to ensure that this function is verified within a
reasonable interval prior to irradiated fuel assembly handling
within containment. Therefore the margin of safety has not been
significantly reduced.

E. CTS 3.7.6 and 3.7.7 both require an inoperable voltage sensing
relay to be restored within 72 hours. ITS 3.3.19 Required Action A.1
and ITS 3.3.20 Required Action A.1 would be

[[Page 53948]]

incorporated to require that the inoperable channel be placed in trip
within 72 hours. This change allows operation to continue indefinitely
when the channel is placed in trip and continues to allow 72 hours to
restore an inoperable channel that cannot be placed in trip.
Basis for proposed no significant hazards consideration
determination:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
This change allows indefinite continued operation with one
voltage sensing channel inoperable, provided the inoperable voltage
sensing channel is placed in trip within 72 hours. This action
leaves the system in a one-out-of-two condition for actuation. Thus,
if another channel were to fail, the DGVP (degraded grid voltage
protection) instrumentation can still perform its function. This
change does not significantly increase the probability of occurrence
for initiation of any analyzed event since the function of the DGVP
instrumentation does not change (and therefore any initiation
scenarios are not changed). Also, the change does not change the
assumed response of the equipment in performing its specified
function from that originally considered. Therefore, the changes do
not significantly increase the consequences of an accident.
2. Does the change create the possibility of a new or different
kind of accident from the accidents previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The change ensures proper availability for the required
DGVP function. Thus, this change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety? This change to the DGVP instrumentation requirements does
not involve a change in setpoints and cannot affect any margin of
safety associated with the response to a design basis accident. The
change does not prevent the DGVP instrumentation from performing
their function since the action places the DGVP instrumentation in a
one-out-of-two condition for actuation versus the normal two-out-of-
three logic. Thus, if another channel were to fail, the DGVP
instrumentation could still perform its initiation functions.
Therefore, this change to allow the DGVP initiation functions to
operate indefinitely with one required DGVP instrument channel
inoperable provided the channel is placed in the tripped condition
within 72 hours, is not considered to involve a significant
reduction in the margin of safety.

F. CTS Table 4.1-3 requires that CFT boron concentration be sampled
monthly and after each makeup. ITS SR 3.5.1.4 requires it be sampled
every 31 days and once within 12 hours after each solution increase
greater than or equal to 80 gallons that is not the result of addition
from a borated water source that meets CFT boron concentration
requirements. Therefore, the ITS frequency is less restrictive than
current requirements because sampling will be required once within 12
hours following the volume increase and source requirement. Also, the
source of makeup would be changed from the ``borated water storage
tank'' to ``a source that meets CFT boron concentration requirements.''
Basis for proposed no significant hazards consideration
determination:

1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
A less frequent performance of a Surveillance Requirement does
not result in any hardware changes. The Frequency of performance
also does not significantly increase the probability of occurrence
for initiation of any analyzed event since the function of the
equipment does not change (and therefore any initiation scenarios
are not changed) and the proposed Frequency has been determined to
be adequate to demonstrate the tank inventory is within the required
parameter limits. Further, the Frequency of performance of a
surveillance does not significantly increase the consequences of an
accident because a change in Frequency does not change the assumed
response of the equipment in performing its specified mitigation
functions from that considered with the original Frequency. The core
flood tank boron concentration change resulting from volume addition
from a source of known concentration is a readily calculated
quantity and hence, a sample and analysis is not required to be
assured of adequate boron concentration. Therefore, this change does
not involve a significant increase in the probability or
consequences of any accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from the accidents previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will still ensure proper
surveillances are required for equipment considered in the safety
analysis. Thus, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change continues to provide assurance of acceptable
boron concentration since addition from a source of known
concentration results in a readily identifiable resulting
concentration. Therefore, a change in the Surveillance Frequency
does not involve a significant reduction in the margin of safety.

G. The proposed change would specify actions to be taken for
Borated Water Storage Tank (BWST) level, boron concentration, or
temperature not being within specifications. Proposed ITS 3.5.4
Required Action C.1 would allow 12 hours to reach Mode 3 (i.e., an
additional 6 hours over what is currently allowed by CTS 3.2.2) under
such conditions.
Basis for proposed no significant hazards consideration
determination:

H. CTS 3.3.4.b requires the BWST minimum boron concentration to be
within the limit specified in the core operating limits report at a
minimum temperature of 50 deg.F and would be changed to 45 deg.F in
ITS SR 3.5.4.1. BWST maximum temperature would be changed from 100
deg.F to 115 deg.F.
Basis for proposed no significant hazards consideration
determination:

[[Page 53949]]

assumed to be the initiators of any analyzed events. As a result,
the probability of an analyzed event is independent of these values.
The proposed change from allowable values based on the uncertainties
associated with the instrument channel to an analytical limit for
the parameter being measured continues to ensure that the limits on
volume and pressure are maintained within analyzed values.
Therefore, this change does not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from the accidents previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The analytical limits of variables established by the
safety analysis have not been changed. Thus, this change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does this change involve a significant reduction in a margin
of safety?
Changing the limits from an allowable value based on the
uncertainties associated with the instrument channel to an
analytical limit for the parameter being measured does not involve a
significant reduction in the margin of safety since the actual
pressure and volume assumed in the safety analyses are not changed.

Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for each of the proposed changes. 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.
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW, Washington, DC.
NRC Project Director: Herbert N. Berkow.

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

Date of amendment request: July 31, 1998.
Description of amendment request: The proposed amendment would
clarify requirements for diesel generator start voltage and frequency.
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) Analyzed events are initiated by the failure of certain
plant structures, systems or components. The proposed changes to the
Clinton Power Stations (CPS) Technical Specifications revise the
acceptance criteria for Surveillance Requirements (SRs) pertaining
to the diesel generators (DGs). The DGs are not considered as
initiators of any analyzed event. Thus, these changes do not
increase the probability of any accident previously evaluated.
The consequences of analyzed events involving the diesel
generators are dependent on the successful functioning of the diesel
generator(s) to mitigate such events when a concurrent loss of
offsite power is postulated. The proposed change in the acceptance
criteria for testing of the DGs per the affected SRs accounts for DG
governor performance in response to a fast start. Notwithstanding,
the revised SRs will continue to ensure that minimum frequency and
voltage are attained within the required time, thus satisfying
permissive conditions required for closure of the DG output breaker.
The SRs will also continue to ensure that proper steady-state
voltage and frequency are attained consistent with proper DG
governor and voltage regulator performance. Additionally,
verification that permanently connected loads are energized within
the required time (in response to a loss of offsite power or in
response to a loss of coolant accident (LOCA) concurrent with a loss
of offsite power) will continue to be performed pursuant to SRs not
affected by the proposed changes. Thus, there is no impact on the
capability of the DGs to perform their required safety function.
Based on the above, IP (Illinois Power Co.) has concluded that
the proposed changes will not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
(2) The proposed changes do not involve a physical alteration of
the plant. No new or different equipment is being installed, and no
installed equipment is being operated in a new or different manner.
There is no alteration to the parameters within which the plant is
normally operated or in the set points that initiate protective or
mitigative actions. As a result, no new failure modes are being
introduced.
Additionally, there are no changes in the methods governing
normal plant operation, nor are the methods utilized to respond to
plant transients altered.
Based on the above, IP has concluded that the proposed changes
will not create the possibility of a new or different kind of
accident not previously evaluated.
(3) As noted previously, the proposed changes to the acceptance
criteria for testing of the DGs per the affected SRs accounts for
the characteristics of the DG governor during a fast start, but they
do not impact the effectiveness of such testing to provide assurance
of DG operability. Thus, the proposed changes do not impact expected
DG performance, including the capability for each DG to attain and
maintain required voltage and frequency for accepting and supporting
plant safety loads within the required time, as assumed in the plant
safety analyses.
Margins of safety are established through the design of the
plant structures, systems and components, the parameters within
which the plant is operated, and the establishment of set points for
the actuation of equipment relied upon to respond to an event. With
respect to any margins of safety associated with the diesel
generators, and as noted previously, the proposed changes do not
impact diesel generator performance. That is, the SRs as revised
will continue to ensure that proper voltage and frequency are
attained for closure of the DG output breaker, and for steady-state
conditions consistent with proper DG governor and voltage regulator
performance. In addition, the proposed changes involve no changes to
any setpoints or settings associated with the diesel generators. On
this basis, the proposed changes do not involve any changes to any
assumptions of the plant safety analyses with regard to the function
of the diesel generators. Thus, no margins of safety are impacted by
the proposed changes.
Based on the above, IP has concluded that the proposed change
will not result in a reduction in a margin of safety.

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

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

Date of amendment request: August 17, 1998.
Description of amendment request: The proposed amendment would
reduce the load at which the diesel generators are tested.
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) Analyzed events (or events bounded by analyzed events) are
initiated by the failure of certain plant structures, systems or
components. The scope of the proposed changes is limited only to the
revision of several Surveillance Requirements (SRs) for testing of
the standby emergency diesel generators (DGs). The DGs are not
considered as initiators of any analyzed event. Thus, the proposed
changes do not impact the probability of any accident previously
evaluated.
The consequences of analyzed events are dependent on the
successful functioning of

[[Page 53950]]

credited equipment to mitigate such events. With respect to the
proposed changes, there is no impact on the capability of credited
equipment, i.e., the diesel generators, to perform as required (in
the event of a loss of coolant accident concurrent with a loss of
offsite power). Testing at reduced load levels reduces stress and
wear on the diesel generators, while still ensuring that the DGs are
adequately challenged at operating temperatures to confirm
operability. In addition, reducing the minimum required load levels
reduces time when, or the probability that, the short-term rating of
any diesel generators is exceeded during testing. The resultant
reduction in stress and wear increases DG availability.
Based on the above, IP (Illinois Power Co.) has concluded that
the proposed changes will not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
(2) The proposed changes do not involve a physical alteration of
the plant. No new or different equipment is being installed, and no
installed equipment is being operated in a new or different manner.
There is no alteration to the parameters within which the plant is
normally operated or in the set points that initiate protective or
mitigative actions. As a result, no new failure modes are being
introduced.
Based on the above, IP has concluded that the proposed changes
will not create the possibility of a new or different kind of
accident not previously evaluated.
(3) The revised Surveillance Requirements are consistent with
the recommendations of RG [Regulatory Guide] 1.9, Revision 3.
Testing at reduced load levels reduces stress and wear on the diesel
generators, while still ensuring that the DGs are adequately
challenged at operating temperatures to confirm operability. In
addition, reducing the minimum required load levels reduces time
when, or the probability that, the short-term rating of any diesel
generators is exceeded during testing. The resultant reduction in
stress and wear increases DG availability.
Margins of safety are established through the design of plant
structures, systems and components, the parameters within which the
plant is operated, and the establishment of set points for the
actuation of equipment relied upon to respond to an event. With
respect to any margins of safety associated with the diesel
generators, the proposed changes do not impact diesel generator
performance, and involve no changes to any setpoints or settings
associated with the diesel generators, nor do the proposed changes
involve any changes to any assumptions of the plant safety analyses
with regard to the function of the diesel generators. Thus, no
margins of safety are impacted by the proposed changes.
Based on the above, IP has concluded that the proposed changes
will not result in a reduction in a margin of safety.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan

Date of amendment request: August 28, 1998.
Description of amendment request: The proposed amendment would
grant relief from the steam generator inspection surveillance
requirement described in technical specification No. 4.4.5.3. The
relief would allow the inspection to be deferred from April 8, 1999,
until the next refueling outage for Donald C. Cook Nuclear Plant , Unit
1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

In accordance with CFR 50.92, the proposed amendment will not
involve a significant hazards consideration if the changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed;
2. Create the possibility of a new or different kind of accident
from any accident previously analyzed or evaluated; or
3. Involve a significant reduction in a margin of safety.

