[Federal Register Volume 63, Number 18 (Wednesday, January 28, 1998)]
[Notices]
[Pages 4308-4330]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-1904]


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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 Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from January 5, 1998, through January 15, 1998. 
The last biweekly notice was published on January 14, 1998 (63 FR 
2271).

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 27, 1998, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or

[[Page 4309]]

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

Carolina Power & Light Company, Docket No. 50-261, H.B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: December 17, 1997.
    Description of amendment request: The requested amendment revises 
Technical Specification Section 5.6.5, ``Core Operating Limits Report 
(COLR).'' The revisions add reference to an additional approved 
methodology for correlating departure from nucleate boiling (DNB) 
ratios. The added methodology is the Siemens Power Corporation Topical 
Report, EMF-92-153(P)(A), ``HTP: Departure from Nucleate Boiling 
Correlation for High Thermal Performance Fuel.''
    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 change adds a methodology that has been previously 
reviewed and approved by the NRC for determining the DNB safety limit. 
The new methodology utilizes the High Thermal Performance (HTP) 
correlation developed by the fuel manufacturer, Siemens Power 
Corporation. The HTP correlation is empirically based and results in a 
DNB safety limit that corresponds to a 95% probability at a 95% 
confidence level that DNB will not occur. The DNB ratio safety limit is 
a conservative design value which is used as a basis for setting core 
safety limits. The DNB correlation is not assumed to be an initiator of 
analyzed events or transients, and use of the new DNB correlation will 
not alter assumptions relative to mitigation of accident or transient 
events. The proposed change has been confirmed to ensure that no 
previously evaluated accident or transient results in a DNB less than 
the DNB correlation safety limit. The HTP DNB correlation assures with 
high confidence that, for accidents and transients that do not result 
in a DNBR less than the HTP DNBR safety limit, departure from nucleate 
boiling and subsequent fuel overheat will not occur in HTP fuel.
    Therefore, the proposed change does not involve any increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration of 
plant systems, structures, or components or changes in parameters 
governing normal plant operation. The proposed change

[[Page 4310]]

will allow use of the new DNB correlation in like manner as the 
existing DNB correlation in the analysis of accidents and transients to 
assure that the acceptance criteria for current analyses are met. 
Therefore, the proposed change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The proposed change allows use of a DNB correlation that determines 
a safety limit that is slightly lower than the currently used DNB 
correlation. While the slightly lower DNB correlation safety limit 
allows a small increase in margin in analyzing accidents and 
transients, the change from the existing DNB correlation to the 
proposed DNB correlation is not directly comparable to the margin of 
safety. This is because the margin of safety for a particular accident 
or transient is that margin that results from the difference between 
the DNBR calculated for the particular accident or transient using the 
DNB correlation and the DNBR safety limit determined by the DNB 
correlation. Since both the safety limit and the accident or transient 
calculated DNB use the same DNB correlation, the margin of safety is 
consistently calculated and evaluated for acceptability. Since both the 
current and proposed DNB correlation closely approximate test data, and 
they still meet the 95/95 criterion, and the new DNB correlation does 
not result in a DNBR from an accident or transient less than the DNBR 
correlation safety limit, the proposed change does not result in 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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Gordon E. Edison, Acting.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: June 6, 1997, as supplemented September 
25, 1997.
    Description of amendment request: The proposed amendment would 
delete the requirement to sample the spray additive tank per Technical 
Specification (TS) Table 4.1-2, ``Frequency for Sampling Tests,'' and 
delete the sodium hydroxide (NaOH) reference in TS Section 5.2.C.1. The 
request to delete the requirement and the reference was inadvertently 
omitted as part of the licensee's original submittal dated August 22, 
1996, supplemented March 28, 1997, to eliminate the requirement for the 
NaOH containment spray additive and spray additive tank.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
evaluated?
    Response:
    The request to remove the requirement for the spray additive tank 
was approved as part of Amendment No. 191 to Operating License No. DPR 
26. By letter dated April 23, 1997, the Commission reviewed and 
approved the amendment request. However, Consolidated Edison failed to 
include the deletion of the requirement to sample the spray additive 
tank. The removal of the requirement for the spray additive tank has 
been analyzed and approved; therefore, there is no further basis for 
continued testing of the tank. Further, the deletion of the requirement 
would not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (2) Does the proposed license amendment create the possibility of a 
new or different kind of accident from any previously evaluated?
    Response:
    The proposed changes allow the containment safeguards to mitigate 
the consequences of a design basis LOCA [loss-of-coolant accident] in a 
manner equivalent to that previously approved. Therefore, the proposed 
changes do not create an accident or malfunction of safety equipment of 
a different type.
    (3) Does the proposed amendment involve a significant reduction in 
margin of safety?
    Response:
    With the proposed changes, all of the safety criteria previously 
evaluated are still valid and remain conservative. Therefore, the 
proposed amendment does not involve a significant reduction in the 
margin of safety.
    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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: S. Singh Bajwa, Director.

