[Federal Register Volume 64, Number 17 (Wednesday, January 27, 1999)]
[Notices]
[Pages 4152-4165]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-1705]


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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 January 4, 1999, through January 14, 1999. 
The last biweekly notice was published on January 13, 1999.

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By February 26, 1999, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended

[[Page 4153]]

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.

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

    Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois.
    Docket Nos. 50-373 and 50-374, LaSalle County Station, Units 1 and 
2, LaSalle County, Illinois.
    Date of application for amendment request: December 17, 1998.
    Description of amendment request: The amendments would revise the 
respective facility Technical Specifications (TS) by adding a new 
Limiting Conditions for Operations which provides an administrative 
enhancement by allowing testing required to return equipment to service 
to be conducted under administrative controls.
    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 has no impact on the design basis of the 
plant. The change has no impact on the response of the plant during 
normal or transient conditions. Incorporation of ISTS [improved 
Standard Technical Specification] 3.0.5 provides the necessary 
administrative controls that allow the return of equipment to 
service to complete testing required to demonstrate operability. 
Without this allowance, certain components could not be restored to 
operable status and a plant shutdown would ensue. It is not the 
intent of the TS to preclude the return to service of a component in 
order to confirm its operability or the operability of other 
equipment. This allowance is deemed to be a safer operation than 
requiring a plant shutdown to complete necessary testing. This 
allowance is considered acceptable because it: (1) is temporary; (2) 
accompanied by appropriate administrative controls, and; (3) 
provides a safety enhancement by restoring the plant status to, or 
confirming the existing plant status is in, a condition that is 
expected to provide for safe operation.
    ISTS 3.0.5 was adopted to address the ambiguity that ACTION 
requirements do not strictly allow the restoration of equipment to 
its normal configuration to perform functional testing required to 
demonstrate operability. The components involved will have completed 
maintenance and or testing that will demonstrate, with reasonable 
assurance, that the component can perform its intended safety 
function.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated:
    The proposed changes do not introduce new features or modify 
plant structures, systems or components that may impact station 
operations under normal or abnormal conditions. The proposed changes 
will allow the necessary testing to ensure safety related equipment 
will perform its design basis safety function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in the margin of safety for 
the following reasons:
    The proposed changes have no impact on any of the Safety Limits 
provided in the Technical Specifications, nor does the change impact 
the operation of structures, systems and components import to plant 
safety. The purpose of the proposed change is to return equipment to 
service, under administrative controls, to complete operability 
testing. Therefore, allowing the return of equipment to service will 
promote timely restoration of, or confirmation of, equipment 
operability thereby increasing the margin of safety from that 
existing with this equipment remaining out of service. Temporarily 
returning inoperable equipment to service for the purpose of 
confirming operability places the plant in a condition which has 
been previously evaluated and determined to be acceptable for short 
periods. Therefore, the proposed change does not involve a 
significant reduction in safety.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.


[[Page 4154]]


    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 
amendments requested involve no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021; for LaSalle, Jacobs Memorial Library, 815 North Orlando 
Smith Avenue, Illinois Valley Community College, Oglesby, Illinois 
61348-9692.
    Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice 
President and General Counsel, Commonwealth Edison Company, P.O. Box 
767, Chicago, Illinois 60690-0767.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: December 24, 1998.
    Description of amendment request: These amendment requests change 
the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2) 
Technical Specifications (TSs) to ensure that Emergency Diesel 
Generator (EDG) requirements contained in Technical Specification 3/
4.8.1 for both units are consistent with assumptions contained in 
design analyses and requirements of plant procedures. Revisions to TS 
3/4.8.1 ``A.C. Sources,'' contained in this amendment provide more 
conservative limiting conditions for operation (LCO) and surveillance 
requirements that affect EDG fuel oil storage volume, EDG load 
rejection and overspeed testing, and EDG operating frequency 
requirements. The applicable bases for each unit are also refined, as 
necessary, to strengthen the explanations regarding EDG fuel oil 
storage systems and provide the EDG overspeed in terms of frequency 
(Hertz) and speed (Revolutions Per Minute).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The addition of the term ``usable'' to LCO 3.8.1.1 and 3.8.1.2 
for both Units will assure that the required quantity of fuel oil 
will be available to operate the diesel during emergency conditions. 
This revision including the discussion contained in the Technical 
Specification Bases has no physical impact on the diesels or their 
setpoints. These revisions also do not delete any function 
previously provided by the diesels. There are no design bases 
accidents for which failure of the diesel is considered an 
initiating event. Therefore, the probability of an accident 
previously evaluated in the safety analysis is not increased by this 
change. The proposed changes do not involve an increase in the 
consequences of an accident previously analyzed, as they make the 
limiting condition for operation and associated bases more 
conservative and involve no physical changes to the diesels.
    The revised EDG single largest load rejection and overspeed 
criteria do not involve an increase in the probability or the 
consequences of accidents previously analyzed. The surveillance 
tests impacted by the proposed revision are performed only during 
shutdown when the opposite train EDG and its connected AC power 
system are relied upon as the emergency AC power source. Further, 
there are no design basis accidents for which changes to EDG load 
rejection test acceptance criteria can be an initiating event. The 
proposed changes affect the diesel testing requirements but do not 
affect the operating or design parameters. The changes also do not 
affect the diesels' ability to mitigate the consequences of an 
accident. They serve to ensure the ability of the diesel to reject 
the largest load. The overspeed criteria ensures that diesel 
frequency does not exceed a certain value subsequent to a load 
rejection. This criteria also ensures compliance with the guidance 
of Safety Guide 9 for Unit 1 and Regulatory Guide 1.9 for Unit 2. It 
does not involve an increase in the consequences of an accident 
previously analyzed. The revision does not impact accidents 
previously analyzed and would not, therefore, affect the 
consequences of accidents previously analyzed.
    Revising the EDG operating frequency as discussed in the 
proposed amendment protects [engineered safety feature] ESF pumps 
from runout conditions and motors from operating in an unanalyzed 
condition. The narrower frequency limits are more restrictive and 
have no adverse effect on the diesel generator operability. The 
proposed revision to decrease the EDG operating frequency limit does 
not involve an increase in the probability of an accident as 
described in the [Updated Final Safety Analysis Report] UFSAR. There 
are no design basis accidents for which failure of the diesel is 
considered an initiating event. A narrower operating frequency does 
not increase the probability of a design basis accident; it ensures 
that equipment performs their intended function. This change is 
intended to prevent the diesel from being loaded beyond analyzed 
loading limits and protect ESF equipment. The more conservative 
surveillance requirements being applied to operating limits will 
provide greater assurance that the diesels will be operable and that 
greater performance requirements are not imposed on ESF equipment. 
This change, therefore, will not result in an increase in the 
consequences of an accident previously described.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed revisions do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
They also will have no adverse impact on the design basis accidents 
previously evaluated in the UFSAR. The revisions contained in the 
proposed amendment are more restrictive to assure that diesel and 
ESF equipment are available and fully operable to perform their 
intended safety function following a design basis accident and a 
loss of offsite power. The proposed changes do not involve physical 
changes to plant equipment or the AC power system configuration. New 
failure modes are not introduced as a result of the proposed 
revisions. A revision of the diesel frequency will prevent motors 
and pumps from being subjected to over-frequency conditions which 
could reduce the life of the equipment. Increasing the load 
rejection criteria for Unit 1 and including overspeed criteria for 
both units revises surveillance test criteria for verifying load 
rejection capability. This does not affect the probability of 
malfunction of a diesel or its connected emergency AC power system. 
Further, it does not create a new failure mode. Revising diesel fuel 
oil storage requirements to include the term ``usable'' reduces the 
potential for misinterpretation of this specification; it does not 
create a new kind of accident from any accident previously 
evaluated.
    The revisions contained in this license amendment have the 
effect of making the BVPS Technical Specifications more conservative 
than previously. This license amendment request will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety is not reduced as a result of the proposed 
revisions. The margin of safety depends on the maintenance of 
specific operating parameters within design limits. The margin of 
safety derived from limiting condition for operation 3.8.1.1 and 
3.8.1.2 for both Units is enhanced by adding ``usable'' in these 
requirements. This revision reduces the possibility of 
misinterpreting Technical Specification requirements. The addition 
of diesel overspeed criteria (both units) and increasing load 
rejection criteria for Unit 1 does not reduce the margin of safety. 
Diesel reliability and performance during a loss of offsite power 
and a design basis accident are enhanced by this more conservative 
surveillance test requirement. Revision of diesel operating 
frequency limits protects engineered safety features equipment from 
overfrequency conditions; this would not be a significant reduction 
in the margin of safety. Though the temporary Unit 1 EDG loading 
limit of 2791.51 exceeds the Safety Guide 9 value of 2745, it still 
is below the EDG 2000 hour rating limit of 2850 kW contained in 
Surveillance Requirement 4.8.1.1.2.b.6. Further, the loading value 
of 2791.51 kW does not exceed the design