Criterion 1

The last unit 1 surveillance was completed in the spring of 1997
and was the most thorough evaluation of the steam generators to
date. Both standard and enhanced eddy current inspection techniques
were employed to inspect the steam generator tubing. Additionally, a
series of in situ pressure tests were performed to verify tubing
integrity. Tube repairs consisting of hot leg tube end re-rolling
and plugging were performed. Pre- and post-tube bundle pressure
tests were conducted to verify the integrity of the repairs. A tube
pull was also conducted to verify continued comformance with generic
letter 95-05 requirements. The tube pull data did not identify any
unexpected conditions or areas of concern. During the 1997
inspection, select secondary side visual and eddy current
inspections were also performed to provide assurance of continued
secondary side internals integrity.
Following the inspection, a condition monitoring and operational
assessment, using data gathered during the steam generator
inspections and tests, was made to determine whether steam generator
leakage and structural integrity could be maintained throughout the
upcoming cycle (cycle 16).
The unit was subsequently restarted and the steam generators
operated without incident when a unit shutdown occurred in September
of 1997.
Throughout the cycle 16 operating period, a relatively low
reactor coolant temperature was maintained. By maintaining a T-hot
temperature of approximately 586 deg.F during the operating period,
corrosion impact on the steam generator tubes was minimized.
Throughout the operating period, steam generator primary-to-
secondary leakrate monitoring was performed to assure conformance
with T/S requirements. Historically, Unit 1 has not experienced a
forced shutdown because of leakrate concerns.
During the shutdown period, the steam generators have been
maintained under lay-up conditions, which comply with or exceed the
industry standard practice. These practices are designed to mitigate
the corrosive environment within the steam generators.
The previous cycle 16 integrity assessment has been re-visited
to provide reasonable assurance conclusions made remain valid given
the extended shutdown period. This re-assessment considered the
initial cycle runtime, the shutdown period and subsequent operation
through the end of the current fuel cycle. These results confirm the
findings of the initial evaluation (i.e., that adequate steam
generator integrity will be maintained throughout the current
cycle).
The proposed change will not affect the scope, methodology,
acceptance limit, or corrective measures of the existing steam
generator examination program. As adequate integrity will be
maintained, the probability and consequences of an accident
previously analyzed due to leaking or degraded tubes is not
increased by the proposed change.

Criterion 2

We have determined that this extension will not result in a
change in plant configuration or operation. Plant systems and
components will not be operated in a different manner as a result of
this change. No plant modifications or changes in methods of
operation will result from this change. Therefore, the extension
will not create the possibility of a new or different kind of
accident from what has been previously evaluated or analyzed.

Criterion 3

We have determined that the proposed extension request will not
involve a significant reduction in a margin of safety. Re-assessment
of the cycle 16 steam generator operational assessment report, which
indicates structural and leakage integrity will be maintained
throughout the cycle, has shown that the shutdown period will not
adversely impact overall steam generator integrity.
This assessment concluded that when the reactor is shut down and
the reactor coolant system is at a reduced temperature, the steam

[[Page 53951]]

generators are not subject to conditions that lead to tube
degradation. The actual number of days that the steam generators
will be subjected to an environment conducive to tube degradation is
not being increased under this request. Therefore, this request is
judged not to 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: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ronald R. Bellamy (Acting).

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

Date of amendment request: August 12, 1998.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) by updating the list of
documents specified in TS 6.9.1.8b that describe the analytical methods
used to determine the core operating limits. The changes can be
categorized as: (1) The analysis methodology is unchanged, but the
reference has been clarified by identifying the specific revision,
supplements, and dates for the revision; (2) the analysis methodology
is unchanged and the reference is being added for completeness and; (3)
the analysis methodology is being changed. 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 change in reference 4 of Technical Specification
Section 6.9.1.8b revises the steam line break analysis methodology
to be applied to Millstone Unit No. 2 and clarifies the references
to the Siemens topical reports. The other changes are clarifications
or additions for completeness and do not represent a change in the
approved methodology for Millstone Unit No. 2. The change in
methodology is associated with the interference between XTGPWR, the
neutronics code, and XCOBRA-IIIC, the thermal hydraulics code. It
has no impact on plant equipment operation. Since the change only
affects the analysis of the events, it cannot affect the likelihood
or consequences of these events. Therefore, this change will not
significantly increase the probability or consequences of an
accident previously evaluated.
The sentence on page 6-19, starting with ``The acceptable
Millstone 2 * * *.'' and ending with ``* * * dated October, 1988,''
references the document ANF-88-126, ``Millstone Unit 2 Cycle 10
Safety Analysis Report,'' which has been outdated because of the
above mentioned changes in the methodology. The removal of this
sentence is necessary to be consistent with methodology changes.
Therefore, this change 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 accident previously evaluated.
The proposed change in reference 4 of Technical Specification
Section 6.9.1.8b revises the steam line break analysis methodology
to be applied to Millstone Unit No. 2 and clarifies the references
to the Siemens topical reports. The other changes are clarifications
or additions for completeness and do not represent a change in the
approved methodology for Millstone Unit No. 2. The proposed change
in reference 4 of Technical Specification Section 6.9.1.8b will not
alter the plant configuration (no new or different type of equipment
will be installed) or require any new or unusual operator actions.
It does not alter the way any structure, system, or component
functions and does not alter the manner in which the plant is
operated.
The sentence on page 6-19, starting with ``The acceptable
Millstone 2 * * *.'' and ending with ``* * * dated October, 1988,''
references an outdated document. The removal of this sentence is
necessary to be consistent with methodology changes. The change does
not alter the way any structure, system, or component functions and
does not alter the manner in which the plant is operated.
The changes do not introduce any new failure modes. Therefore,
the proposed changes will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed change in reference 4 of Technical Specification
Section 6.9.1.8b revises the steam line break analysis methodology
to be applied to Millstone Unit No. 2 and clarifies the references
to the Siemens topical reports. The other changes are clarifications
or additions for completeness and do not represent a change in the
approved methodology for Millstone Unit No. 2. The change in steam
line break methodology is associated with the interface between
XTGPWR, the neutronics code, and XCOBRA-IIIC, the thermal hydraulics
code. The change will result in a better correlation between the two
computer codes, which is the intent of the iteration process. This
will result in more accurate results while still maintaining a
conservative modeling of the event. The most significant impact is
on the low RCS [reactor coolant system] flow cases associated with
loss of offsite power. These cases are not limiting when compared to
the offsite power available cases. The improved references will
clearly identify the approved Siemens Topical Reports applicable to
Millstone Unit No. 2 and will ensure that methodology changes will
be identified and submitted to the NRC for approval as required. The
sentence on page 6-19, starting with ``The acceptable Millstone 2 *
* *.'' and ending with ``* * * dated October, 1988,'' references an
outdated document. The removal of this sentence is necessary to be
consistent with methodology changes.
Therefore, the proposed changes will not result in a significant
reduction in the margin of safety as defined in the Bases for
Technical Specifications covered in this License Amendment Request.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff 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, PO Box 270, Hartford,
Connecticut.
NRC Project Director: William M. Dean.

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

Date of amendment request: August 15, 1996, as supplemented March
19, 1998.
Description of amendment request: The proposed amendment revises
the Technical Specifications so that either 8 or 12 hour shifts will be
considered ``normal'' and 40 hours will be considered a ``nominal''
week, changes the wording for surveillances required ``once per shift''
to ``once per 12 hours,'' clarifies the ``once per hour'' wording
related to fire watch patrols, and makes a number of other
typographical corrections and clarifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards

[[Page 53952]]

consideration, which is presented below:

(1) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(With respect to shift definition and editorial changes:) This
change does not affect the physical configuration of the plant or
how it is operated, as such, it is not the initiator of any plant
event. Working a ``normal'' 12-hour shift is no different from
working a ``normal'' 8-hour shift with 4-hours of overtime which has
been an accepted and approved practice for years. Therefore, the
proposed changes will not result in any increase in the probability
of an accident occurring. The intent is still that operators will
not work excessive overtime either on a daily, or weekly basis.
The typographical errors, clarifications and title changes do
not involve technical issues and as such do not involve safety
issues, and therefore do not effect [sic] the chances or
consequences of an accident.
(With respect to surveillance and fire watch patrol interval:)
This change does not affect the physical configuration of the plant
or how it is operated. As such, it is not the initiator of any plant
event. This change clarifies the intervals in which Sensor Checks,
Surveillances, and fire watch patrols must be completed. As
described above [in the supplement], the 12-hour interval has been
determined acceptable for the specified Sensor Checks and
Surveillances based on Monticello and industry experience which
demonstrates instrumentation and channel failures are rare. This
change conforms the Monticello TS (Technical Specifications) to
NUREG-1433 and clarifies the intervals in which checks must be
completed.
Completing fire watch patrols on a one hour +25% interval will
require patrols on an hourly basis, while providing flexibility to
complete the patrols within a 15 minute window. In addition to the
Technical Specification required fire watches, additional
individuals are often in the plant proper, so the required hourly
fire watch patrols are only part of the entire program for fire
detection.
Therefore, the proposed changes will not result in a significant
increase in the probability of an accident occurring.
(2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
(With respect to shift definition and editorial changes:) This
change does not affect the physical configuration of the plant or
how it is operated. Therefore, revising the length of a ``normal''
shift or correcting minor errors does not create the possibility of
a new or different kind of accident from any previously evaluated.
As such, it is not the initiator of any plant event.
(With respect to surveillance and fire watch patrol interval:)
Revising the wording to ``once per 12 hours'' or ``once per hour
(+25%)'' does not create the possibility of a new or different kind
of accident from any previously evaluated. No new or different
surveillance activities are proposed, nor are any being deleted. As
such, it is not the initiator of any plant event.
(3) The proposed amendment will not involve a significant
reduction in the margin of safety.
(With respect to shift definition and editorial changes:) This
change does not affect the physical configuration of the plant or
how it is operated. The level of expertise on shift will not be
diminished or changed as a result of this change. Therefore, this
change will not reduce the margin of safety.
(With respect to surveillance and fire watch patrol interval:)
This change does not affect the physical configuration of the plant
or how it is operated. The level of expertise on shift will not be
diminished or changed, nor will it reduce the functionality of plant
equipment. This change requires Sensor Checks, surveillances, and
fire watch patrols be completed within industry guidelines.
The 12 hour interval has been determined acceptable based on
industry experience which demonstrates channel failure is rare. The
one hour interval for fire watch patrols has also been an accepted
industry standard. In addition to the Technical Specification
required fire watches, additional individuals are often in the plant
proper, so the required hourly fire watch patrols are only part of
the entire program for fire detection. The proposed change simply
defines the acceptable interval during which the task must be
performed. Therefore, this change does not constitute 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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.