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 amendment request: December 17, 1997.
    Description of amendment request: The proposed amendments would 
revise Section 6.9.1.9 of the Technical Specifications (TS) to 
reference updated or recently approved topical reports, which contain 
methodologies used to calculate cycle-specific limits contained in the 
Core Operating Limits Report. These topical reports have all been 
previously approved by the staff under licensing actions separate from 
the current amendment request.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below.
    1. Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes do not involve any modification to 
existing systems, components, operating limits, or operating procedure. 
Therefore, these proposed changes will have no impact on the 
consequences or probabilities of any previously evaluated accidents.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?

[[Page 4311]]

    No. No actual plant equipment or operating procedure will be 
affected by the proposed changes. Hence, no new equipment failure modes 
or accidents from those previously evaluated will be created.
    3. Will the change involve a significant reduction in a margin of 
safety?
    No. Margin of safety is associated with confidence in the design 
and operation of the plant. The proposed changes to the TS do not 
involve any change to plant design or operation. Thus, the margin of 
safety previously analyzed and evaluated is maintained.
    Based on this analysis, 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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.
    Attorney for licensee: Mr. Paul R. Newton, Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28242-0001.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: December 17, 1997.
    Description of amendment request: The proposed amendments would 
revise Section 6.9.1.9 of the Technical Specifications (TS) to 
reference updated or recently approved topical reports, which contain 
methodologies used to calculate cycle-specific limits contained in the 
Core Operating Limits Report. These topical reports have all been 
previously approved by the staff under licensing actions separate from 
the current amendment request.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is 
presented below.
    1. Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes do not involve any modification to 
existing systems, components, operating limits, or operating procedure. 
Therefore, these proposed changes will have no impact on the 
consequences or probabilities of any previously evaluated accidents.
    2. Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No. No actual plant equipment or operating procedure will be 
affected by the proposed changes. Hence, no new equipment failure modes 
or accidents from those previously evaluated will be created.
    3. Will the change involve a significant reduction in a margin of 
safety?
    No. Margin of safety is associated with confidence in the design 
and operation of the plant. The proposed changes to the TS do not 
involve any change to plant design or operation. Thus, the margin of 
safety previously analyzed and evaluated is maintained.
    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: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina.
    Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 
422 South Church Street, Charlotte, North Carolina.

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

    Date of amendment request: September 23, 1997.
    Description of amendment request: The proposed amendment changes 
the Reactor Protective System and Engineering Safety Actuation System 
trip set point and allowable values for steam generator low pressure. 
The proposed amendment also relocates the RPS and ESFAS response time 
tables from the Technical Specifications to the Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of and Accident Previously Evaluated.
    The proposed changes included in this amendment request do not 
affect the accident initiators in any of the accidents previously 
evaluated. The proposed trip setpoints and allowable values for Steam 
Generator Pressure--Low are being reduced by this proposed amendment 
request. This change is necessary to increase the operating margin 
between the full power steam generator pressure and these setpoints. 
The change should reduce the probability of an inadvertent Main Steam 
Isolation Signal (MSIS) from occurring at power since it will increase 
the operating space between the operating pressure and the setpoints. 
Therefore, this amendment request will not increase the probability of 
any accident previously evaluated.
    The secondary system pipe break safety analyses were reanalyzed for 
the Steam Generator Pressure--Low setpoint reduction effort. This 
effort included the removal of unnecessary analysis conservatisms 
resulting in a significant reduction in the associated setpoints. The 
proposed changes do not involve any change to the configuration or 
method of operation of any plant equipment used to mitigate the 
consequences of an accident. The previously evaluated accidents which 
were determined to be impacted by this setpoint change were evaluated 
with no significant increase in the consequences.
    This amendment request contains the relocation of the Reactor 
Protective System (RPS) and Engineered Safety Features Actuation System 
(ESFAS) response time information from the Technical Specifications 
(TS) to the Safety Analysis Report. This proposed change adopts the TS 
``line-item improvement'' as recommended in NRC Generic Letter 93-08, 
``Relocation of Technical Specification Tables of Instrument Response 
Time Limits,'' dated December 29, 1993. The NRC has concluded that 10 
CFR 50.36 does not require the response time tables to be retained in 
TSs and has issued Generic Letter 93-08 as a line item improvement to 
allow their removal. Response time testing will still be required by 
the ANO-2 TS after the relocation of the associated response time 
information in this amendment request. Relocating the response time 
information for the RPS and ESFAS from the TS to the SAR will not alter 
these surveillance requirements. Therefore, the relocated response time 
portion of this amendment request is considered administrative in 
nature and will not affect the probability or consequences of any 
accident previously evaluated.
    Therefore, this change does not involve a significant increase in 
the