[[Page 4155]]

loading capability of the EDG. Based on engineering analyses, the 
revisions contained in the proposed amendment will not significantly 
reduce the margin of safety. Engineered safety features equipment 
will continue to function, as assumed in the safety analysis, to 
ensure that fuel, reactor coolant system and containment design 
limits are not exceeded.
    Therefore, this change will not involve a significant reduction 
in a margin of safety due to the continued availability and 
reliability of the A.C. electrical power sources.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: December 24, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specification (TS) requirements for the axial flux 
difference [AFD] monitor, quadrant power tilt ratio [QPTR] monitor, rod 
position deviation monitor, and rod insertion limit (RIL) monitor. The 
changes would (1) relocate requirements for the AFD monitor and the 
QPTR monitor to the Licensing Requirements Manual (LRM); (2) delete 
requirements for the rod position deviation monitor and RIL monitor 
from the TSs; (3) modify Unit 1 surveillance requirements (SR) 4.1.3.5 
and 4.1.3.6 by incorporating the Unit 2 wording to provide 
surveillances more consistent with the Limiting Condition for Operation 
(LCO); (4) change Unit 1 SR 4.1.3.2.2, SR 4.1.3.5, SR 4.1.3.6 and Unit 
2 SR 4.1.3.5 from 24 hour surveillance frequencies to 12 hour 
frequencies; and (5) delete Unit 1 SR 4.1.3.2.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment would modify applicable Technical 
Specifications (TS) by deleting requirements associated with the rod 
position deviation monitor and rod insertion limit (RIL) monitor and 
relocating the requirements associated with the axial flux 
difference (AFD) monitor and quadrant power tilt ratio (QPTR) 
monitor from the following specifications and Bases:

Unit 1: 4.1.3.1.2, 3.1.3.2, 4.1.3.2.2, 4.1.3.2.3, 4.1.3.6, 4.2.1.1, 
4.2.4;
Unit 2: 4.1.3.1.2, 4.1.3.2, 4.1.3.6, 4.2.1.1, 4.2.4.

    The TS contains requirements where a reduced surveillance 
interval is required in the event the monitors referenced in the 
above specifications, surveillance requirements (SR) and associated 
Bases are inoperable. Removing the requirements associated with 
these monitors from the TS will not affect the ability of any system 
to perform its design function.
    Nuclear Electric Institute (NEI) Technical Specification Task 
Force (TSTF) 110 Revision 2 provides the basis for these changes and 
recommends relocating the requirements for these monitors to ``plant 
administrative practices.'' The AFD monitor and the QPTR monitor 
requirements will be relocated to the LRM and changes to these 
requirements will be controlled in accordance with the 10 CFR 50.59 
process which will require NRC approval if the change constitutes an 
unreviewed safety question. However, based on the smaller change in 
surveillance intervals, deletion and not relocation of the rod 
position deviation monitor and the RIL monitor requirements can be 
justified and is proposed.
    Although these monitors are being removed from the TSs, they 
will continue to be maintained as described in the [Updated Final 
Safety Analysis Report] UFSAR (subject to revisions via the 10 CFR 
50.59 process). Removing the rod deviation monitor requirements from 
Unit 1 SR 4.1.3.2.3 makes the remaining portion of SR 4.1.3.2.3 
redundant to SR 4.1.3.2.2.a; therefore, SR 4.1.3.2.3 has been 
deleted. In addition, the 24-hour surveillance frequency in Unit 1 
SR 4.1.3.2.2, 4.1.3.5 and 4.1.3.6 as well as in Unit 2 SR 4.1.3.5 is 
being changed to 12 hours to assure the required parameters are 
adequately monitored and to provide consistency between the units 
and related requirements as well as the Improved Standard Technical 
Specifications (ISTS).
    Removing these monitors from the TS is consistent with the NRC 
approved changes to the ISTS identified in TSTF-110, Revision 2. 
Verification that plant conditions are within specified limits at 
the frequency specified in the normal SR provides sufficient 
information that allows the operator to detect a parameter that is 
beginning to deviate from its expected limits. The specified 
frequency takes into account other information (i.e., rod position 
indication system, rod bottom alarm and excore neutron detectors) 
that is continuously available to the operator in the control room, 
so that during changes in plant conditions, deviation from the 
limits can be readily detected.
    The proposed changes do not affect the operation of the system 
or the accident analyses and are consistent with the NRC approved 
changes to the surveillances identified for the ISTS of NUREG-1431 
identified in TSTF-110, Revision 2. These changes do not involve a 
change to plant equipment and do not affect the performance of plant 
equipment used to mitigate an accident. Although the deletion of 
these monitor requirements from the TS results in elimination of the 
reduced surveillance interval when the alarm is inoperable (for 
those requirements not being relocated to the LRM) the change in 
frequency is not significant considering the indications available 
to the operator and the relatively slow changes in the parameters 
being monitored during steady state operation. Therefore, based on 
the above, these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Unit 1 SRs 4.1.3.5 and 4.1.3.6 have been additionally modified 
by incorporating the Unit 2 wording which more closely provides a 
surveillance appropriate for the LCO. The LCO requires the shutdown 
rods/control banks to be within the insertion limits and the revised 
SR requires a determination that each shutdown rod/control bank is 
within the insertion limits on a 12-hour frequency. Therefore, the 
revised SRs are consistent with the LCO requirements and more 
clearly provide verification that the LCO is met. This change does 
not affect the operation of the rod position indication system or 
any other system and is consistent with the Unit 2 and ISTS wording. 
This change will not affect the ability of any system to perform its 
design function; therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Changing the surveillance frequency from 24 to 12 hours is more 
conservative and assures the affected parameters are adequately 
monitored. In addition, the change removes monitors from the TSs and 
provides consistency between the SRs, the units and the ISTS. 
Changing the surveillance frequency, correcting the Unit 1 SRs and 
removing reference to the identified monitors from the TS will not 
cause a significant reduction in system reliability nor affect the 
ability of any system to perform its design function. There are no 
hardware changes associated with this license amendment nor are 
there any changes in the method by which any safety-related plant 
system performs its safety function. No new accident scenarios, 
transient precursors, failure mechanisms or limiting single failures 
are introduced as a result of these changes. These changes do not 
introduce any adverse effects or challenges to any safety-related 
systems. No change is required to any system configurations, plant 
equipment or analyses. Therefore, these changes will not create the 
possibility of any new or different kind of accident from any 
accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?