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

Date of amendment request: January 14, 1998, as supplemented by
letter dated May 19, 1998.
Description of amendment request: The proposed amendment would
approve a modification to the Diablo Canyon Power Plant, Unit Nos. 1
and 2, 230 kV transmission system. The modifications include
installation of new 230/12kV startup transformers with automatic load
tap changers, along with installation of shunt capacitor banks. The
transformers will assure that voltage on the plant 12 kV and 4 kV buses
is maintained within limits, while the capacitor banks assure adequate
VAR support.
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 replacement of the startup transformers (SUTs) with new
transformers equipped with load tap changers (LTCs) for voltage
control does not alter the original configuration of the electrical
distribution system and hence, will not increase the probability of
occurrence of an accident previously evaluated.
The replacement of the SUTs with new transformers equipped with
LTCs will enhance the capability of the 12 kV and 4 kV electrical
distribution systems to maintain sufficient voltage for successful
transfer of the plant auxiliary loads to the startup source
following a unit trip. This change eliminates the potential for
``double sequencing'' (starting loads from the 230 kV system,
subsequent voltage degradation causes load shedding and restarting
from the diesel generators) of the 4 kV vital loads during an
accident by providing adequate voltage to the 4 kV vital buses from
the 230 kV source. The maintenance of adequate voltage at the 4 kV
vital buses prevents the second level undervoltage relay (SLUR)
action. The LTC will automatically maintain adequate voltage at the
terminals of the vital equipment under design basis accident
conditions. Therefore, engineered safety feature equipment will
function as previously evaluated.
The manual operation of the Unit 2 LTC while in a standby mode
will not increase the probability of an accident since normally none
of the plant loads are energized from the 230 kV system. Plant loads
are only powered from the 230 kV system during short periods of unit
startup and shutdown. Loss of the 230 kV system while the operating
plant loads are fed from the 25/500 kV system cannot initiate an
accident since the system is not connected to plant equipment if the
loads are supplied by the 25/500 kV system. Therefore, the proposed
modifications will not increase the probability of an accident
previously evaluated. The manual operation of the Unit 2 LTC assures
adequate voltage is supplied to Unit 2 safety equipment in the event
of an accident. Therefore, the proposed modification will not
increase the consequences of an accident.
The installation of the shunt capacitors at the Diablo Canyon
Power Plant switchyard and Mesa Substation to replace the VAR
support from Morro Bay Power Plant (MBPP), assuming no MBPP
generation, does not alter the capability or availability of the
offsite

[[Page 53953]]

power source. Since shunt capacitors are considered more reliable
than generators, it adds to the reliability of the 230 kV system and
will not increase the probability of an accident previously
evaluated.
Even if 230 kV voltage were lost or became degraded, the first
or second level undervoltage relays will initiate transfer to the
diesel generators should there be a loss or degraded 230 kV system
while feeding the vital loads from the 230 kV system. This scenario
is evaluated in Final Safety Analysis Report (FSAR) Update Section
15.2.9.1 ``Loss of Offsite Power to the Station Auxiliaries.''
Therefore, the changes will not increase the consequences of an
accident previously evaluated since the safety-related loads will
function as required.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The change does not result in a change in operation,
maintenance, physical change, or procedural change that could create
the possibility of an accident that is of a new or different type
than previously evaluated.
The replacement SUTs and the installation of the shunt
capacitors to replace MBPP serves the same function as the original
design and do not create the possibility of a new or different type
of accident. Should there be a loss of offsite power, the onsite
power source (diesel generators) will provide power to the loads.
The FSAR already includes an evaluation for station blackout if
there is a total loss of both onsite and offsite power.
Therefore, the 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.
The replacement transformers and the installation of the shunt
capacitors will not cause a reduction in the margin of safety as
defined in the basis for any Technical Specification (TS). The
minimum voltage required for safe shutdown is defined in TS Table
3.3.4, Functional Unit 7.b, ``Second Level Undervoltage Relay (SLUR)
setting.'' By replacing the existing SUTs with automatic LTC
transformers, the vital 4 kV bus voltage will be automatically
maintained at a sufficiently higher value during normal operation
such that during an accident, the minimum 4 kV vital bus voltages
after the bus transfer will be adequate to prevent SLUR actuation.
The installation of the shunt capacitors will assure adequate VAR
support that was previously provided by operation of the MBPP in the
Los Padres Region of PG&E's service territory for present peak load
and future peak load growth under worse case line outage conditions.
During the interim period between January and February 1998,
when manual control of the Unit 2 SUT LTC will be utilized to
maintain adequate voltage at the 12 kV and 4 kV buses, the margin of
safety is not reduced since the adjustment of the LTC will assure
stable voltage for the vital buses.
Therefore, there is no reduction in a margin of safety as
defined in the basis for any TS.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room Location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas &
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.

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

Date of amendment request: March 18, 1998.
Description of amendment request: The proposed amendment would
approve a change in the way passive failures in the auxiliary saltwater
(ASW) and component cooling water (CCW) systems are mitigated during
the long-term recovery period following a loss-of-coolant accident
(LOCA). Specifically, plant procedures would no longer require ASW and
CCW system train separation after the transfer to hot leg recirculation
following a LOCA.
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 changes revise the way passive failures are mitigated in the
auxiliary saltwater (ASW) and component cooling water (CCW) systems.
Specifically, plant procedures would no longer require ASW and CCW
train separation after transfer to hot leg recirculation following a
loss-of-coolant accident. The decision to separate trains would be
made by the Technical Support Center (TSC) after evaluation of plant
conditions. Operation of the ASW and CCW systems during this period
is required to mitigate the accident, therefore, the change in plant
operation would not affect the probability of that accident
occurring.
The change ensures the ASW and CCW systems will be able to
mitigate an active or passive failure without the loss of safety
function during the long-term (beginning 24 hours after the
accident) period of recovery following an accident. Since the ASW
and CCW systems will continue to perform their safety function,
overall system performance is not affected, assumptions previously
made in evaluating the consequences of the accident are not altered,
and the consequences of the accident are not increased as a result
of the change in plant operation.
Therefore, the changes will not increase 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 ASW and CCW systems function to mitigate the consequences of
an accident. The change in operation ensures these system will be
able to mitigate an active or passive failure without loss of safety
function during the long-term (beginning 24 hours after the
accident) period of recovery following an accident. Operation of the
ASW and CCW systems in accordance with plant procedures, and the
guidance on train separation provided to the TSC, ensure the design
basis requirements for the ASW and CCW systems will continue to be
met. Therefore, the ability of the ASW and CCW systems to mitigate
the accident is not degraded. Required operator actions are similar
to other operator actions specified in the FSAR that are considered
acceptable by the NRC.
Therefore, the 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.
The change ensures the ASW and CCW systems will be able to
mitigate an active or passive failure without loss of safety
function during the long-term (beginning 24 hours after the
accident) period of recovery following an accident. Since the ASW
and CCW systems will continue to perform their safety function,
there is no impact on any acceptance limits for ASW and CCW system
operation assumed in the safety analysis, or on any Technical
Specification (TS).
Therefore, the change does not involve a significant reduction in a
margin of safety as defined in the basis for any TS.

[[Page 53954]]

NRC Project Director: William H. Bateman.

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

Date of amendment request: August 10, 1998,
Description of amendment request: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 to revise TS 3/4.3.2, Table 3.3-5,
``Engineered Safety Features Response Times,'' to add the response
times for closure of the main feedwater regulating valves (MFRVs) and
MFRV bypass valves, and trip of the main feedwater pumps (MFWPs). The
change would also revise TS 3/4.7.1.7 to add a limiting condition for
operation (LCO), actions, and surveillance requirements for the MFWP
turbine stop valves, and would revise the actions and surveillance
requirements for the MFRVs, MFRV bypass valves, and main feedwater
isolation valves (MFIVs) to be consistent with the NUREG-1431
requirements. Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the Technical Specifications (TS) to add
response time requirements for the main feedwater regulating valve
(MFRV) and associated bypass valves and the main feedwater pump
(MFWP) trip provide more restrictive TS requirements that are
consistent with current plant practice. They do not change the
function or operation of any plant equipment or affect the response
of that equipment if it is called upon to operate. These more
restrictive requirements are imposed to ensure the affected
components are maintained consistent with the safety analyses and
licensing bases.
The proposed changes to: (1) Revise the actions to apply to one
or more main feedwater isolation valves (MFIVs), and MFRVs and
associated bypass valves, (2) extend the action completion time from
4 hours to 72 hours, (3) provide actions when two valves affecting
the feedwater isolation capability for a flow path are inoperable,
(4) add actions for an inoperable MFWP turbine stop valve, and (5)
allow separate action entry for each inoperable valve unless the
feedwater isolation capability for a flow path is affected, do not
change the function or operation of any plant equipment or affect
the response of that equipment if it is called on to operate. The
actions account for the redundancy provided by the remaining valves
and the MFWP trip, and the low probability of an event occurring
during this time period that would require isolation of the main
feedwater flow path. A probabilistic risk assessment, performed to
assess the increase in annual core damage frequency (CDF) associated
with the increase in allowable outage time, determined the increase
in annual CDF to be approximately 1.5 percent. That increase in
annual CDF is considered non-risk significant per the Electric Power
Research Institute ``PSA Application Guide.''
The addition of the limiting condition for operation, actions,
and surveillance requirements for the MFWP turbine stop valves, and
the addition of the surveillance requirement for the MFIVs, MFRVs,
and MFRV bypass valves are more restrictive requirements that ensure
these components are operable and capable of performing their safety
function. They do not change the function or operation of any plant
equipment or affect the response of that equipment if it is called
on to operate. The proposed surveillance intervals are supported by
the operating, maintenance, and surveillance histories of the
valves.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not necessitate a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in the parameters governing normal plant operation. The
changes imposed are consistent with the assumptions made in the
accident analyses and licensing basis.
Therefore, the 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.
The proposed changes to the TS impose requirements consistent
with the assumptions in the safety analyses and current licensing
bases, and reflect current plant practice. They do not alter the
margins of safety established in previous accident and transient
analysis.
Therefore, none of the proposed changes involves a significant
reduction in a margin of safety.