[[Page 4312]]

probability or consequences of any accident previously evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from and Previously Evaluated.
    The proposed changes do not involve any physical modifications 
(i.e., new systems, new components, etc.) to the plant. The proposed 
changes do not involve any change to the configuration or method of 
operation of any plant equipment used to mitigate the consequences of 
an accident. The results of the accident reanalyzes suggest no 
different phenomena or plant behavior than previously considered. The 
Steam Generator Pressure Low setpoint change does not create any new or 
different system actuations or interactions than evaluated previously. 
The relocated response time portion of this amendment request is 
considered administrative in nature and is not considered an accident 
initiator. Therefore, this change does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The accidents which were determined to be impacted by the Steam 
Generator Pressure Low setpoint change were evaluated to ensure 
acceptable results are maintained. The instrument error calculations 
supporting the lower Steam Generator Pressure Low setpoint and 
allowable values will ensure the present accident analysis assumptions 
are still maintained. The methodology used to determine the instrument 
loop errors and uncertainties is the same as that used in previous 
amendment requests that have been reviewed and approved by the NRC. 
Based on these evaluations, the proposed changes do not involve a 
significant reduction in a margin of safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the reauested chance does not involve a significant 
hazards consideration.
    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, N.W., Washington, DC 20005-3502.
    NRC Project Director: John Hannon.

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

    Date of amendment request: September 23, 1997.
    Description of amendment request: The proposed amendment reduces 
the minimum primary system flow that is specified in the technical 
specifications to reflect the effects of increased primary system 
resistance caused by steam generator tube plugging.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    Entergy Operations is proposing a change to the Technical 
Specifications for Arkansas Nuclear One--Unit 2 (ANO-2) to accommodate 
a larger number of plugged steam generator tubes. The proposed 
amendment request will revise the Technical Specifications to 
conservatively account for the reduced reactor coolant system (RCS) 
flow effects of plugging up to 30 percent of the tubes in either steam 
generator. This change will reduce the minimum RCS total flow rate from 
120.4 x 106 lbm/hr to 108.4 x 106 lbm/hr until 
the steam generators are replaced. The steam generators are currently 
scheduled for replacement during the fall of the year 2000. After the 
steam generators are replaced, the minimum RCS flow will then return to 
the current value of 120.4 x 106 lbm/hr.
    The tube plugs that are installed in the steam generators are 
passive components by nature. This amendment request does not change 
the type of plugs which may be installed in the steam generators nor 
does it change the criteria for plugging steam generator tubes. 
Reducing the minimum required RCS flow does not change the plant's 
required mode of operation or modify any active component. Therefore, 
this amendment request will not significantly increase the probability 
of the occurrence of a previously evaluated accident.
    The installation of steam generator tube plugs removes the affected 
tube from service thus reducing the heat transfer surface area and 
increasing the steam generator primary side flow resistance. The 
increased flow resistance in the affected steam generator leads to a 
reduction in the RCS flow available for core cooling. The reduced RCS 
flow rate and heat transfer surface area resulted in a change in 
several primary and secondary parameters that required reanalysis. The 
ANO-2 accident reanalyses supporting the additional steam generator 
tube plugging and the reduction in RCS flow have been completed.
    The Design Basis Accidents (DBAs) affected by these changes were 
reanalyzed to determine if the effects of increased steam generator 
tube plugging and the reduced RCS flow could result in exceeding the 
acceptance criteria applicable to each of these events. It was 
determined that the DBA acceptance criteria would not be exceeded as a 
result of increased steam generator tube plugging and reduction in the 
minimum RCS flow rate.
    Based on the results of the analysis, it is concluded that the 
emergency core cooling system design satisfies the acceptance criteria 
of 10 CFR 50.46(b) for a spectrum of small break and large break loss 
of coolant accidents (LOCAs). The specified acceptable fuel design 
limits (SAFDLs) and the RCS pressure boundary limits also are not 
violated. The fuel and core performance were also determined to remain 
within acceptable limits. Primary and secondary system pressures remain 
below their respective pressure limits.
    Analyses and evaluations of the DBAs have been performed 
demonstrating that the NRC acceptance criteria for these events are 
met. The revised analyses and evaluations consider reduced RCS flow, 
increased RCS temperatures, and increased steam generator tube plugging 
conditions. Although the offsite dose during a steam generator tube 
rupture event could increase, the results remain well within 10 CFR 
[Part] 100 limits. Therefore, the consequences of a previously 
evaluated accident are not significantly increased.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated.
    The proposed amendment reduces the minimum RCS total flow to 
account for the effects of steam generator tube plugging. This 
amendment request will not change the modes of operation defined in the 
Technical Specifications. This change does not add any new equipment, 
modify any interfaces with any existing equipment, change the 
equipment's function, or the method of operating the equipment. The 
proposed change does not change plant conditions in a manner which 
could