[[Page 4156]]

    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor impact any plant safety analyses since the 
assumptions used will remain unchanged. The safety limits assumed in 
the accident analyses and the design function of the equipment 
required to mitigate the consequences of any postulated accidents 
will not be changed since the proposed changes do not affect the 
accident analyses assumptions or equipment required to mitigate 
design basis accidents described in the UFSAR. Although the deletion 
of these monitor requirements from the TSs results in elimination of 
the reduced surveillance interval when the alarm is inoperable (for 
those requirements not being relocated to the LRM) the effect is not 
significant considering the indications available to the operator 
and the relatively slow changes in the parameters being monitored 
during steady state operation. The TSs continue to assure the 
applicable operating parameters are maintained within the required 
limits. Based on engineering judgement, incorporating these changes 
will not involve a significant reduction in the margin of safety.
    The margin of safety depends upon maintenance of specific 
operating parameters within design limits. The TSs continue to 
require that these limits be maintained and provide appropriate 
remedial actions if a limit is exceeded. The maintenance of these 
limits continues to be assured through performance of the normal 
surveillance at the proposed frequency and the requirements for 
increased monitoring that are relocated to the LRM. Additional 
assurance that the required parameters are adequately monitored is 
provided through other information readily available (i.e., rod 
position indication system, rod bottom alarm and excore neutron 
detectors) that allows the operator to detect a parameter that is 
beginning to deviate from its expected limits and through the 
proposed changes which reduce the normal surveillance interval from 
24 hours to 12 hours to assure the affected parameters are 
adequately monitored. Although these monitors are being removed from 
the TSs, they will continue to be maintained as described in the 
UFSAR (subject to revisions via the 10 CFR 50.59 process). 
Therefore, the plant will be maintained within the analyzed limits 
and the proposed changes will not involve a significant reduction in 
a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: November 24, 1998.
    Description of amendment request: The proposed changes implement 
the consolidated Entergy Operations Quality Assurance Plan Manual 
approved by the NRC on November 6, 1998. The proposed changes also 
clarify the responsibilities of the shift technical advisor position on 
shift, simplify the contents of the monthly operating report 
description in accordance with Generic Letter (GL) 97-02, complete the 
relocation of fire protection requirements from the TS to the fire 
protection program in accordance with GL 88-12, and replace position 
titles with descriptions of functional responsibility in accordance 
with GL 88-06.
    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:

Criterion 1--Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.

    The proposed changes only affect the administrative controls 
contained in Section 6.0 of the Arkansas Nuclear One--Unit 1 (ANO-1) 
and Unit 2 (ANO-2) Technical Specifications (TSs). The proposed 
changes either add additional administrative controls, reduce 
regulatory duplication of requirements consistent with NUREG-1430 
``Standard Technical Specifications--Babcock and Wilcox Plants'' 
dated April 1995, and NUREG-1432 ``Standard Technical 
Specifications--Combustion Engineering Plants'' dated April 1995, or 
revise or relocate administrative controls in accordance with NRC 
guidance. The proposed changes do not affect the operation of any 
structure, system, or component or the assumptions of any accident 
analysis. The details relocated from the ANO-1 and ANO-2 TSs, and 
changes to these details, are controlled under the ANO 10 CFR 50.59 
or 10 CFR 50.54 processes as appropriate.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different 
Kind of Accident from any Previously Evaluated.

    The proposed changes to the ANO-1 and ANO-2 Section 6.0 
administrative controls do not involve a change in the plant design 
or affect the configuration or operation of any structure, system, 
or component.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin 
of Safety.

    The proposed changes to the ANO-1 and ANO-2 TSs affect only 
administrative requirements and do not involve changes to safety 
limits, limiting conditions for operation, or surveillance 
requirements on equipment required to operate the station.
    Therefore, this 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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: November 30, 1998.
    Description of amendment request: The proposed amendment would 
change the CR-3 Improved Technical Specifications (ITS) Section 3.9.3, 
Containment Penetrations. The proposed changes recognize the use of an 
outage equipment hatch (OEH) during refueling operations. The proposed 
changes would also allow both doors in the personnel air locks, and the 
single door in the OEH, to be open during core alterations or movement 
of irradiated fuel assemblies within containment provided certain 
specified conditions are met.
    The licensee stated that the ability to open these doors under 
administrative controls would assist in the maintenance of cleanliness 
and housekeeping, and would provide a safer work environment inside 
containment. In addition, the licensee stated that evacuation of 
personnel could be quickly achieved in the unlikely event of a fuel 
handling accident or other radiological event inside containment, 
reducing the potential for exposures.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.


[[Page 4157]]


    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed change would allow both doors in the personnel air 
locks and the door in the outage equipment hatch (OEH) to remain 
open during core alterations or the movement of irradiated fuel 
inside containment. These doors are normally closed during this 
period in order to prevent the escape of radioactive materials in 
case of a fuel handling accident.
    Operations involving the personnel air locks during refueling 
operations cannot be an initiator of a fuel handling accident or 
other radiological event inside containment. Similarly, operations 
involving the OEH during refueling operations cannot be an initiator 
of a fuel handling accident or other radiological event inside 
containment. The personnel air locks and the OEH are remotely 
located to the fuel handling equipment and cannot affect the 
function of this equipment. The personnel air locks and the OEH are 
not in the immediate vicinity of the reactor vessel and the 
contained irradiated fuel, or any of the paths used for movement of 
irradiated fuel. Additionally, allowing both doors in the personnel 
air locks and the door in the OEH to be open during core alterations 
or the movement of irradiated fuel inside containment cannot create 
the possibility of a fuel handling accident or other radiological 
event inside containment. Therefore, the probability of occurrence 
of any accident previously evaluated is unaffected.
    The approved fuel handling accident analysis does not take 
credit for containment closure. This analysis results in a maximum 
calculated offsite dose well within the limits of 10 CFR 100, and 
the existing analysis as presented in the CR-3 Final Safety Analysis 
Report does not require revision as a result of this proposed 
change. By providing a designated individual readily available to 
close at least one door in the personnel air locks and the door in 
the OEH, containment closure is assured following any required 
evacuation of containment terminating any release of radioactive 
materials outside of the containment. Therefore, the consequences of 
accidents will not be greater than that previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from previously evaluated accidents?
    The operations involving the personnel air locks and the OEH 
cannot be an initiator of any type of accident during refueling 
operations. The personnel air locks and the OEH are passive 
structural features designed to retain structural integrity under 
the expected environmental conditions when installed. Operation of 
the personnel air lock doors and the door in the OEH does not affect 
any safety-related component or structure. Additionally, allowing 
both doors in the personnel air locks and the door in the OEH to be 
open during core alterations or the movement of irradiated fuel 
inside containment cannot initiate any type of accident. Therefore, 
the possibility of a new or different kind of accident occurring as 
a result of this change is not created.
    3. Involve a significant reduction in a margin of safety?
    The margin of safety as defined by 10 CFR 100 has not been 
reduced. The existing approved fuel handling accident analysis does 
not credit containment closure, and remains bounding with both doors 
in the personnel air locks and the door in the OEH open. Closing at 
least one door in the personnel air locks and the door in the OEH 
after evacuation of containment further reduces the offsite doses in 
case of a fuel handling accident, and provides additional margin to 
the calculated offsite doses. Therefore, the existing margin of 
safety will not be reduced.