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

Date of amendment request: September 8, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Appendix C, ``Additional
Conditions,'' to authorize the use of non-Class 1E single cell battery
chargers, with proper electrical isolation, for charging connected
cells in OPERABLE Class 1E batteries. The single cell chargers would be
used to restore individual cell float voltage to the normal limit
specified in TS Table 4.8.2.1-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change permits the use of an industry accepted
method to restore a battery cell to its design basis from an
OPERABLE but degraded condition or to prevent a cell from becoming
degraded. IEEE Std 450-1995, ``IEEE Recommended Practice for
Maintenance, Testing, and Replacement of Vented Lead Storage
Batteries for Stationary Applications,'' states that single cell
charging is an acceptable method of correcting low cell voltage or
low specific gravity conditions for a single cell or for a small
number of cells.
At least two class 1E fuses in series will be used on both the
positive and negative leads between the battery and the charger to
protect the battery if a fault should develop in the charger. The
battery charger design includes diodes, a power transformer and
control circuitry to prevent draining the connected cells in the
event of a short circuit in the 120 Volt ac source or a loss of
charger input or output voltage. Charger output is controlled
automatically to prevent overcharging the connected cells.
In the event of a controller failure resulting in charger
overvoltage, procedural controls governing the use of the charger
ensure the condition is detected and corrected before failure of a
connected cell occurs. While the single cell charger is connected,
procedures will require periodic checks to verify proper charger
operation and to measure electrolyte level, temperature and specific
gravity for the cells being charged. Monitoring will be

[[Page 53955]]

performed at least once every eight hours, a frequency sufficient to
ensure compliance with the ACTION requirements of Technical
Specification 3.8.2.1.
An insulating material will be used to minimize the possibility
of shorting leads or clips at the battery. Administrative controls
governing the use and storage of transient loads are sufficient to
ensure the use of single cell battery chargers does not create a
potential missile hazard to safety related systems, structures and
components.
The Class 1E dc system is not an accident initiator. It supports
the operation of safety related equipment required for the safe
shutdown of the plant and for the mitigation of accident conditions.
Therefore, the proposed change does not increase the probability of
an accident previously evaluated.
The station's dc systems will be operable to mitigate the
consequences of an accident previously evaluated. Single cell
charging would be limited to one OPERABLE class 1E battery bank at a
time. Therefore, failure of a class 1E battery as a result of single
cell charging would be limited to a single channel and would not
reduce the number of OPERABLE dc sources below that required to
safely shutdown the plant. Administrative controls would also
prohibit the use of single cell charging for an OPERABLE class 1E
battery if less than the minimum number of class 1E batteries
required by Technical Specifications are OPERABLE.
The proposed change does not cause the capability of the class
1E dc system to be degraded below the level assumed for any accident
described in the (safety analysis report) SAR. It would enhance the
availability of safety related equipment required for the safe
shutdown of the plant and for the mitigation of accident conditions.
Therefore the radiological consequences of an accident will remain
inside the design basis while single cell charging is performed on
an OPERABLE battery.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The potential to adversely affect the Class 1E batteries is
minimized by the use of Class 1E fuses and by appropriate
administrative controls. Failure modes associated with the proposed
change are bounded by the loss of a Class 1E battery bank which was
previously evaluated. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
(3) The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change permits the use of non-Class 1E single cell
battery chargers, with proper electrical isolation, for charging
connected cells in OPERABLE class 1E batteries. This would allow
parameters for an individual cell or for a small number of cells to
be restored to the normal values specified in Technical
Specifications without affecting the remainder of the cells in the
battery. Increased cell monitoring after single cell charging,
together with PSE&G's corrective action program which requires
degraded and non-conforming conditions to be documented and
evaluated, provides assurance that the use of single cell charging
will not cause long-term cell degradation to go undetected. Since
all battery cells are required to be maintained within the allowable
values specified in Technical Specifications, and since the use of
the single cell charger will not adversely affect battery capacity
or capability, the proposed change does not involve a significant
reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: Robert A. Capra.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1

Fairfield County, South Carolina.
Date of amendment request: July 1, 1998.
Description of amendment request: The proposed amendment would
revise Virgil C. Summer Nuclear Station (VCSNS) Technical
Specifications (TS) Surveillance Requirement 4.7.7.e to remove the
``during shutdown'' condition from the specified test interval.
Removing the ``during shutdown'' wording from the TS would allow VCSNS
to perform on-line snubber testing, and would make the up to 25 percent
allowable interval extension in Surveillance Requirement 4.0.2 apply to
the specified snubber surveillance interval. The proposed amendment
would also make administrative changes to Surveillance Requirement
4.7.7.g and BASES 3/4.2.2 and 3/4.2.3 to correct typographical errors.
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 probability or consequences of an accident previously
evaluated is not significantly increased.
The proposed change will not affect system operation or
performance, nor do they affect any Engineered Safety Features
actuation setpoints or accident mitigation capabilities. NUREG/CR-
6027 supports the determination that piping failure due to a snubber
single failure is considered low. Therefore, the proposed changes
will not significantly increase the consequences of an accident or
malfunction of equipment important to safety previously evaluated in
the FSAR.
2. The possibility of an accident or a malfunction of a
different type than any previously evaluated is not created.
The changes to the situational testing requirements will not
affect the method of operation of any system to which a snubber is
attached. The proposed changes only address the plant mode at which
a surveillance activity may be performed. No new or different
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures will be introduced as a result of these
changes. Therefore, the possibility of a new or different kind of
accident other than those already evaluated will not be created by
this change.
3. The margin of safety has not been significantly reduced.
This proposed change will not have an impact on the overall
reliability of the snubber population. This is due, in part, to the
fact that the snubber test plans are self correcting. As functional
test failures are identified, additional snubbers are required to be
tested. Thus, the reliability of the snubber population is
maintained. The proposed change does not alter the intent or method
by which the surveillances are conducted, does not involve any
physical changes to the plant, does not alter the way any structure,
system, or component functions, and does not modify the manner in
which the plant is operated. Therefore the proposed change will not
degrade the ability of the snubbers to perform their safety function
or significantly decrease 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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Acting Project Director: P. T. Kuo.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia

Date of amendment request: June 26, 1998, as supplemented by letter
dated September 18, 1998.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) as follows: (1) The
applicability of

[[Page 53956]]

Limiting Condition for Operation (LCO) 3.3.6 would be revised to refer
to TS Tables 3.3.6-1 and 3.3.6-1; the TS Tables would be revised to add
a column entitled ``APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS.''
Then, the applicable modes for Manual Initiation, Automatic Actuation
Logic and Actuation Relays, and Safety Injection functions would be
revised to include only Modes 1, 2, 3, and 4. Consistent with this
proposed change, LCO 3.3.6, Condition C and Required Action C.2 would
be revised to reflect that system level manual initiation and automatic
actuation would not be required during core alterations and/or during
movement of irradiated fuel assemblies within containment. Appropriate
Bases changes are included to reflect the proposed changes; (2) LCO
3.9.4 would be revised to allow the equipment hatch and the emergency
air locks to be open during core alterations and/or during movement of
irradiated fuel assemblies within containment. In addition, the LCO
statement would be revised to reflect that containment ventilation
isolation (CVI) would be accomplished by manually closing the
individual CVI valves as opposed to a system level manual or automatic
initiation, consistent with the proposed changes to LCO 3.3.6. The
surveillance requirements (SRs) would be revised to reflect the
proposed change to the CVI and to reflect that the equipment hatch
would be allowed to be open. Appropriate Bases changes are included to
reflect the proposed changes; (3) LCO 3.7.6a, ``Condensate Storage Tank
(CST)--(Non-redundant CSTs),'' would be deleted. This LCO was created
to address a design condition that rendered the CSTs nonredundant. A
note was added stating that this LCO was only applicable to the unit(s)
that have not completed design modifications required for redundant
CSTs and that the LCO would no longer be required when both units
completed the design modifications. These design modifications have
been completed; therefore, LCO 3.7.6a is no longer applicable, and LCO
3.7.6, ``Condensate Storage Tank (CST)--(Redundant CSTs),'' would be
revised to delete the words ``(Redundant CSTs)'' from the title.
Appropriate Bases changes are included to reflect the proposed changes.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes would revise the VEGP [Vogtle Electric
Generating Plant] Unit I and Unit 2 TS by removing requirements for
automatic and system level manual containment ventilation isolation,
and allow the emergency air lock and the equipment hatch to be open
during core alterations and movement of Irradiated fuel assemblies
inside containment. The containment penetrations affected by the
proposed changes are not initiators for any accident previously
evaluated. Allowing these penetrations to be open under the
conditions specified will not affect the probability of any accident
previously evaluated.
The existing VEGP TS allow the personnel air look doors to be
open during core alterations and movement of irradiated fuel
assemblies inside containment. The radiological consequences of a
fuel handling accident inside containment have been determined to be
below the Standard Review Plan (SRP) section 15.7.4 criteria and
General Design Criteria (GDC) 19 criteria with the personnel air
lock doors open. The proposed changes will not alter these
previously determined consequences. The existing dose analysis
bounds the proposed changes. Therefore, the proposed changes will
not increase the consequences of any accident previously evaluated.
The proposed deletion of LCO 3.7.6a is an administrative change
only. The requirements of LCO 3.7.6a applied only during the time
that the condensate storage tanks (CSTs) were not redundant. Due to
the implementation of design changes which make the CSTs redundant
for each unit, the requirements of LCO 3.7.6a are no longer
applicable. The CSTs (redundant or not) are not initiators for any
accident previously evaluated. Now that the CSTs are redundant, the
requirements of LCO 3.7.6a are no longer necessary to ensure the
capability of the auxiliary feedwater system to perform its safety
function. Therefore, the proposed deletion of LCO 3.7.6a will not
affect the probability or consequences of any accident previously
evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed change does not create any new failure modes
for any system or component, nor does it adversely affect plant
operation. The previously determined radiological consequences of a
fuel handling accident inside containment with the personnel air
lock doors open remain bounding for operation under the proposed
changes. No new single failure scenarios are created, and the
proposed changes do not introduce any new challenges to components
and systems that could result in a new or different kind of accident
from any previously evaluated.
The proposed deletion of LCO 3.7.6a is an administrative change
only. The requirements of LCO 3.7.6a applied only during the time
that the condensate storage tanks (CSTs) were not redundant. Due to
the implementation of design changes which make the CSTs redundant
for each unit, the requirements of LCO 3.7.6a are no longer
applicable. Now that the CSTs are redundant, the requirements of LCO
3.7.6a are no longer necessary to ensure the capability of the
auxiliary feedwater system to perform its safety function. No new
single failure scenarios are created, and the proposed changes do
not introduce any new challenges to components and systems that
could result in a new or different kind of accident from any
previously evaluated. Therefore, the proposed deletion of LCO 3.7.6a
will not create a new or different kind of accident from any
accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The margin of safety for fission product release is 300 rem
thyroid and 25 rem whole body as defined by 10 CFR (Part) 100. The
previously determined radiological dose consequences for a fuel
handling accident inside containment with the personnel air lock
doors open remain bounding for operation under the proposed changes.
These previously determined dose consequences were determined to be
well within the limits of 10 CFR (Part) 100 by virtue of the fact
that they meet SRP Section 15.7.4 and GDC 19 acceptance criteria.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The proposed deletion of LCO 3.7.6a is an administrative change
only. The requirements of LCO 3.7.6a applied only during the time
that the condensate storage tanks (CSTs) were not redundant. Due to
the implementation of design changes which make the CSTs redundant
for each unit, the requirements of LCO 3.7.6a are no longer
applicable. Now that the CSTs are redundant, the requirements of LCO
3.7.6a are no longer necessary to ensure the capability of the
auxiliary feedwater system to perform its safety function.
Therefore, LCO 3.7.6a is not necessary to maintain margin of safety
and the proposed change will not involve a reduction in a margin of
safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Project Director: Herbert N. Berkow.