[[Page 4313]]

affect other plant components. Reactor core, RCS, and steam generator 
parameters remain within appropriate design limits during normal 
operation. The proposed change could not cause any existing equipment 
to become an accident initiator. Therefore, this change does not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    The margins of safety associated with this change are defined in 
the fuel and core related analyses, and in each of the transient and 
accident analyses affected by the reduced RCS flow. An evaluation of 
the affected analyses confirmed that the established acceptance 
criteria for specified acceptable fuel design limits, primary and 
secondary system over-pressurization, and the acceptance criteria for 
the emergency core cooling systems have been satisfied by this license 
amendment request. The evaluation concludes that, when considering the 
proposed Limiting Conditions for Operation for the minimum RCS total 
flow rate, all applicable acceptance criteria limits are met. The 
margins of safety associated with the transient and accident analyses 
affected by this change will not be significantly reduced. Therefore, 
this change does not involve a significant reduction in the margin of 
safety.
    Therefore, based upon the reasoning presented above and the 
previous discussion of the amendment request, Entergy Operations has 
determined that the requested change does not involve a significant 
hazards consideration.
    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, N.W., Washington, DC 20005-3502.
    NRC Project Director: John Hannon.

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

    Date of amendment request: December 10, 1997.
    Description of amendment request: To clarify certain sections of 
the Technical Specifications (TSs) and Bases which have been 
demonstrated to be unclear or conflicting. Administrative changes 
include TS 2.3 Bases, Table 3.1.1.G.1, Table 3.1.1.M.2, Section 4.3.C, 
and Section 6.1.1. Technical changes include Table 3.3.3, note b, 
Section 3.4 Bases, Section 3.8 Bases and Section 4.5 Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    With respect to the administrative changes, they are typical of the 
example I.c.2.e.i in 51 FR 7744 and therefore, they do 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 the margin of safety; in that 
they are purely administrative changes to achieve consistency or 
correct an error in the TS.
    With respect to technical change, Table 3.1.1, note b:
    1. Involve a siginificant increase in the probability or 
consequences of an accident previously evaluated; (or)
    The proposed change would restore the original value of less than 
600 psig. This lower value would not increase the probability of any 
accident as it provides a more conservative level below which 
protection can be bypassed.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated; (or)
    The proposed change would restore the original value of less than 
600 psig. The setpoint of a bypass cannot create a different kind of 
accident, it can only affect the severity.
    3. Involve a significant reduction in a margin of safety; As the 
requested change lowers the bypass setpoint, the margin of safety will 
be increased.
    With respect to Section 3.4 Bases:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated; (or)
    The proposed change to the Bases removes a possible area of 
confusion from the [TS], and updates the Bases to reflect the results 