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

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: November 4, 1998.
    Description of amendment request: The proposed change would revise 
Technical Specifications Surveillance Requirement 4.5.2b.1 to delete 
the prescribed method of venting the Emergency Core Cooling System 
(ECCS) which would allow alternate methods to verify that the ECCS 
piping is full of water. In addition, the associated Bases would be 
expanded to reflect the intent of the surveillance requirement.
    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 does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. The proposed change does not alter or prevent the 
ability of structures, systems and components (SSCs) to perform 
their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR).
    Removal of the prescriptive requirements will not subject the 
ECCS system to conditions adverse to nuclear safety. The proposed 
change does not affect the source term, containment isolation or 
radiological release assumptions used in evaluating the radiological 
consequences of an accident previously evaluated in the Seabrook 
Station UFSAR. The use of proven alternative techniques to verify 
that the ECCS piping is full of water will continue to ensure that 
the ECCS system is capable of performing its intended designed 
safety function. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated and maintained in a state of readiness. Existing 
system and component redundancy is not being changed by the proposed 
change. The proposed change has no adverse affect on component or 
system interactions. The use of proven alternative techniques to 
verify that the ECCS piping is full of water will continue to ensure 
that the ECCS system is capable of performing its intended designed 
safety function. Therefore, since there are no changes to the design 
assumptions, conditions, configuration of the facility, or the 
manner in which the plant is operated and maintained in a state of 
readiness, the proposed change does not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed change does not adversely affect equipment design 
or operation and there are no changes being made to the Technical 
Specification required safety limits or safety system settings that 
would adversely affect plant safety. The proposed change does not 
change the intent of the surveillance requirement of ensuring that 
the system will perform properly, injecting its full capacity into 
the RCS upon demand without subjecting the system to hydraulic 
transients, pump cavitation, and pumping of non-condensable gas 
(e.g., air, nitrogen, or hydrogen) into the reactor vessel following 
a safety injection (SI) signal or during shutdown cooling.
    Thus, it is concluded that the ECCS will continue to be 
available upon demand to mitigate the consequences of an accident 
and, therefore, there is no significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 4158]]

    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: William M. Dean.

Northeast Nuclear Energy Company (NNECO), et al., Docket Nos. 50-245, 
50-336, and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, 
and 3, New London County, Connecticut

    Date of amendment request: December 22, 1998.
    Description of amendment request: The proposed amendment would 
replace specific titles in Section 6.0 of the Technical Specifications 
of all three Millstone units with generic titles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with 10 CFR 50.92, NNECO has reviewed the attached 
proposed changes and ha[s] concluded that they do not involve a 
Significant Hazard Consideration (SHC). The basis for this 
conclusion is that the three criterion of 10 CFR 50.92 are not 
compromised. The proposed change is not a[n] SHC because the 
proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No design basis accidents are affected by these proposed 
changes. The proposed changes are administrative in nature and are 
being proposed to eliminate the need for a Technical Specification 
change each time there is a change in the organization.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no changes in the way the plant is operated due to 
these administrative changes. The potential for an unanalyzed 
accident is not created. There is no impact on plant response, and 
no new failure modes are introduced. The proposed administrative and 
editorial changes have no impact on safety limits or design basis 
accidents, and have no potential to create a new or unanalyzed 
event.
    3. Involve a significant reduction in a margin of safety.
    These changes do not directly affect any protective boundaries 
nor do they impact the safety limits for the protective boundaries. 
These proposed changes are administrative and editorial in nature. 
Therefore there is no reduction in the margin of safety.

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

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

    Date of amendment request: November 13, 1998.
    Description of amendment request: NNECO has determined that the 
increase in radiological consequences, due to changes in the 
assumptions used in the updated dose consequence analysis of the Steam 
Generator Tube Rupture (SGTR) event in the Millstone Unit No. 2 Final 
Safety Analysis Report (FSAR), involves an unreviewed safety question 
(USQ). The changes include a change in High Pressure Safety Injection 
(HPSI) pump runout flowrate, a change in Auxiliary Feedwater Pump (AFW) 
flowrate, a change in the iodine partition factor for the air ejector, 
inclusion of the potential of flashing of the primary-to-secondary 
leakage, and a change in the atmospheric release point assumed 
following actuation of the Enclosure Building Filtration Actuation 
Signal (EBFAS). Therefore, per 10CFR50.59(c), NNECO requested that the 
NRC review and approve the changes to the FSAR through an amendment to 
Operating License DPR-65, pursuant to 10CFR50.90.
    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 10CFR50.92, NNECO has reviewed the proposed 
changes and has concluded that they do not involve a Significant 
Hazards Consideration (SHC). The basis for this conclusion is that 
the three criteria of 10CFR50.92(c) are not compromised. The 
proposed changes do not involve an SHC because the changes would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The FSAR changes reflect changes in the updated SGTR analysis. 
The analysis was updated because of changes in the assumptions used 
in the dose consequence analysis of the SGTR event in Millstone Unit 
No. 2 FSAR. These changes include a change in the iodine partition 
factor for the air ejector, inclusion of the potential of flashing 
of the primary-to-secondary leakage, and a change in the atmospheric 
release point assumed following actuation of the EBFAS. In addition, 
the operator actions associated with Reactor Coolant System (RCS) 
cooldown that are specified in the Emergency Operating Procedures 
have been incorporated, mass releases assuming an RCS cooldown to 
Shutdown Cooling Entry conditions have been used in the dose 
consequence analysis, thyroid doses were calculated using ICRP-30 
dose conversion factors, Iodine releases account for potential 
flashing of the primary-to-secondary leakage, and the Reactor 
Coolant pumps are assumed to be tripped following actuation of a 
safety injection actuation signal. The revised HPSI flowrate is 
higher than that used in the previous analysis. Higher HPSI 
flowrates would increase the primary-to-secondary break flow and, 
thereby, increase the dose consequences. A more conservative iodine 
partition factor for the air ejector has been used along with more 
limiting atmospheric dispersion coefficients as a result of manual 
realignment of the air ejector discharge path to the atmosphere. 
These changes in radiological assumptions are the major reason for 
the increase in calculated dose. The revised AFW flowrate is lower 
than that used in the previous analysis. Lower AFW flowrate would 
tend to increase the steaming required and, thereby, increase the 
dose consequences. The probability that an accident could occur due 
to these changes is not increased since changing the analysis and 
its description can not cause a steam generator tube rupture. 
Therefore, these changes will not significantly increase the 
probability of an accident previously evaluated.
    The dose consequences for the updated SGTR analysis are higher 
than the dose consequences for the previous analysis. However, the 
dose consequences are within the acceptance criteria of SRP 
[Standard Review Plan] 15.6.3 and GDC [General Design Criterion] 19. 
Therefore, these changes will not significantly increase the 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The FSAR changes reflect changes in the updated SGTR analysis. 
The updated analysis does not introduce any new or unanalyzed 
failure modes of equipment or systems, and does not change the 
configuration of the plant. While the updated analysis incorporates 
operator actions that are in accordance with the Emergency Operating 
Procedures, it does not alter the way any structure, system, or 
component functions, and does not alter the manner in which the 
plant is operated. Therefore, there are no new or different types of 
failures of systems or equipment important to safety