[[Page 53957]]

Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric
Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: July 13, 1998
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 1.1 Definitions for
``Engineered Safety Feature (ESF) Response Time'' and ``Reactor Trip
System (RTS) Response Time'' to provide for verification of response
time for selected components provided that the components and the
methodology for verification have been previously reviewed and approved
by the NRC. Changes to the TS Bases have also been proposed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) The proposed license amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
This change to the Technical Specifications does not result in a
condition where the design, material, and construction standards
that were applicable prior to the change are altered. The same RTS
[reactor trip system] and ESFAS [engineered safety features
actuation system] instrumentation is being used; the time response
allocations/modeling assumptions in the Chapter 15 analyses are
still the same; only the method of verifying time response is
changed. The proposed change will not modify any system interface
and could not increase the likelihood of an accident since these
events are independent of this change. The proposed activity will
not change, degrade or prevent actions or alter any assumptions
previously made in evaluating the radiological consequences of an
accident described in the SAR [safety analysis report]. Therefore,
the proposed amendment does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
(2) The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
This change does not alter the performance of the pressure and
differential pressure transmitters and switches, Process Protection
racks, Nuclear Instrumentation, and Logic Systems used in the plant
protection systems. Applicable sensors, Process Protection racks,
Nuclear Instrumentation, and Logic Systems will still have response
time verified by test before placing the equipment into operational
service and after any maintenance that could affect the response
time. Changing the method of periodically verifying instrument
response times for certain equipment (assuring equipment
operability) from time response testing to calibration and channel
checks will not create any new accident initiators or scenarios.
Periodic surveillance of these instruments will detect significant
degradation in the equipment response time characteristics.
Implementation of the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(3) The proposed license amendment does not involve a
significant reduction in margin of safety.
This change does not affect the total system response time
assumed in the safety analysis. The periodic system response time
verification method for selected pressure and differential pressure
sensors and for Process Protection racks, Nuclear Instrumentation,
and Logic Systems is modified to allow use of actual test data or
engineering data. The method of verification still provides
assurance that the total system response time is within that assumed
in the safety analysis, since calibration tests will detect any
degradation which might significantly affect equipment response
time. Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Project Director: Herbert N. Berkow.

Date of amendment request: September 3, 1998.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) to: (1) Support the
replacement of the Nuclear Instrumentation System Source Range and
Intermediate Range Channels and Post-Accident Neutron Flux Monitoring
System; and (2) delete the requirement for performing response time
testing of the source range channels and power range detector plateau
voltage determinations.
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 power range low trip, the intermediate range trip, and
the source range trip are designed to provide protection against
power excursions during reactor startup or low-power operation. The
source and intermediate range trips provide redundant protection
during reactor startup or low-power operation. The changes to the
source range and intermediate range instrumentation and setpoints,
as well as the deletion of source range response time testing, do
not affect any safety analysis conclusions because the source range
and intermediate range trips are not explicitly credited in any
design basis accident. Only the power range low trip setpoint is
assumed to actuate to mitigate the uncontrolled rod cluster control
assembly withdrawal accident. The high flux at shutdown alarm
function during a boron dilution event will continue to be provided
by the new source range detector system. No changes have been made
to the setpoint assumed in the safety analyses. The new detector
system is qualified in compliance with Regulatory Guide 1.97 and
will also be used to provide post-accident monitoring. The
functional and operability requirements for the power range channels
are not affected by deleting the requirement for determining
detector voltage plateaus.
Therefore, based on the conclusions of the above evaluation, the
proposed changes will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The functional and operability requirements for the new
detector system are the same as for the existing system as defined
by the Technical Specifications. No credit is taken for the source
and intermediate range trips in any of the design basis accidents.
The high flux at shutdown alarm and post-accident monitoring
functions continue to be met. The functional and operability
requirements for the power range channels are not affected by
deleting the requirement for determining detector voltage plateaus.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The functional and operability requirements for the new
detector system are the same as for the existing system. The
functional and operability requirements for the power range channels
are not affected by deleting the requirement for determining
detector voltage plateaus. The margin of safety provided by the
previous Technical Specifications is not significantly affected
because the proposed changes are based on the same accident analysis
acceptance limits.

[[Page 53958]]

Therefore, the proposed changes in this license amendment will
not result in a significant reduction in the plant's 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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket No. 50-260 Browns Ferry Nuclear
Plant Unit 2 Limestone County, Alabama

Date of amendment request: September 8, 1998.
Description of amendment request: The proposed amendment would
revise the Browns Ferry Nuclear Plant (BFN) Unit 2 technical
specifications (TS) to include provisions for enabling the Oscillation
Power Range Monitor (OPRM) Upscale trip function in the Average Power
Range Monitor (APRM). The APRM is part of the Power Range Neutron
Monitoring (PRNM) system. The OPRM Upscale trip function provides
protection from exceeding the fuel Minimum Critical Power Ratio (MCPR)
safety limit in the event of thermal-hydraulic power oscillations, and
thereby, provides compliance with Title 10 Code of Federal Regulations,
Part 50, Appendix A, General Design Criteria (GDC) 10 and 12.
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. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed amendment is to enable the OPRM Upscale trip
function which is contained in the previously installed PRNM
equipment. Enabling the OPRM hardware provides the long term
stability solution required by Generic Letter 94-02.
This hardware incorporates the Option III detect and suppress
solution reviewed and approved by the NRC in NEDO-31960, ``BWROG
[Boiling Water Reactor Owners Group] Long Term Stability Solutions
Licensing Methodology.'' The OPRM is designed to meet all
requirements of GDC 10 and 12 by automatically detecting and
suppressing design basis thermal-hydraulic power oscillations prior
to violating the fuel MCPR Safety Limit. The OPRM system provides
this protection in the region of the power-to-flow map where
instabilities can occur, including the region where ICAs (interim
corrective actions) restricted operation because of stability
concerns. Thus, the ICA restrictions on plant operations are deleted
from the TS, including region avoidance and the requirement for the
operator to manually scram the reactor with no recirculation loops
operating. Operation at high core powers with low core flows may
cause a slight, but not significant, increase in the probability
that an instability can occur. This slight increase is acceptable
because subsequent to the automatic detection of a design basis
instability, the OPRM Upscale trip provides an automatic scram
signal to the RPS [reactor protection system] which is faster
protection than the operator-initiated manual scram required by the
current ICAs. Because of this rapid automatic action, the
consequences of an instability event are not increased as a result
of the installation of the OPRM system because it eliminates
dependence on operator actions.
Based on the above discussion, the proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed amendment permits BFN to enable the OPRM power
oscillation detect and suppress function provided in previously
installed PRNM hardware, and it simultaneously deletes certain
restrictions which preclude operation in regions of the power-to-
flow map where oscillations potentially may occur. Enabling the OPRM
Upscale trip function does not create any new system hardware
interfaces nor create any new system interactions. Potential
failures of the OPRM Upscale trip result either in failure to
perform a mitigation action or in spurious initiation of a reactor
scram. These failures would not create the possibility of a new or
different kind of accident. Based on the above discussion, the
proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The OPRM Upscale trip function implements BWROG Stability Option
III, which was developed to meet the requirements of GDC 10 and GDC
12 by providing a hardware system that detects the presence of
thermal-hydraulic instabilities and automatically initiates the
necessary actions to suppress the oscillations prior to violating
the MCPR Safety Limit. The NRC has reviewed and accepted the Option
III methodology described in Licensing Topical Report NEDO-31960 and
concluded this solution will provide the intended protection.
Therefore, it is concluded that there will be no reduction in the
margin of safety as defined in TS as a result of enabling the OPRM
Upscale trip function and simultaneously removing the operating
restrictions previously imposed by the ICAs.

Based on the above discussion, the proposed amendment does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
its 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: Athens Public Library, 405 E.
South Street, Athens, Alabama 35611.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.

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

Date of amendment request: August 27, 1996, and as supplemented on
July 22, 1998.
Description of amendment request: The amendment request removes the
Technical Specification requirements for the Main Steam Isolation Valve
Leakage Control System, and increases the allowable leak rate specified
for the main steam lines. The Perry facility is a pilot plant in the
collaborative efforts of the Nuclear Regulatory Commission, the Nuclear
Energy Institute, and the Electric Power Research Institute for
implementation of the NRC research documented in NUREG-1465, ``Accident
Source Terms for Light-Water Nuclear Power Plants.'' The proposed
changes are based on reanalysis of the design basis Loss of Coolant
Accident using the revised accident source term from NUREG-1465 and the
NEI document entitled ``Generic Framework for Application of Revised
Accident Source Term to Operating Plants.''
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.

[[Page 53959]]

The proposed change removes the Technical Specification
requirements for the Main Steam Isolation Valve Leakage Control
System (MSIV-LCS), and increases the allowable leak rate specified
for the main steam lines. Although the requirements for the MSIV-LCS
are being removed (since credit is no longer taken for the system as
part of the design basis accident analysis), OPERABILITY
requirements on the Main Steam Shutoff Valves are being retained
since the valves meet Criterion 3 of 10 CFR 50.36(c)(2)(ii).
Removing the Technical Specification requirements of the MSIV-LCS
and increasing main steam line allowable leakage rates has been
addressed in the Loss of Coolant Accident (LOCA) reanalysis and does
not adversely affect operation of other equipment or systems
important to safety. These changes do not affect the precursors for
accidents or transients analyzed in Chapter 15 of the Perry Nuclear
Power Plant (PNPP) Updated Safety Analysis Report (USAR). Therefore,
there is no increase in the probability of accidents previously
evaluated.
The spectrum of LOCAs was considered to determine which would be
most limiting with respect to radiological consequences. The worst
case LOCA (i.e., main steam line break upstream of the inboard MSIV)
off-site and Control Room doses have been reanalyzed using the
revised design basis accident (DBA) source term (from NUREG-1465 and
the Nuclear Energy Institute (NEI) document ``Generic Framework for
Application of Revised Accident Source Term to Operating Plants'')
in order to assess the radiological consequences of the increased
main steam line leak rates, and not taking credit for the MSIV-LCS.
The radiological analysis used conservative assumptions and
analytical techniques. These conservatisms in the LOCA reanalysis
have been determined to be comparable to the conservatisms utilized
in the original analyses.

The results of the off-site and Control Room dose reanalysis are
provided below.

Dose Results (REM)
----------------------------------------------------------------------------------------------------------------
Existing USAR Regulatory
Proposed USAR dose* dose limit **
----------------------------------------------------------------------------------------------------------------
Control Room......................... Whole Body............... 0.1 0.4 5
Thyroid.................. 16.2 29.2 30
Skin..................... 4.8 2.5 30
EAB.................................. Whole Body............... 1.9 3.6 25
Thyroid.................. 157.9 140.8 300
LPZ.................................. Whole Body............... 1.7 1.9 25
Thyroid.................. 130.3 144.7 300
----------------------------------------------------------------------------------------------------------------
* Rounded to nearest tenth.
** Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) dose limits are per 10 CFR 100.11. Control Room
dose limits are per 10 CFR part 50 Appendix A, General Design Criterion (GDC) 19 and NUREG 0800 Standard
Review Plan (SRP) Section 6.4.