of newer, approved methodologies. Therefore, no change to any 
probability calculation occurs.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated; (or)
    The proposed change addresses an existing accident (Small Break 
LOCA) and removes outdated and possibly confusing information. 
Therefore, no new or different kind of accident is created.
    3. Involve a significant reduction in a margin of safety;
    The proposed change does not change the way the plant is operated 
or the way design Bases are maintained. It only removes an outdated and 
possibly confusing paragraph from the Bases, therefore, no margin of 
safety is affected.
    With respect to Section 3.8 Bases:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated; (or)
    The Isolation Condenser Radiation Monitors had no impact o[n] the 
operation of any plant system. Additionally, the monitors were not 
relied upon for any post accident evaluations. They were removed from 
the plant using the 10 CFR 50.59 process. As this request updates the 
[TS] Bases to reflect the plant as currently configured, no impact on 
the probability or consequences of any previously evaluated accident is 
possible.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated; (or)
    The Isolation Condenser Radiation Monitors had no impact o[n] the 
operation of any plant system. Additionally, the monitors were not 
relied upon for any post accident evaluations. They were removed from 
the plant using the 10 CFR 50.59 process. As this request updates the 
[TS] Bases to reflect the plant as currently configured, no new or 
different kind of accident is created.
    3. Involve a significant reduction in the margin of safety;
    The Isolation Condenser Radiation Monitors had no impact o[n] the 
operation of any plant system. Additionally, the monitors were not 
relied upon for any post accident evaluations. They were removed from 
the plant using the 10 CFR 50.59 process. As this request updates the 
[TS] Bases to reflect the plant as currently configured, no reduction 
in any margin of safety can occur.
    With respect to Section 4.5 Bases:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated; (or)
    No change to any procedure, nor any modification to any system is 
requested. The same surveillance will be performed at the same 
frequency. Only the brand of chemical used to perform the surveillance 
will be affected. As an equivalent chemical will be selected, no 
increase in the probability or consequences of an accident previously 
evaluated can be created.

[[Page 4314]]

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated; (or)
    No change to any procedure, nor any modification to any system is 
requested. The same surveillance will be performed at the same 
frequency. Only the brand of chemical used to perform the surveillance 
will be affected. As an equivalent chemical will be selected, no new or 
different kind of accident previously evaluated can be created.
    3. Involve a significant reduction in the margin of safety;
    No change to any procedure, nor any modification to any system is 
requested. The same surveillance will be performed at the same 
frequency. Only the brand of chemical used to perform the surveillance 
will be affected. As an equivalent chemical will be selected, no margin 
of safety can be affected.
    The 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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ronald B. Eaton.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, 
Linn County, Iowa