[[Page 4159]]

which could cause a new or different type of accident from any 
accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The FSAR changes reflect changes in the updated SGTR analysis. 
The updated analysis shows that the dose consequence acceptance 
criteria are met. The updated analysis incorporates operator actions 
that are in accordance with the Emergency Operating Procedures, and 
credits equipment consistent with its capabilities. Therefore, the 
updated analysis does not reduce the margin of safety. The FSAR 
changes do not alter the acceptance limits of the safety parameters 
of the accident analyses stated in the FSAR. Therefore, these 
changes do not significantly reduce the margin of safety.
    The NRC has provided guidance concerning the application of 
standards in 10CFR50.92 by providing certain examples (March 6, 
1986, 51 FR 7751) of amendments that are considered not likely to 
involve an SHC. The changes proposed herein are covered by example 
(vi) in that the consequences for the updated SGTR analysis are 
higher than dose consequences for the previous analysis. However, 
the dose consequences are within the acceptance criteria of SRP 
15.6.3 and GDC 19.
    As described above, this License Amendment Request does not 
involve a significant increase in the probability of an accident 
previously evaluated, does not involve a significant increase in the 
consequences of an accident previously evaluated, does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated, and does not result in a significant 
reduction in a margin of safety. Therefore, NNECO has concluded that 
the proposed changes do not involve an SHC.

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

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric 
Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: November 20, 1998.
    Description of amendment request: This amendment request updates 
the Emergency Diesel Generator (EDG) day tank volume Surveillance 
Requirement (SR) 3.8.1.4.
    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.
    This proposal does not involve an increase in the probability or 
consequences of an accident previously evaluated. The proposed 
amendment changes EDG day tank volume requirements to reflect the 
[Susquehanna Steam Electric Station] SSES design.
    The safety function of the EDG day tanks is to supply the EDG's 
with enough fuel to ensure the availability of necessary power to 
[engineered safety feature] ESF systems so that fuel, reactor 
coolant and containment system design limits are not exceeded. The 
proposed change increases the minimum diesel fuel oil day tank 
volume for Unit 1 and Unit 2 SR 3.8.1.4 from 325 gallons to 420 
gallons for EDG A-D and 425 gallons for EDG E.
    This volume corresponds to the tank volume at which automatic 
refill occurs. This volume provides for 55 minutes of EDG A-D and 62 
minutes for EDG E operation at continuous rated load conditions.
    Currently, the bases for SR 3.8.1.4 identifies that 
``administrative controls ensure a useable volume of the fuel oil in 
the day tank adequate for approximately 60 minutes of DG operation 
plus 10% at the continuous rated load.'' These administrative 
controls ensure compliance with the Regulatory Guide 1.137 
requirements. Regulatory Guide 1.137 revision 1 endorses American 
National Standards Institute (ANSI) N195-1976. The ANSI N195-1976 
requires each diesel to be equipped with a day tank whose capacity 
is sufficient to maintain at least 60 minutes of operation. This 
capacity is to be based on the fuel consumption at a load of 100% of 
the continuous rating of the diesel plus a minimum margin of 10%.
    These administrative controls on day tank level ensure that the 
required initial fuel oil supply is available to meet the intent of 
the Standard as it applies to the Technical Specification 
surveillance. This Technical Specification change eliminates these 
unnecessary controls needed to conform to the ANSI standard.
    An assessment of the proposed change based on the guidance 
provided in Regulatory Guide 1.174, July 1998, ``An Approach for 
Using Probabilistic Risk Assessment in Risk-Informed Decisions on 
Plant Specific Changes to the Licensing Basis'' concludes that the 
increase in risk is insignificant. It is therefore concluded that 
the proposed changes to SSES Unit 1 and Unit 2 Technical 
Specification SR 3.8.1.4 day tank volume requirements ensures the 
volume is adequate to support the EDG's post accident design basis 
safety function to ensure the availability of necessary power to ESF 
systems so that fuel, reactor coolant system, and containment design 
limits are not exceeded.
    Based upon the above, PP&L concludes that the proposed action 
does not involve an increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed changes does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This proposal does not create the probability of a new or 
different type of accident from any accident previously evaluated. 
The change to the day tank required minimum volume does not change 
any plant systems, structures, or components, nor does the change 
affect any existing or create any new or different kind of accident.
    An assessment of the proposed change based on the guidance 
provided in Regulatory Guide 1.174, July 1998, ``An Approach for 
Using Probabilistic Risk Assessment in Risk-Informed Decisions on 
Plant Specific Changes to the Licensing Basis'' concludes that the 
increase in risk is insignificant. Based on this, it is concluded 
that the proposed changes to SSES Unit 1 and Unit 2 Technical 
Specification SR 3.8.1.4 day tank volume requirements ensures the 
volume is adequate to support the EDG's post accident design basis 
safety function to ensure the availability of necessary power to ESF 
systems so that fuel, reactor coolant system, and containment design 
limits are not exceeded.
    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 the margin of safety.
    An assessment of the proposed change based on the guidance 
provided in Regulatory Guide 1.174, July 1998, ``An Approach for 
Using Probabilistic Risk Assessment in Risk-Informed Decisions on 
Plant Specific Changes to the Licensing Basis'' concludes that the 
increase in risk is insignificant.
    It is concluded that the proposed changes to SSES Unit 1 and 
Unit 2 Technical Specification SR 3.8.1.4 day tank volume 
requirements ensures the volume is adequate to support the EDG's 
post accident design basis safety function to ensure the 
availability of necessary power to ESF systems so that fuel, reactor 
coolant system, and containment design limits are not exceeded.
    Based on this, the proposed changes do not involve a reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.