As noted in the NEI Generic Framework Document (``Generic
Framework for Application of Revised Accident Source Term to
Operating Plants,'' EPRI TR-105909, Interim Report, November 1995),
the acceptability of applications utilizing the revised accident
source terms ``may be judged by the same licensing acceptance limits
(e.g., dose limits in 10 CFR part 100) in use with the TID-14844
source term. That is, the licensee would show that the revised
design basis, with either selective or essentially complete
application of NUREG-1465 together with the plant changes under
evaluation, results in doses no greater than these licensing
acceptance limits.'' The off-site dose licensing acceptance limit
for PNPP is 10 CFR part 100.11 (see Question 3 for details on the
source of this PNPP licensing acceptance limit). The newly
calculated radiological doses were lower for six of the seven
factors evaluated. For the one factor which was higher, i.e., at the
EAB for thyroid dose (from 140.8 REM to 157.9 REM), the dose
remained significantly below the 10 CFR part 100 limit of 300 REM to
the thyroid. This analysis demonstrated that the resulting off-site
and Control Room doses were well below the regulatory limits
contained in 10 CFR part 100, Reactor Site Criteria, and 10 CFR part
50, Appendix A, General Design Criterion (GDC) 19, Control Room.
Therefore, the proposed changes do not involve a significant
increase in the consequences of previously evaluated accidents.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change removes the Technical Specification
requirements for the MSIV-LCS, retains the Technical Specification
requirements for the Main Steam Shutoff Valves, and increases the
allowable leak rate specified for the main steam lines.
Removing the Technical Specification requirements for the MSIV-
LCS is based on reanalysis of off-site and Control Room doses, where
the MSIV-LCS is not credited in the calculation. As noted above, the
reanalysis utilizes the revised design basis accident (DBA) source
terms. The limiting reanalysis case assumes that main steam line
leakage is attenuated in the main steam line from the reactor vessel
out to the outboard MSIV. This is the limiting scenario since the
worst case single failure, and hence the most limiting analysis
case, involves a failure to close the valve downstream of the
outboard MSIV in each main steam line, i.e., the Main Steam Shutoff
Valves (1N11F0020A,B,C AND D). Although this most limiting analysis
case assumes a failure to close the Main Steam Shutoff Valves,
retention of OPERABILITY requirements on these valves is appropriate
to ensure the single failure analysis remains valid.
Not crediting the MSIV-LCS in the design basis accident analysis
is consistent with the approach taken by several BWR licensees,
which have applied for NRC approval of this change using an approach
developed by the Boiling Water Reactor Owners Group (BWROG). The
BWROG methodology involves seismically qualifying the main steam
lines out to and including the non-safety related, non-seismic drain
line and main condenser, and then using that volume to attenuate
leakage past the MSIVs. At PNPP, the existence of safety related,
seismically qualified piping leading to the safety related, Class 1E
powered Main Steam Shutoff Valves (downstream of the outboard MSIV),
together with the characteristics of the revised accident source
term (i.e., predominantly aerosol which is largely retained in the
drywell, containment and main steam lines) provides the option of
taking credit only for the volume within the main steam lines for
leakage attenuation.
Knowledge of the more physically correct source term timing and
chemical form permits use of more appropriate mitigation techniques.
Specifically, natural forces such as gravitational settling of
aerosol (particulates) has been credited inside the drywell and in
portions of the main steam lines, which significantly reduces the
amount of radionuclides that could escape from the containment and
into the environment. Also, based on a high radiation signal in the
Control Room, the Containment Spray system would be operated post-
LOCA for up to 24 hours (previous analyses assumed 6 hours of spray
operation), in order to scrub released radionuclides from the
containment atmosphere and into the suppression pool, and thus
reduce the post-LOCA off-site and Control Room dose. Once the
containment sprays have been successful in sweeping the iodine to
the suppression pool, the iodine must be retained in the water. To
achieve this, the pH level of the suppression pool will now be
raised to 7 or above following the accident, and then maintained at
7 or above. This prevents significant fractions of the dissolved
iodine from being converted to elemental iodine and then re-evolving
to the containment atmosphere. During the course of the accident the
pH of the suppression pool can decrease due to radiolysis of reactor
coolant and chloride-bearing electrical insulation, which would
create acids. The

[[Page 53960]]

method for pH control will use the existing Standby Liquid Control
(SLC) system for raising (and maintaining) long-term post-accident
pH levels to 7 or above. Calculations have shown that the contents
of one tank of the Standby Liquid Control solution will be effective
in raising and maintaining pH levels for 30 days following the DBA.
Post-accident operator actions are minimized. The operator
action associated with initiating the Containment Spray system does
not change. Containment Spray is initiated via a push button in the
Control Room. The previously required manual initiation of the MSIV-
LCS involved multiple operator actions to open and close numerous
valves and start the blowers, which will no longer be required.
Replacing these actions, the new analysis simply assumes the
operator closes the Main Steam Shutoff Valves (which was previously
one of the steps in manually initiating the MSIV-LCS system), and
based on post-accident pH samples of the suppression pool, initiates
the Standby Liquid Control system, which is accomplished via two key
lock switches in the Control Room. These operator actions are less
complex than those previously required, and minimize the probability
of an error.
Other accidents, as described in USAR 15, were reviewed. The
original methodology, input parameters and overall conclusions
contained within these accident evaluations were found to be
unaffected by the changes proposed by this activity. Removing the
Technical Specification requirements of the MSIV-LCS and increasing
MSIV allowable leakage rates has been addressed in the LOCA
reanalysis and does not adversely affect operation of other
equipment or systems important to safety. This activity does not
alter or impact plant systems, structures or components which were
not appropriately addressed in the LOCA reanalysis. No new accident
initiator or failure mode is introduced. The physical isolation of
the MSIV-LCS from the Main Steam system will eliminate leakage
pathways. This modification will be performed as part of the PNPP
design change process.
With respect to the change in main steam line leakage limits,
the BWROG has concluded, based on an in-depth evaluation of MSIV
leakage (as discussed in NEDC-31858 ``BWROG Report for Increasing
MSIV Leakage Rate Limits and Elimination of Leakage Control
Systems,'' Revision 2, and summarized in NUREG-1169 ``Technical
Findings Related to Generic Issue C-8; Boiling Water Reactor Main
Steam Isolation Valve and Leakage Treatment Methods''), that leakage
rates of up to 500 scfh are not indicative of substantial mechanical
defects in the valves which would challenge the capability of the
valves to fulfill their safety function of isolating the steam
lines. Therefore, as demonstrated in the design basis LOCA
radiological reanalysis, the proposed increased allowable MSIV
leakage rate (i.e., each line less than or equal to 100 scfh and
total leakage less than or equal to 250 scfh when tested at Pa) will
not affect each MSIV's isolation function capability. Additionally,
no new operator actions or errors are introduced as a result of the
increased main steam line leakage limits, other than those addressed
above.
Based on the above discussions, the proposed change would not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change will not involve a significant reduction
in the margin of safety.
The worst case LOCA (i.e., a main steam line break upstream of
the inboard MSIV) has been reanalyzed using the revised DBA source
term (NUREG-1465 and the NEI generic framework document) in order to
assess the radiological consequences of the increased MSIV leak
rate, and not taking credit for MSIV-LCS. The radiological analyses
used conservative assumptions and analytical techniques. The results
of the revised DBA source term dose calculations should be
determined acceptable using the current licensing basis acceptance
limits (those that were used for initial plant licensing).
As noted in the NEI Generic Framework Document (``Generic
Framework for Application of Revised Accident Source Term to
Operating Plants,'' EPRI TR-105909, Interim Report, November 1995),
``to demonstrate that an adequate margin of safety is maintained,
the licensee may show that the doses associated with the revised
design basis (resulting from the revised source term together with
the plant change under evaluation) are less than the licensing
acceptance limits for the plant.''
The licensing acceptance limits for off-site dose are discussed
in Supplement 8 to the NRC Safety Evaluation Report (SER) for PNPP,
Section 15.3, ``Radiological Consequences of Design Basis
Accidents.'' The licensing acceptance limits are the guideline
values of 10 CFR 100.11, ``Reactor Site Criteria.'' The SER states
``The doses computed for this accident are less than the guideline
values of 10 CFR 100.11 and the staff concludes that the Perry plant
is adequately designed to mitigate the off-site consequences arising
from a LOCA.'' For Control Room doses, the licensing acceptance
limit is discussed in Supplement 10 to the NRC SER, Section 6.4,
``Control Room Habitability.'' The licensing acceptance limits are
as stated therein, i.e., ``The staff's LOCA analysis indicates that
the Control Room doses are within the guidelines of General Design
Criterion (GDC) 19 of Appendix A to 10 CFR part 50 and of Section
6.4 of the Standard Review Plan (SRP, NUREG-0800).''
The revised PNPP design basis calculations (i.e., the revised
DBA source term coupled with the plant changes under evaluation)
demonstrated that the resulting off-site and Control Room doses were
below the licensing acceptance limits contained in 10 CFR part 100,
10 CFR part 50, Appendix A, General Design Criterion 19, and SRP
Section 6.4. An acceptable margin of safety is inherent in these
licensing acceptance limits. The improvement in the technical
knowledge base and in the analytical techniques that are part of the
revised accident source term, and the modeling of the increased MSIV
leakages without taking credit for MSIV-LCS, do not alter the
acceptability of the margin. Therefore, the resulting calculated
Control Room and off-site doses, which are well within regulatory
limits, ensure that the proposed change does not involve a
significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Acting Project Director: Ronald R. Bellamy.

Date of amendment request: September 3, 1998.
Description of amendment request: The proposed license amendment
increases the present Division 3 Diesel Generator (High Pressure Core
Spray System) fuel level requirements to account for (1) a rounding
error in the calculation, and (2) the unusable volume due to vortex
formation at the eductor nozzles located in the fuel oil storage tank.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:

(1) The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change revises the Division 3 Diesel Generator (DG)
7-day fuel oil supply requirement and the 6-day fuel oil supply
requirement due to a rounding error in the calculation and due to
the consideration of vortex formation near the eductor suction
nozzle located near the bottom of the fuel oil storage tank. The
proposed change ensures a sufficient DG fuel oil volume to maintain
submergence of the eductor suction nozzle so that a vortex formation
does not occur. Eliminating the concerns of a vortex formation will
provide assurance that the DG fuel oil system will perform its
intended function. Analyzed events are initiated by the failure of
plant structures, systems, or components. The DGs are not considered
as initiators of any analyzed event. The proposed change does not
have a detrimental

[[Page 53961]]