    Date of amendment requests: October 3, 1997.
    Description of amendment requests: The proposed amendment would 
revise the Operating License to allow the start of core offload as soon 
as 60 hours after shutdown instead of the 120 hours currently 
specified.
    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 Operating License Amendment will not significantly 
increase the probability or consequences of any previously evaluated 
accidents.
    The proposed change will allow initiation of core offload earlier 
after shutdown than is currently allowed. Thermal-hydraulic analysis 
shows that maximum bulk SFP, local water, and fuel clad temperatures 
will remain within acceptable limits and, in fact, do not exceed those 
previously reviewed and approved for Amendment 195.
    Thermal-hydraulic analysis shows the minimum time to action is 
calculated at 4.5 hours versus 5.5 hours previously reviewed and 
approved for Amendment 195. In the event of a loss of forced cooling 
with cask pit isolation gate failure event, the DAEC will use Emergency 
Service Water (ESW), a Seismic Category I system, to provide makeup to 
the SFP. It is estimated to take no more than 2 hours to provide ESW 
makeup to the SFP, therefore the minimum time to action of 4.5 hours is 
sufficient time to prevent uncovering the fuel in the SFP.
    The DAEC design basis refueling accident, as discussed in Section 
15.10.2 of the Updated Final Safety Analysis Report, assumes a twenty-
four hour decay time before core offload begins. The proposed change 
does not adversely affect that accident analysis.
    Therefore, the proposed change will not result in an increase in 
probability or consequences of an accident previously evaluated.
    2. The proposed changes will not create a new or different kind of 
accident from those previously evaluated.
    Thermal-hydraulic analysis shows that the proposed change will not 
result in maximum bulk SFP, local water, or fuel clad temperatures 
which would initiate bulk pool boiling, challenge fuel rod integrity or 
jeopardize the structural integrity of the pool.
    As stated above, the minimum time to action of 4.5 hours allows 
sufficient time to provide ESW makeup to the SFP. Therefore, this 
change does not create the possibility of a new or different type of 
accident.
    3. The proposed change will not result in a significant reduction 
in any margin of safety.
    This change will not result in maximum bulk SFP, local water, and 
fuel clad temperatures in excess of those previously evaluated and 
accepted per Amendment 195. The thermal-hydraulic analysis for Case C 
does show a reduction in the minimum time to action by one hour. 
However, 4.5 hours does provide sufficient time to provide ESW makeup 
to the SFP as this task is estimated to require no more than 2 hours. 
Furthermore, this change does not result in any change to the Technical 
Specifications. Therefore, this change does not result in a significant 
reduction in a margin of safety.
    Based upon the above, we have determined that the proposed 
amendment will not involve a significant hazards consideration.
    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: December 15, 1997.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 2.1 and 3/4.4.1 to change the 
safety limit minimum critical power ratio (MCPR) for the upcoming fuel 
operating cycle (Cycle 7) from 1.07 to 1.09 for two recirculation loop 
operation and from 1.08 to 1.10 for single loop operation. An obsolete 
footnote in TS 3/4.4.1, which states that ``the MCPR Safety Limit of 
1.07 will be used through the first operating cycle,'' would be 
deleted. The associated Bases 2.1 would be changed to (1) reflect the 
new MCPR values, (2) delete certain details (including Bases Table 
B2.1.2-1, ``Uncertainties Used in the Determination of the Fuel 
Cladding Safety Limit,'' and Bases Table B2.1.2-2, ``Nominal Values of 
Parameters Used in the Statistical Analysis of Fuel Cladding Integrity 
Safety Limit,'') and (3) substitute for the deleted detail a reference 
to General Electric Standard Application for Reactor Fuel (GESTAR II), 
NEDE-24011-P-A, and to the cycle-specific analysis. The TS Index would 
be changed to reflect deletion of Bases Tables B2.1.2-1 and B2.1.2-2.
    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 operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The derivation of the revised Safety Limit MCPR was performed using 
the NRC approved methodology in GESTAR II. The Safety Limit MCPR is a 
TS numerical value that cannot initiate an event. Maintaining 
compliance with this

[[Page 4315]]