[[Page 4160]]

    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric 
Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: November 23, 1998.
    Description of amendment request: These amendments would modify the 
Susquehanna Steam Electric Station, Units 1 and 2, Technical 
Specifications (TS) limiting condition for operation (LCO) 3.8.3 and 
surveillance requirement (SR) 3.8.3.1 to increase the minimum fuel oil 
storage tank (FOST) volume ranges. The Bases would be modified to 
reflect that the proposed volumes equal the 7-day fuel oil consumption 
at the continuous emergency diesel generator (EDG) ratings, which are 
greater than design basis analysis (DBA) loads, plus the unusable 
volume in the storage tanks.
    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.
    This proposal does not involve an increase in the probability or 
consequences of an accident previously evaluated. The proposed 
amendment increases FOST volume requirements so to increase the 
margin of safety thus providing further assurance that the EDG FOST 
volume is adequate to support the EDG's post accident design basis 
safety function.
    The safety function of the EDG FOST is to supply the emergency 
diesel generators with enough fuel to ensure the availability of 
necessary power to ESF systems so that fuel, reactor coolant and 
containment system design limits are not exceeded. The current 
Technical specification FOST specified volume is based on the EDG 
post DBA load profile. The proposed FOST volume is based on EDG 
continuos [sic] [continuous] rated load rating which is greater than 
the post DBA load profile providing margin and further assurance 
that the EDG FOST will support the EDG safety function. The proposed 
required FOST volumes are calculated in accordance with ANSI N195-
1976.
    Based upon the above, PP&L concludes that the proposed action 
does not involve an 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.
    This proposal does not create the probability of a new or 
different type of accident from any accident previously evaluated. 
The FOST required minimum values do not change any plant systems, 
structures, or components, nor do they change any existing or create 
any new or different kind of accident. The proposed amendment 
changes FOST volume requirements so to increase the margin of safety 
thus providing further assurance that the EDG FOST volume is 
adequate to support the EDG's post accident design basis safety 
function. 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 the margin of safety.
    The proposed change increases the margin of safety since the 
proposed FOST values are based on the EDG continuos [sic] 
[continuous] rated load ratings which bound the post DBA load 
profile.

    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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: November 6, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications for the Nuclear Instrumentation 
System [NIS] Power Range daily surveillance requirement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed surveillance change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The proposed surveillance change does not significantly increase 
the probability or consequences of an accident previously evaluated 
in the FSAR [Final Safety Analysis Report]. This modification does 
not directly initiate an accident. The consequences of accidents 
previously evaluated in the FSAR are not adversely affected by this 
proposed change because the change to the NIS Power Range channel 
adjustment requirement ensures the conservative response of the 
channel even at part power levels.
    2. Does the proposed surveillance change create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    The proposed surveillance change does not create the possibility 
of a new or different kind of accident than any accident already 
evaluated in the FSAR. No new accident scenarios, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. The proposed Technical Specifications change 
does not challenge the performance or integrity of any safety-
related systems. Therefore, the possibility of a new or different 
kind of accident is not created.
    3. Does the proposed surveillance change involve a significant 
reduction in a margin of safety?
    The proposed surveillance change does not involve a significant 
reduction in a margin of safety. The proposed change does require a 
revision to the criterion for implementation of Power Range channel 
adjustment based on secondary power calorimetric calculation; 
however, the change does not eliminate any RTS [Reactor Trip 
Setpoint] surveillances or alter the frequency of surveillances 
required by the Technical Specifications. The revision to the 
criterion for implementation of the daily surveillance will have a 
conservative effect on the performance of the NIS Power Range 
channel, particularly at part power after normalization at 100% RTP 
[Rated Thermal Power] conditions. The nominal trip setpoints 
specified by the Technical Specifications and the safety analysis 
limits assumed in the transient and accident analysis are unchanged. 
The margin of safety associated with the acceptance criteria for any 
accident is unchanged. Therefore, the proposed change will not 
significantly reduce the margin of safety as defined in the 
Technical Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama.

[[Page 4161]]

    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. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia.

    Date of amendment request: December 4, 1998.
    Description of amendment request: The proposed amendments would 
make two changes to the Technical Specifications (TSs). Change 1 would 
delete the footnote in Hatch Unit 1 TS Section 2.1.1.2 that ties the 
Safety Limit Minimum Critical Power Ratio to Cycle 18. Change 2 would 
delete TS Section 5.6.5.b.2 for Units 1 and 2, and incorporate TS 
Section 5.6.5.b.2 into TS Section 5.6.5.b.1 for both units.
    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:

Basis for Proposed Change 1

    The change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The footnote in Section 2.1.1.2 of the Hatch-1 Technical 
Specifications restricts the applicability of the Safety Limit for 
MCPR [minimum critical power ratio] (SLMCPR) [safety limit minimum 
critical power ratio] to Cycle 18 only. By applying the same NRC-
approved methods used to calculate the Cycle 18 SLMCPR it has been 
determined that the current value is bounding for Cycle 19 as well. 
However, because of the footnote, it [cannot] be applied to Cycle 19 
without a Technical Specifications amendment. In order to eliminate 
future Technical Specifications revisions that do not change the 
SLMCPRs values, SNC [Southern Nuclear Operating Company, Inc.] 
proposes to delete the footnote which ties those values to a 
specific operating cycle. Removing the footnote does not change the 
method of calculating SLMCPR for other cycles, nor does it eliminate 
the requirement to revise the Technical Specifications if a 
different value is used for future cycles. Deletion of the cycle-
specific footnote does not change the operation of any plant 
structure, system or component; therefore, it has no affect on the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    Deleting the cycle-specific footnote in Section 2.1.1.2 of the 
Technical Specifications does not result in any new methods of 
operating the facility and does not involve any facility 
modifications. No new initiating events or transients result from 
this change.
    Therefore, this proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The purpose of the SLMCPR in the Technical Specifications is to 
ensure at least 99.9% of the fuel pins in the core are expected to 
avoid transition boiling during the worst anticipated operational 
occurrence (AOO) throughout an operating cycle. The footnote in 
Section 2.1.1.2 of the Hatch-1 Technical Specifications is intended 
to ensure the correct SLMCPR is used each cycle. Prior to the Spring 
of 1996, the Safety Limits had been calculated for each fuel type, 
independently of operating cycle. As long as the limiting fuel type 
in the core did not change from cycle to cycle, the Safety Limit did 
not change. It was discovered in 1996, however, that generic SLMCPRs 
based on fuel type alone may not be bounding for all cycles for all 
reactors. In response to this discovery GE committed to evaluating 
SLMCPRs based on cycle-unique information as a more accurate method 
of ensuring 99.9% of the fuel pins in the core are expected to avoid 
transition boiling during AOOs. The new methodology, which is now 
applied each cycle, is based on NRC-approved methods and 
incorporates implementing procedures that model cycle-specific 
parameters. This methodology was used to calculate the Cycle 18 
value that is currently in the Technical Specifications. The same 
procedure was also employed to determine that the Hatch-1 Cycle 19 
SLMCPR and it was determined the Cycle 19 value is bounded by the 
Cycle 18 value. Thus, except for the footnote in Section 2.1.1.2, 
there is no need to revise the Hatch-1 Technical Specifications in 
order to ensure the correct SLMCPR is implemented for Cycle 19. As a 
way of avoiding similar changes in the future, SNC proposes that the 
footnote be deleted. Since NRC-approved methodology will still be 
used to determine the cycle-specific SLMCPRs to ensure that [ ] 
99.9% of the fuel rods are expected to avoid transition boiling 
during AOOs, there will be no reduction of margin of safety as a 
result of this change.