impact on the integrity of any plant structure, system, or component
that initiates an analyzed event. The proposed change will not alter
the operation of, or otherwise increase its failure probability of
any plant equipment that initiates an analyzed event. As such, the
probability of occurrence for a previously analyzed accident is not
significantly increased.
The consequences of a previously analyzed event are dependent on
the initial conditions assumed for the analysis, the availability
and successful functioning of the equipment assumed to operate in
response to the analyzed event, and the setpoints at which these
actions are initiated. The proposed change ensures a sufficient DG
fuel oil volume to maintain submergence of the eductor suction
nozzle so that a vortex formation does not occur. The proposed
change continues to ensure that the DG fuel oil system will
adequately support the design basis performance and mitigative
function of the DG. The proposed change does not affect the
performance of any credited equipment. As a result, no analyses
assumptions are violated and there are no adverse effects on the
factors that contribute to offsite or onsite dose as the result of
an accident. The proposed change does not affect setpoints that
initiate protective or mitigative actions. The proposed change
ensures that plant structures, systems, or components are maintained
consistent with the safety analysis and licensing bases. Based on
this evaluation, there is no significant increase in the
consequences of a previously analyzed event.
Therefore, this change will not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
(2) The proposed change would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change revises the Division 3 DG 7-day fuel oil
supply requirement and the 6-day fuel oil supply requirement due to
a rounding error in the calculation and due to the consideration of
vortex formation near the eductor suction nozzle located near the
bottom of the fuel oil storage tank. The proposed change ensures a
sufficient DG fuel oil volume to maintain submergence of the eductor
suction nozzle so that a vortex formation does not occur.
Eliminating the concerns of a vortex formation will provide
assurance that the DG fuel oil system will perform its intended
function. The proposed change does not involve a physical change to
the DG fuel oil system or tank, nor does it change the operating
characteristics or the safety function of the DG. The proposed
change does not involve a physical alteration of the plant. No new
or different equipment is being installed and no installed
equipment, which might initiate a new or different kind of accident,
is being operated in a different manner. The proposed change does
not impact core reactivity or the manipulation of fuel bundles. The
DG performs a mitigative function. There is no alteration to the
parameters within which the plant is normally operated or in the
setpoints that initiate protective or mitigative actions. As a
result no new failure modes are being introduced. There are no
changes in the methods governing normal plant operation, nor are the
methods utilized to respond to plant transients altered.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) The proposed change will not involve a significant reduction
in the margin of safety.
The margin of safety is established through the design of the
plant structures, systems, and components, the parameters within
which the plant is operated, and the establishment of the setpoints
for the actuation of equipment relied upon to respond to an event.
The proposed change revises the Division 3 DG 7-day fuel oil supply
requirement and the 6-day fuel oil supply requirement due to
rounding error in the calculation and due to the consideration of
vortex formation near the eductor suction nozzle located near the
bottom of the fuel oil storage tank. The margin of safety is being
maintained by the proposed change from the margin of safety
established by the original design. The proposed change ensures a
sufficient DG fuel oil volume to maintain submergence of the eductor
suction nozzle so that vortex formation does not occur. Eliminating
the concerns of a vortex formation will provide assurance that the
DG fuel oil system will perform its intended function. The proposed
change does not significantly impact the condition or performance of
structures, systems, and components relied upon for accident
mitigation. The proposed change, in fact, provides assurance of the
DG's ability to perform its intended function as previously
evaluated. The proposed change does not significantly impact any
safety analysis assumptions or results.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Acting Project Director: Ronald R. Bellamy.

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

Date of amendment request: September 8, 1998.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 5.3.1, ``Design Features--
Reactor Core--Fuel Assemblies.'' A different type of fuel rod cladding
would be added. The associated bases would also be changed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:

The Davis-Besse Nuclear Power Station has reviewed the proposed
changes and determined that a significant hazards consideration does
not exist because operation of the Davis-Besse Nuclear Power Station
in accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because it has been demonstrated that
the material properties of the M5 alloy are not significantly
different from those of Zircaloy-4. Further, there are no evaluated
accidents in which the fuel cladding or fuel assembly structural
components are assumed to arbitrarily fail as an accident initiator.
The fuel handling accident assumes that the cladding does, in fact,
fail as a result of an undefined fuel handling event. However, the
probability of that undefined initiating event is independent of the
properties of the fuel rod cladding.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because it has been demonstrated that
the material properties of the M5 alloy are not significantly
different from those of Zircaloy-4. Therefore, in both non-LOCA and
LOCA accident scenarios, there will be no significant increase in
cladding failure or fission product release.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because it has been
demonstrated that the material properties of the M5 alloy are not
significantly different from those of Zircaloy-4. Therefore, M5 fuel
cladding and fuel assembly structural components will perform
similarly to those fabricated from Zircaloy-4, thus precluding the
possibility of the fuel becoming an accident initiator and causing a
new or different kind of accident.
3. Not involve a significant reduction in a margin of safety
because it has been demonstrated that the material properties of the
M5 alloy are not significantly different from those of Zircaloy-4.
The M5 alloy is expected to perform similarly to Zircaloy-4 for all
normal operating and accident scenarios, including both non-LOCA and
LOCA scenarios. For LOCA scenarios, where the slight differences in
M5 material properties relative to Zircaloy-4 could have

[[Page 53962]]

some impact on the overall accident scenario, plant-specific LOCA
analyses will be performed prior to the use of batch quantities of
fuel assemblies containing either fuel rod cladding, fuel rod end
plugs, or fuel assembly structural components fabricated from M5.
These plant-specific LOCA analyses, required by TS 6.9.1.7, ``Core
Operating Limit Report,'' will either demonstrate that all current,
applicable, and appropriate margins of safety will be maintained
during the use of the M5 alloy or their results will be submitted
for NRC review and approval prior to use of the M5 alloy.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Acting Project Director: Ronald R. Bellamy.

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

Date of amendment request: August 20, 1998.
Description of amendment request: By letter dated August 20, 1998,
the licensee submitted a License Amendment request related to three
Technical Specification (TS) administrative changes. The first is to
remove a definition from the DEFINITIONS section of the TS that is
provided in 10 CFR part 20. The second change is to transfer the site
map from Section 5.0 of the TS to the Final Safety Analysis Report and
to replace the map with a textual description of the site location.
Lastly, to delete TS 5.1.1--EXCLUSION AREA.
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 are administrative in nature and in no way
affect the safety of the Yankee Nuclear Power Station (YNPS). The
proposed deletion of the definition for SITE BOUNDARY in no way
reduces or eliminates any regulatory requirement which Yankee Atomic
Electric Company must currently satisfy. Likewise, the relocation of
the YNPS site map from the YNPS Technical Specifications to the YNPS
Final Safety Analysis Report is devoid of any safety implications.
Therefore, the proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. The administrative
nature of the changes will not affect safety related systems or
components and, therefore, involve no significant increase in the
probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different accident from
any previously evaluated. The proposed changes do not modify any
plant systems or components and, therefore, do not create the
possibility of a new or different accident from any previously
evaluated.
3. Involve a significant reduction in the margin of safety. The
proposed changes do not involve any physical changes to the plant
nor any changes in plant procedures. Therefore, there will be no
reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Greenfield Community College,
1 College Drive, Greenfield, Massachusetts 01301.
Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One
International Place, Boston, Massachusetts 02110-2624.
NRC Project Director: Seymour H. Weiss.

Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed no Significant
Hazards Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.

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: August 27, 1998.
Brief description of amendment request: The amendment revises
Technical Specifications 3.0.4 and 4.0.4 to be consistent with the
guidance provided in Generic Letter 87-09 dated June 4, 1987.
Date of publication of individual notice in Federal Register:
September 8, 1998 (63 FR 47529).
Expiration date of individual notice: October 8, 1998.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

Date of amendment request: August 21, 1998.
Description of amendment request: The amendment would remove the
requirement for the Automatic Depressurization System function of the
Electromatic Relief Valves to be operable during Reactor Vessel
Pressure Testing. Additionally, note h of Table 3.1.1 will be corrected
due to a typographical error introduced in the issuance of Amendment
75.
Date of publication of individual notice in Federal Register:
September 10, 1998 (63 FR 48527).
Expiration date of individual notice: October 13, 1998.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Wisconsin Public Service Corporation, Wisconsin Power and Light Company
and Madison Gas and Electric Company, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, WI

Date of application for amendment: April 8, 1998, modified by
letter dated August 27, 1998.
Brief description of amendment request: The proposed amendment
would reduce the maximum allowable level of reactor coolant system
activity (dose equivalent 1-131) to provide a means of accepting higher
projected leak rates for steam generator tubes while still meeting
offsite and control room dose criteria. Also included is a change to
the secondary coolant activity level for which an increased sampling
frequency applies.

[[Page 53963]]

Date of publication of individual notice in Federal Register:
September 14, 1998 (63 FR 49137).
Expiration date of individual notice: October 14, 1998.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

Notice of Issuance of Amendments to Facility Operating Licenses

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 application for amendment: March 12, 1998, as supplemented
August 14, 1998. The August 14, 1998, supplemental letter provided
clarifying information only, and did not change the initial no
significant hazards consideration determination.
Brief description of amendment: This amendment deletes Technical
Specification surveillance requirement 4.9.12.d.4, which requires
verification at least once every 18 months that the Fuel Handling
Building Emergency Exhaust System filter cooling bypass valve is locked
in the balanced position.
Date of issuance: September 11, 1998.
Effective date: September 11, 1998.
Amendment No: 82.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 8, 1998 (63 FR
17222).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 11, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

Date of application for amendment: January 28, 1998 (NRC-98-0003)
as supplemented March 10, 1998
Brief description of amendment: The amendment revises technical
specification (TS) 3.4.10, TS Figure 3.4.10-1, and the associated bases
by changing the prohibited and restricted operating region associated
with core thermal-hydraulic stability. Also, TS 3.4.1.4, TS Figure
3.4.1.4-1, and the associated bases are revised to reflect stability-
related improvements in operating restrictions for idle recirculation
loop startup. Finally, in an unrelated change, TS Tables 3.3.7.5-1 and
4.3.7.5-1 are revised to delete neutron flux from the list of accident
monitoring instrumentation in TS 3.3.7.5.
Date of issuance: September 16, 1998
Effective date: September 16, 1998, with full implementation within
90 days.
Amendment No.: 128.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9598). The March 10, 1998, letter provided clarifying information
that was within the scope of the original Federal Register notice and
did not change the staff's initial proposed no significant hazards
considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.

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

Date of application for amendment: June 5, 1998 (NRC-98-0067), as
supplemented August 24, 1998.
Brief description of amendment: The amendment revises Technical
Specification (TS) 2.1.2, ``Thermal Power, High Pressure and High
Flow,'' by changing the values for the safety limit minimum critical
power ratio from 1.09 to 1.11 for two recirculation loop operation and
from 1.11 to 1.13 for single recirculation loop operation for Cycle 7.
The amendment also revises the footnote to TS 2.1.2 to indicate that
these revised values are applicable for Cycle 7 operation only.
Date of issuance: September 21, 1998.
Effective date: September 21, 1998, with full implementation prior
to restart from the sixth refueling outage.
Amendment No.: 129.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 1, 1998 (63 FR
35988). The August 24, 1998, letter provided clarifying information
that was within the scope of the original Federal Register notice and
did not change the staff's initial proposed no significant hazards
considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 21, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.

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

Date of application for amendments: May 8, 1998.
Brief description of amendments: The amendments revise the Power
Range Neutron Flux Trip setponts in the event of inoperable main steam
safety valves.

[[Page 53964]]

Also, the amendments delete the reference to three-loop operation.
These changes are consistent with the proposed Improved Standard
Technical Specifications submitted by the licensee on May 27, 1997.
Date of issuance: September 17, 1998.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1--181; Unit 2--163.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40554).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 17, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.

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

Date of application for amendments: October 6, 1998, as
supplemented by letter dated August 24, 1998.
Brief description of amendments: The amendments delete all
references to the steamline low pressure safety injection function.
Date of issuance: September 22, 1998.
Effective date: As of the date of issuance to be implemented in the
refueling outage associated with the plants' hardware modifications.
Amendment Nos.: Unit 1--182; Unit 2--164.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 19, 1997 (62
FR 61841).
The August 24, 1998, submittal provided clarifying information that
did not change the scope of the original Federal Register notice, and
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 22, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas

Date of amendment request: October 2, 1996, as supplemented by the
letter dated June 18, 1997.
Brief description of amendments: The amendments relocate the
Radiological Effluents Technical Specifications (RETS) to the Offsite
Dose Calculation Manual and the Process Control Program. The NRC
provided guidance to all power reactors licensees and applicants on the
proposed TS changes in Generic Letter 89-01, ``Implementation of
Programmatic Controls for Radiological Effluent Technical
Specifications in the Administrative Controls Section of the Technical
Specifications and the Relocation of Procedural Details of RETS to the
Offsite Dose Calculation Manual or to the Process Control Program,''
dated January 31, 1989.
Date of issuance: September 23, 1998.
Effective date: September 23, 1998.
Amendment Nos.: Unit 1; 193 and Unit 2; 193.
Facility Operating License Nos. DPR-51 and NPF-6: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 15, 1997 (62 FR
2188).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of application for amendments: March 3, 1998.
Brief description of amendments: The amendments modify surveillance
requirement 4.6.4.2.b.4 for Unit 1 and the Technical Specification
bases 3/4.6.4 for Unit 1 and 2.
Date of issuance: September 17, 1998.
Effective date: September 19, 1998, with full implementation within
45 days.
Amendment Nos.: 223 and 207.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 1, 1998 (63 FR
35990).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 17, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, MI 49085.