limit will assure that 99.9 percent of the fuel rods will not 
experience transition boiling during transient events. The deletion of 
the footnote that is no longer necessary and the revision to the Bases 
information are administrative only. The proposed change does not 
modify any of the accident initiators described in the USAR [Updated 
Safety Analysis Report]. No equipment malfunctions or procedural errors 
are created as a result of this change, therefore, no accidents are 
affected by it. The change does not adversely impact the integrity of 
the fuel cladding, which is the first barrier to the release of 
radioactivity to the environment. The change does not affect the 
operation of any systems necessary to mitigate the radiological 
consequences of an accident or to safely shutdown the plant. Therefore, 
this change will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The Safety Limit MCPR is a TS numerical value designed to prevent 
fuel damage from transition boiling. It cannot create the possibility 
of a transient or accident. The deletion of the footnote that is no 
longer necessary and the revision to the Bases information are 
administrative only. The proposed change does not directly impact the 
operation of any systems or equipment important to safety. The analyses 
show that all fuel licensing acceptance criteria are met. The fuel 
cladding, reactor vessel, and reactor coolant system integrity will be 
maintained. Therefore, this change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The Safety Limit MCPR calculation was performed using the NRC 
approved methodology in GESTAR II. Analyses of limiting USAR transients 
establish Operating Limit MCPR values that ensure that the Safety Limit 
MCPR is not violated. The revised cycle specific Safety Limit MCPR 
preserves the existing margin of safety and will continue to assure 
that 99.9 percent of the fuel rods will not experience transition 
boiling during transient events. The deletion of the footnote that is 
no longer necessary and the revision to the Bases information are 
administrative only. Thus, the margin of safety to fuel cladding 
failure due to insufficient cladding heat transfer during transient 
events is not reduced. Therefore, this change will not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: November 13, 1997.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to (1) modify the low 
temperature overpressure protection (LTOP) requirements; (2) modify the 
reactor coolant system (RCS) heatup and cooldown limits; and (3) make 
changes to correct various items based on the licensee's review of the 
current TSs. The supporting TS Bases sections would also be changed to 
reflect the proposed TS changes.
    The affected TSs are: TS 3.1.2.1, ``Flow Paths--Shutdown;'' TS 
3.1.2.2, ``Flow Paths--Operating;'' TS 3.1.2.3, ``Charging Pump--
Shutdown;'' TS 3.1.2.4, ``Charging Pumps--Operating;'' TS 3.1.2.5, 
``Boric Acid Pumps--Shutdown;'' TS 3.1.2.6, ``Boric Acid Pumps--
Operating;'' TS 3.1.2.8, ``Borated Water Sources--Operating;'' TS 
3.4.1.3, ``Coolant Loops and Coolant Circulation--Shutdown;'' TS 3.4.3, 
``Relief Valves;'' TS 3.4.9.1, ``Reactor Coolant System;'' TS 3.4.9.2, 
``Pressurizer;'' TS 3.4.9.3, ``Overpressure Protection Systems;'' TS 
3.5.3, ``ECCS Subsystems--Tavg < 300  deg.F;'' and TS 
3.10.3, ``Pressure/Temperature Limitation--Reactor Criticality.''
    The November 13, 1997, submittal provides specific details related 
to each of the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Each of the proposed changes have been grouped together, as 
appropriate, to address this criteria.
    HPSI Pump Not Required To Be Operable In Modes 5 and 6. The 
proposed change to only require one charging pump to be operable in 
Modes 5 and 6, instead of the current requirement for one charging pump 
and one high pressure safety injection pump (HPSI) pump to be operable, 
will result in sufficient, but not excessive, Reactor Coolant System 
(RCS) makeup capability. When the plant is in Mode 5 or 6 there are two 
major factors to consider with respect to the number of RCS makeup 
pumps required to be operable. If too many RCS makeup pumps are 
required, an inadvertent start of these pumps can result in a mass 
addition transient beyond the capacity of the Low Temperature 
Overpressure Protection (LTOP) System. This may result in an RCS 
pressure increase that exceeds the 10CFR50 Appendix G pressure/
temperature limits. Compliance with the mass input and venting 
restrictions contained in the proposed Technical Specification 3.4.9.3 
will ensure the Appendix G limits are not exceeded.
    The minimum number of RCS makeup pumps required to be operable in 
Modes 5 and 6 ensures sufficient makeup capability is available for RCS 
inventory control and RCS boration requirements. RCS inventory control 
is necessary in Modes 5 and 6 to ensure sufficient water is available 
for core cooling. A rapid loss of RCS inventory due to catastrophic 
pipe failures is unlikely in Modes 5 and 6 due to the reduced RCS 
pressure and temperature. An inventory loss is more likely to occur due 
to small system component failures or during infrequently performed 
evolutions, such as reduced inventory operation. This type of inventory 
loss will occur at a slower rate. Plant operators will have time to 
perform the necessary actions to mitigate the event. Reliance on 
automatic operation of the Emergency Core Cooling System is not 
necessary and Technical Specifications do not require automatic 
actuation by the Engineered Safety Features Actuation System to be 
operable in Mode 4 or below. Operator action is sufficient to mitigate 
a loss of RCS inventory in Mode 4 or below, provided sufficient

[[Page 4316]]

