[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: 

    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 time 
to be in MODE 3 is not assumed to be the initiator of any analyzed 
events. As a result, the probability of an analyzed event is 
independent of the time permitted to be in MODE 3. The consequences 
of an accident occurring during the 12 hours permitted to be in MODE 
3 are no greater than the consequences of an accident occurring 
during the 6 hours currently permitted to place the unit in Hot 
Shutdown. Therefore, the probability and 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 time to place the unit in MODE 5 is appropriately 
limited. Therefore, 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 extended time to place the unit in MODE 3 is not 
significantly greater than the time currently permitted to place the 
unit in Hot Shutdown and represents a reasonable time to accomplish 
the shutdown. Therefore, the extended time to place the unit in MODE 
3 does not involve a significant reduction in the margin of safety.

    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: 

    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. BWST 
water temperature and volume are not

[[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.

    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).

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.

    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.

[[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.

    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.

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.

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: 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.

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: 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

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

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