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

Date of application for amendment: July 16, 1998, as supplemented
September 3, 1998. The application dated July 16, 1998, supersedes a
July 2, 1997, application in its entirety.
Brief description of amendment: The amendment changes Technical
Specification 3/4.2.3 regarding reactor coolant chemistry in accordance
with a report by Electrical Power Research Institute, Inc., TR-103515-
R1, ``BWR Water Chemistry Guidelines, 1996 Revision.''
Date of issuance: September 18, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 163.
Facility Operating License Nos. DPR-63 and NPF-69: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: August 13, 1998 (63 FR
43432).
The September 3, 1998, submittal contained clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 18, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Project Director: S. Singh Bajwa.

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

Date of application for amendment: August 23, 1995.
Brief description of amendment: The amendment extends the Technical
Specification (TS) Allowed Outage Time (AOT) for an inoperable Safety
Injection Tank (SIT) from 1 hour to 24 hours, unless the SIT is
inoperable due to

[[Page 53965]]

either boron concentration not within its limits or an inoperable water
level or nitrogen cover pressure instrument. The proposed change, for
these two special cases, extends the AOT for an inoperable SIT to 72
hours. In addition, the completion times and conditions for action
statements and the criteria for surveillance requirements are changed.
The TS Bases are also updated to reflect the changes.
Date of issuance: September 3, 1998.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 220 Facility Operating License No. DPR-65: Amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: September 13, 1995 (60
FR 47621).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 3, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.

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

Date of application for amendment: August 6, 1998, as supplemented
September 3 and 21, 1998.
Brief description of amendment: The latest Millstone Unit 3 steam
generator tube inspection began on September 24, 1996, and was
completed on October 1, 1996. The inspection results placed the steam
generators in Category C-2. Technical Specification Surveillance
Requirement 4.4.5.3.a establishes an allowable inspection interval of
24 calendar months for this category. Without an extension of the
interval, Millstone Unit 3 must shut down prior to September 24, 1998,
to perform the necessary inspections. The amendment allows a one-time
extension to the surveillance interval until the next refueling outage
or July 1, 1999, whichever date is earlier.
Date of issuance: September 23, 1998. Effective date: As of the
date of issuance to be implemented within 30 days from the date of
issuance.
Amendment No.: 163.
Facility Operating License No. NPF-49: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 17, 1998 (63 FR
43964).
The September 3 and 21, 1998, letters provided clarifying
information that did not change the scope of the August 6, 1998,
application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 23, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.

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

Date of application for amendment: July 26, 1996, as supplemented
September 5, 1997, as revised December 4, 1997, and as supplemented
March 6, March 26, April 8, April 17, April 22, May 5, May 12, May 29,
June 15, July 1, July 20, and July 30, 1998.
Brief description of amendment: The amendment revises the operating
license and the Technical Specifications to allow increase of the
maximum reactor core thermal power level from 1670 megawatts-thermal
(MWt) to 1775 MWt.
Date of issuance: September 16, 1998.
Effective date: September 16, 1998. Full implementation within 90
days of issuance.
Amendment No.: 102.
Facility Operating License No. DPR-22. Amendment revised the
License and the Technical Specifications.
Date of publication of individual notice in Federal Register:
February 25, 1998 (63 FR 9606).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 16, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.

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

Date of application for amendments: February 27, 1998, as
supplemented July 14, 1998.
Brief description of amendments: The amendments allow a design
modification of the existing Anticipated Transient Without Scram (ATWS)
Mitigation System Actuation Circuitry (AMSAC). The design modification
installs a Diverse Scram System (DSS) designed to meet the requirements
of a DSS described by 10 CFR 50.62 (ATWS Rule) for non-Westinghouse
designed plants and make major modifications to the existing AMSAC.
Date of issuance: September 22, 1998. Effective date: September 22,
1998, with full implementation by the completion of the next scheduled
refueling outage.
Amendment Nos.: 138 and 129.
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the license to authorize a design modification of the existing
Anticipated Transient Without Scram (ATWS) Mitigation System Actuation
Circuitry (AMSAC).
Date of initial notice in Federal Register: August 17, 1998 (63 FR
43965).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 22, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.

Philadelphia Electric Company, Docket No. 50-171, Peach Bottom Atomic
Power Station, Unit 1, York County, Pennsylvania

Date of amendment request: March 2, 1998.
Brief description of amendment: This amendment will revise the
Peach Bottom Atomic Power Station, Unit 1, Technical Specifications
(TS) to include requirements for control of effluents and annual
reporting in accordance with the requirements of 10 CFR 50.36a.
Date of issuance: September 14, 1998.
Effective date: As of the date of its issuance and must be fully
implemented no later than 30 days from the date of issuance.
Amendment No.: 9.
Facility Operating License No. DPR-12: Amendment revised the
Technical Specifications.

[[Page 53966]]

Date of initial notice in Federal Register: July 1, 1998 (61 FR
35994). The NRC's related evaluation of the amendment is contained in a
Safety Evaluation dated September 14, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.

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

Date of amendment request: August 28, 1998.
Brief description of amendments: The amendments modified TS 4.0.5
to state that the inservice testing requirement for exercise testing in
the closed direction for specified Unit 1 containment isolation valves
shall not be required until the next plant shutdown to Mode 5 of
sufficient duration to allow the testing or until the next refueling
outage scheduled in March 1999.
Date of issuance: September 24, 1998.
Effective date: September 24, 1998, to be implemented within 7
days.
Amendment Nos.: Unit 1--Amendment No. 95; Unit 2--Amendment No. 82.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (63 FR 48254). The notice provided an
opportunity to submit comments on the Commission's proposed NSHC
determination. No comments have been received. The notice also provided
for an opportunity to request a hearing by October 8, 1998, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final determination of NSHC are contained in
a Safety Evaluation dated September 24, 1998.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

Date of amendment request: February 16, 1998, as supplemented April
2, July 15, and August 13, 1998.
Brief description of amendments: The amendments revised TS 3/4.4.5,
``Steam Generators,'' and its Bases to allow the implementation of 1-
volt voltage-based repair criteria for the steam generator tube support
plate-to-tube intersections for Unit 2 in accordance with Generic
Letter 95-05, and made related Unit 1 administrative changes for
consistency of wording (the Nuclear Regulatory Commission (NRC) had
previously approved a similar 1-volt voltage-based repair criteria
application for Unit 1). In addition, the amendments made an
administrative change to Bases 3/4.4.6.2, ``Operational Leakage,'' to
clarify that the allowable steam generator leakage specification
applies to both Unit 1 and Unit 2.
Date of issuance: September 24, 1998.
Effective date: September 24, 1998, to be implemented within 30
days.
Amendment Nos.: Unit 1--Amendment No. 96; Unit 2--Amendment No. 83.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: May 20, 1998 (63 FR
27765).
The additional information contained in the supplemental letters
dated July 15 and August 13, 1998, were clarifying in nature and thus,
within the scope of the initial notice and did not affect the staff's
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 24, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

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

Date of application for amendment: August 29, 1995, supplemented
June 25, 1998
Brief description of amendment: This amendment revises Technical
Specification Tables 3.3.5.1-1, ``Emergency Core Cooling System
Instrumentation,'' and 3.3.6.1-1, ``Primary Containment and Drywell
Isolation Instrumentation,'' by revising allowable values for selected
plant process instrumentation in accordance with Instrument Setpoint
Methodology Group and GE Topical Report NEDC-31336, ``General Electric
Instrument Setpoint Methodology,'' dated October 1986.
Date of issuance: September 15, 1998.
Effective date: September 15, 1998.
Amendment No.: 93.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62496)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 15, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.

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

Date of application for amendment: June 1, 1998, supplemented July
14, 1998

Brief description of amendment: The changes revise the F* and elevated
F* (EF*) criteria used to disposition indications in the roll expansion
joint of degraded steam generator (SG) tubes within the tubesheet.

Date of issuance: September 22, 1998.
Effective date: September 22, 1998.
Amendment No.: 138.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 1, 1998 (63 FR
35996)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated September 22, 1998. .
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001

[[Page 53967]]

Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of no Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)

During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Ch. I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By November 6, 1998, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner

[[Page 53968]]

must provide sufficient information to show that a genuine dispute
exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. A copy of the petition should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

Date of amendment request: September 18, 1998, as superseded by
letter dated September 23, 1998.
Description of amendment request: The amendment changes the
Appendix A TSs by revising Note ``1'' in Table 2.2-1, ``Reactor
Protective Instrumentation Trip Setpoint Limits'' and Note ``a'' in
Table 3.3-1, ``Reactor Protective Instrumentation,'' both applicable to
high logarithmic power reactor trip instrumentation. Additionally, the
requested changes clarify the terms RATED THERMAL POWER and THERMAL
POWER used in Tables 2.2-1, 3.3-1 and 4.3-1. A Bases change is made to
support these changes.
Date of issuance: September 24, 1998.
Effective date: September 24, 1998.
Amendment No: 145.
Facility Operating License No. NPF-38: Amendment revises the
Technical Specifications Public comments requested as to proposed no
significant hazards consideration: No. The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated September 24, 1998.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street NW., Washington, D.C. 20005-3502.
NRC Project Director: John N. Hannon.

Southern California Edison Company, et al., Docket No. 50-361, San
Onofre Nuclear Generating Station, Unit No. 2, San Diego County,
California

Date of application for amendment: September 22, 1998.
Brief description of amendment: The amendment revises the technical
specifications (TS) to change the operative parameter for setting and
removing the operating bypass bistables for Logarithmic Power Level--
High, Reactor Coolant Flow--Low, Local Power Density--High, and
Departure from Nucleate Boiling Ratio--Low trips. The operative
parameter specified in the TS is being changed from ``THERMAL POWER''
to logarithmic power.
Date of issuance: September 25, 1998.
Effective date: September 25, 1998.
Amendment No.: 142.
Facility Operating License No. NPF-10: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No. The Commission's related evaluation of the
amendments, finding of emergency circumstances, and final determination
of no significant hazards consideration are contained in a Safety
Evaluation dated September 25, 1998.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, P.O. Box 800, Rosemead, California 91770.
Local Public Document Room location: Main Library, University of
California, Irvine, California 92713

Dated at Rockville, Maryland, this 30th day of September 1998.
For the Nuclear Regulatory Commission.
John N. Hannon,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-26746 Filed 10-6-98; 8:45 am]
BILLING CODE 7590-01-P

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Document Information

Effective Date:: 9/11/1998
Published:: 10/07/1998
Department:: Nuclear Regulatory Commission
Entry Type:: Notice
Document Number:: 98-26746
Dates:: September 11, 1998.
Pages:: 53943-53968 (26 pages)
PDF File:: 98-26746.pdf