Basis for Proposed Change 2

    The change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Section 5.6.5.b.2) no longer describes NRC-approved methods for 
analyzing fuel in the Unit 1 and Unit 2 reactors because the ANF 
[advanced nuclear fuel] LUAs [lead use assemblies] have been 
permanently discharged. Deleting Section 5.6.5.b.2) from the 
Administrative Controls portion of the Technical Specifications does 
not change the operation of any structure, system, or component in 
the facility. Therefore, this amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    Deleting Section 5.6.5.b.2), which describes the use of ANF 
methods for analyzing LUAs, from the Technical Specifications does 
not result in any new methods of operating the facility and does not 
involve any facility modifications. No new initiating events or 
transients result from this change. Therefore, this proposed change 
does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    ANF LUAs are no longer used as fuel in the Plant Hatch reactors, 
therefore, ANF NRC-approved methods described in Technical 
Specifications Section 5.6.5.b.2) are not used to determine power 
distribution limits which appear in the COLR [Core Operating Limit 
Report]. GE's [General Electric's] reload licensing methodology 
described in Section 5.6.5.b.1) will be incorporated into Section 
5.6.5.b. and will continue to be used to analyze the GE fuel in both 
units. Therefore, this 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: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: November 4, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) Sections 4.6.A.1.b and Basis 
3.16 for Units 1 and 2 to revise the start/load time testing and 
ratings for emergency diesel generators (EDGs). The changes will bring 
the TS into conformance with the Updated Final Safety Analysis Report.

[[Page 4162]]

    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:

    Criterion 1--Operation of the Surry Units 1 and 2 in accordance 
with the proposed Technical Specification change does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    The currently specified ``less than 30 seconds'' time to be 
replaced has no specific safety significance or design basis 
regarding EDG starting. The proposed time change to ``less than or 
equal to 10 seconds'' is more conservative and in agreement with 
current accident analysis and surveillance testing. These changes do 
not, in any way, affect the as-built conditions of the plant and do 
not affect the initiators of analyzed events or the assumed 
mitigation of accident or transient events. Analyzed events are 
initiated by the failure of plant structures, systems, or 
components. The proposed changes do not impact the condition or 
performance of these structures, systems or components. Consequences 
of analyzed events are the result of the plant being operated within 
assumed parameters at the onset of any event, and the successful 
functioning of at least one train or division of the equipment 
credited with mitigating the event. There is no impact on the 
capability of the credited equipment to perform, nor is there any 
change in the likelihood that credited equipment will fail to 
perform. As a result, there is no significant increase in the 
probability or consequences of any accident previously evaluated and 
Criterion 1 is, thereby, satisfied.
    Criterion 2--The proposed Technical Specifications change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not involve a physical alteration of 
the plant, or a change in the methods used to operate the plant or 
to respond to plant transients. No new or different equipment is 
being installed and no installed equipment is being removed or 
operated in a different manner. There is no alteration to the 
parameters within which the plant is normally operated or in the 
setpoints, which initiate protective or mitigative actions. 
Consequently, no new failure modes are introduced and the proposed 
changes do not create the possibility of a new or different kind of 
accident from any previously evaluated and Criterion 2 is, thereby 
satisfied.
    Criterion 3--The proposed Technical Specifications change does 
not involve a significant reduction in a margin of safety.
    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 
replacement of the ``less than 30 seconds'' requirement for loading 
the EDGs with the more stringent ``less than or equal to 10 
seconds'' requirement makes no change to the condition or 
performance of equipment or system used in accident mitigation or 
assumed for any accident analysis that could reduce a margin of 
safety as described in the basis for any TS. Therefore, the proposed 
changes do not involve a significant reduction in any margin of 
safety described in the bases for the Technical Specifications and 
Criterion 3 is, thereby, satisfied.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: September 28, 1998 (TSCR 208).
    Description of amendment request: The proposed amendments will 
clarify the notation definition of ``R'' in the Technical 
Specifications (TS) and add a new frequency of ``A.'' The revision of 
``R'' would specify the refueling frequency as 18 months and ``A'' 
would be defined as an annual or 12-month 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. Operation of the Point Beach Nuclear Plant [PBNP] in 
accordance with the proposed amendments will not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    These changes do not involve a significant increase in the 
probability of an accident previously evaluated because no such 
accidents are affected by the proposed revisions to clarify that the 
provisions of TS 15.4.0.2 apply to notation ``R'' in TS Table 
15.4.1-1. The proposed TS changes do not introduce any new accident 
initiators since no accidents previously evaluated have as their 
initiators anything related to the change in the frequency of 
surveillance testing.
    The increased time potential between surveillance frequencies 
does not significantly increase the probability [of] failure of the 
instrumentation contained in TS Table 15.4.1-1. As noted above, 
instrument drift studies concluded that the magnitude of the 
instrument drift (for instrumentation affected by drift) that could 
occur over a 22.5-month interval was bounded by the uncertainty 
allowances used in determining safety system setpoints, and the 
review of historical calibration data concluded that the as-found 
and as-left data has not exceeded acceptable limits for the 
calibration intervals reviewed, except on rare occasions.
    In addition, initiating conditions and assumptions are unchanged 
and remain as previously analyzed for accidents in the PBNP Final 
Safety Analysis Report. The proposed TS changes do not involve any 
physical changes to systems or components, nor do they alter the 
typical manner in which the systems or components are operated. 
Therefore, these changes do not increase the probability of 
previously evaluated accidents.
    These changes do not involve a significant increase in the 
consequences of an accident previously evaluated because the source 
term, containment isolation or radiological releases are not being 
changed by these proposed revisions. Existing system and component 
redundancy and operation is not being changed by these proposed 
changes. The assumptions used in evaluating the radiological 
consequences in the PBNP Final Safety Analysis Report are not 
invalidated; therefore, these changes do not affect the consequences 
of previously evaluated accidents.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    These changes do not introduce nor increase the number of 
failure mechanisms of a new or different type than those previously 
evaluated since there are no physical changes being made to the 
facility. The surveillance test requirements and the way they are 
performed will remain unchanged. The design and design basis of the 
facility remain unchanged. The plant safety analyses remain 
unchanged. Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated is not introduced.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not involve a significant reduction in 
a margin of safety.
    The proposed changes do not involve a significant reduction in 
the margin of safety because existing component redundancy is not 
being changed by these proposed changes. There are no new or 
significant changes to the initial conditions contributing to 
accident severity or consequences, and safety margins established 
through the design and facility license including the Technical 
Specifications remain unchanged. Therefore, there are no significant 
reductions in a margin of safety introduced by [these] proposed 
amendment[s].

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 4163]]

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: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: October 5, 1998 (TSCR 200).
    Description of amendment request: The proposed change modifies 
Technical Specifications Section 15.4.1, ``Operational Safety Review,'' 
by removing the requirement to check environmental monitors on a 
monthly basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Operation of the Point Beach Nuclear Plant [PBNP] in 
accordance with the proposed amendments does not result in a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change eliminates a surveillance requirement for 
environmental monitors. The environmental monitors referred to by 
this surveillance were eliminated from the Radiological 
Environmental Monitoring Program and from the Technical 
Specifications by previous amendments. Therefore, this change is 
administrative in nature in that it corrects a previous 
administrative oversight. The requirement is not related to any 
accident initiator or accident mitigation structures, systems or 
components for any previously evaluated accident. Therefore, no 
increase in the probability or consequences of a previously 
evaluated accident can result.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendment[s] does not create a new or different kind of 
accident from any accident previously evaluated.
    The amendments remove a surveillance requirement from the 
Technical Specifications related to environmental monitors. The 
environmental monitors were removed from the environmental 
monitoring program by previously approved amendments. The 
surveillance requirement is not related to an existing design 
feature of PBNP. Therefore, elimination of the surveillance 
requirement cannot create a new or different kind of accident from 
any accident previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendment[s] does not result in a significant reduction 
in a margin of safety.
    Margins of safety are defined by the safety limits and design 
limits for PBNP. The surveillance is not related to, nor does it 
affect, these limits. Monitoring of the environment continues under 
an approved Radiological Environmental Monitoring Program which 
ensures that any changes in radiation levels in the environs is 
detected, thus ensuring the impact of PBNP operation on the 
environment is minimized. Therefore, the proposed change cannot 
result in a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

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.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1 (WBN), Rhea County, Tennessee

    Date of application for amendment: December 22 and revision dated 
December 23, 1998.
    Brief description of amendment: In order to prevent a potential 
shutdown due to sporadic grounds encountered on an annunciator circuit 
used to confirm operability of an ice condenser inlet door position 
monitoring system, the proposed amendment would provide a temporary, 
optional method of satisfying the requirements for the channel check 
until the next operating Mode, planned in late February 1999, for the 
next refueling outage. Date of publication of individual notice in the 
Federal Register: December 31, 1998 (63 FR 72339).
    Expiration date of individual notice: February 1, 1999.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.

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

[[Page 4164]]

local public document rooms for the particular facilities involved.

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

    Date of application for amendments: November 11, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification Surveillance Requirements (SRs) 3.6.11.6 AND 3.6.11.7, 
regarding the Containment Pressure Control System (CPCS), of the units' 
joint Technical Specifications. The revision brings the SRs into 
conformity with the current design of the CPCS.
    Date of issuance: January 14, 1999.
    Effective date: As of the date of issuance to be implemented 
concurrently with implementation of Amendment Nos. 173 (Unit 1) and 165 
(Unit 2).
    Amendment Nos.: 174--Unit 1; Unit 2--166.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66591). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 14, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of amendment request: September 17, 1998.
    Brief description of amendment: The amendment incorporates the use 
of a range rather then a specific setpoint for the automatic removal of 
the operating bypasses for the core power calculator (CPC) generated 
trips and the high logarithmic power level trip to accommodate the 
design of the plant protection system (PPS) which uses a single 
bistable to control both of these functions.
    Date of issuance: December 31, 1998.
    Effective date: December 31, 1998.
    Amendment No.: 196.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56247).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of application for amendment: June 29, 1998.
    Brief description of amendment: The amendment modifies the TS 
surveillance requirements for SR 4.8.2.3.b.2, SR 4.8.2.3.c.4 and the 
Bases for TS 3.8.2.3 Action b. The licensee is planning to modify the 
120 volt vital alternating current (ac) electrical distribution system 
by installing new inverters during the 2R13 refueling outage. Normally, 
the present inverters for ANO-2 are ac powered and automatically shift 
to direct current (dc) power on a loss of the ac source. The new 
inverters will be powered from the 125 dc system at all times.
    Date of issuance: January 13, 1999.
    Effective date: January 13, 1999, with implementation following 
completion of the required modifications but prior to restart from the 
2R13 outage.
    Amendment No.: 198.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56244).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 13, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie 
Plant, Unit No. 1, St. Lucie County, Florida

    Date of application for amendment: October 29, 1998.
    Brief description of amendment: The amendment revised the 
terminology used in the St. Lucie Plant Technical Specifications (TS) 
relative to the implementation and automatic removal of certain 
protection system trip bypasses to ensure that the meaning of explicit 
terms used in the TS are consistent with the intent of the stated 
requirements.
    Date of Issuance: January 5, 1999.
    Effective Date: As of date of issuance and shall be implemented 
within 30 days of receipt.
    Amendment No.: 159.
    Facility Operating License No. DPR-67: Amendment revised the TS.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66594) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 5, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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 4, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) relating to the condensate storage tank (CST) 
relating to the required minimum water volume and also adds a new TS 
which establishes requirements for the atmospheric steam dump valves 
(ASDVs) to assure their operability. The applicable TS Bases for the 
CST is updated to reflect the proposed changes and a new TS Bases 
section is added to discuss the new TS for the ASDVs.
    Date of issuance: December 31, 1998.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 223.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 26, 1998 (63 FR 
45526).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 31, 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.

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

    Date of amendment request: March 26, 1997, as supplemented by 
letters dated March 18, 1998, and November 17, 1998.
    Brief description of amendment: The amendment revises Technical

[[Page 4165]]

Specifications (TS) 2.1.6 and its associated Basis to restrict the 
number of inoperable main steam safety valves when the reactor is 
critical.
    Date of issuance: December 31, 1998.
    Effective date: December 31, 1998.
    Amendment No.: 189.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38137). The March 18, 1998, and November 17, 1998, supplemental letters 
provided additional clarifying information and did not change the 
original no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated December 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

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

    Date of application for amendment: April 14, 1997, as supplemented 
October 17, 1997, March 20, 1998, May 18, 1998, and August 17, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications to allow for a Safety Review Committee review of plant 
performance as opposed to an audit of plant performance and replaces 
the position title of Vice President Regulatory Affairs and Special 
Projects with Director Regulatory Affairs and Special Projects.
    Date of issuance: December 30, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 186.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45460).
    The October 17, 1997, March 20, 1998, May 18, 1998, and August 17, 
1998, letters provided clarifying information that did not change the 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated December 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: March 22, 1996, as revised and 
supplemented on February 6, 1998, April 17, 1998, and October 30, 1998.
    Brief description of amendment: The amendment provides function-
specific actions and allowed outage times for certain instrumentation, 
and relocates some instrumentation requirements to licensee-controlled 
documents.
    Date of issuance: January 12, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 250.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20855).
    The revision and supplemental information provided on February 6, 
1998, April 17, 1998, and October 30, 1998, provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated January 12, 1999.
    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.

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

    Date of application for amendments: June 30, 1997.
    Brief description of amendments: The amendments delete License 
Condition 2.C(19)b for San Onofre Nuclear Generating Station (SONGS) 
Unit 2 and revises TSs 3.3.1, 3.3.2, 3.3.5, 3.3.10, 3.3.11, 3.4.7, 
3.4.12.1, 3.7.5, 5.5.2.10 and 5.5.2.11 for both SONGS units. These 
changes reinstate provisions of the SONGS Units 2 and 3 TS previously 
revised as part of NRC Amendment Nos. 127 and 116, respectively, make 
corrections to the TS, or remove information inadvertently added to the 
TS that are not applicable to the SONGS units design.
    Date of issuance: December 22, 1998.
    Effective date: December 22, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 2--147; Unit 3--139.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised Facility Operating License No. NPF-10 and the technical 
specifications for both licenses.
    Date of initial notice in Federal Register: March 11, 1998 (63 FR 
11921). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 22, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

    Dated at Rockville, Maryland, this 20th day of January 1999.
    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-1705 Filed 1-26-99; 8:45 am]
BILLING CODE 7590-01-P