RCS makeup capability is available. Plant procedures and shutdown risk 
management will provide adequate administrative control to ensure 
sufficient RCS makeup capability is available, or that contingency 
plans have been developed.
    The minimum number of RCS makeup pumps required to be operable in 
Modes 5 and 6 ensures sufficient makeup capability is available for RCS 
boration requirements. The RCS is required to be borated to a 
sufficient boron concentration to ensure the Technical Specification 
Shutdown Margin (SDM) requirements are met. The appropriate SDM 
requirements must be met before entry is allowed into Mode 5 or 6. RCS 
boron concentration is increased to establish the required SDM. This is 
normally accomplished by adding borated water to the RCS during plant 
cooldown to compensate for the contraction of the RCS inventory. The 
proposed change will restrict the number of pumps available, which will 
increase the time required to adequately borate the RCS. However, the 
change will not affect the ability to add boric acid to the RCS.
    Even though the proposed change will remove the Technical 
Specification requirement for an operable HPSI pump in Modes 5 and 6, 
sufficient RCS makeup capability will be available to meet RCS boration 
and inventory requirements. Therefore, the proposed change will not 
result in a significant increase in the probability or consequences of 
an accident previously evaluated.
    LTOP Mass Input and Vent Size Requirements. The proposed changes to 
the RCS venting requirements currently contained in Technical 
Specification 3.4.9.3, and the RCS makeup requirements that will be 
relocated to Technical Specification 3.4.9.3 are necessary to be 
consistent with the new LTOP analysis. These changes will ensure the 
10CFR50 Appendix G limits are not exceeded.
    The proposed changes to the mass input restrictions will still 
allow two charging pumps and one HPSI pump to be capable of injecting 
into the RCS when the RCS is operating in Mode 4 [less than or equal 
to] 275  deg.F. This combination will be allowed in Mode 5 until RCS 
temperature is [less than or equal to] 190  deg.F. When RCS cold leg 
temperature is at or below 190  deg.F only one charging pump will be 
allowed to be capable of injecting into the RCS. This restriction will 
continue to apply until the RCS is vented through a passive vent 
[greater than or equal to] 2.2 in \2\. If this passive vent size is 
established, two charging pumps and one HPSI pump are allowed to be 
capable of injecting.
    The passive vent required if one or two power operated relief 
valves (PORVs) are inoperable (Technical Specification Action 
Statements (TSASs) a, b, and c) will be changed from 2.8 in \2\ or 1.4 
in \2\ to 2.2 in \2\. A passive vent of 1.4 in \2\ is equivalent to the 
vent area of one PORV. Since the LTOP analysis assumes 2 operable PORVs 
initially, and then one PORV fails to actuate, RCS overpressure 
protection will be ensured by a passive vent of 1.4 in \2\. However, a 
passive vent is established by removing a pressurizer PORV or the 
pressurizer manway, the normal vent path. The value of 2.2 in \2\ is 
the minimum size of vent that will ensure RCS pressure remains [less 
than or equal to] 300 psia, which is more conservative than the 
Appendix G limits. This vent size will also ensure that RCS pressure 
does not exceed the SDC System design pressure. In addition, this is 
the size of vent that will satisfy Technical Specification 3.4.9.1 to 
allow a 50  deg.F/hr cooldown rate below 190  deg.F.
    TSAS d will be added to address excessive pumping capacity. The 
required completion time of ``immediate'' reflects the importance of 
this restriction, and is consistent with current Technical 
Specification requirements (Technical Specification 3.1.2.3 TSAS b and 
Technical Specification 3.5.3 TSAS c) for this situation.
    These proposed changes are all more restrictive than the previous 
requirements, except for allowing 2 charging pumps and one HPSI in Mode 
5 between 200  deg.F and 190  deg.F, and requiring a vent of 2.2 in \2\ 
instead of 2.8 in \2\ when two PORVs are inoperable. However, the 
proposed mass input and venting restrictions are consistent with the 
new LTOP analysis. This analysis has demonstrated that with the 
proposed restrictions the required LTOP system will provide adequate 
protection for RCS overpressurization transients. Therefore, the 
proposed changes will not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    Increase in Technical Specification Applicability. The 
applicability of Technical Specification 3.4.9.3 will be expanded to 
include all of Mode 5, and Mode 6 until the reactor vessel head is 
removed. The current applicability is limited in Modes 5 and 6 to when 
the RCS is not vented through a vent [greater than or equal to] 2.8 in 
\2\. Expanding the applicability will ensure an LTOP System is in 
place, except when RCS pressurization is not possible (reactor vessel 
head removed). This will ensure the 10 CFR 50 Appendix G limits are not 
exceeded.
    The applicability of Technical Specification 3.4.9.1 will be 
expanded. The current applicability is Modes 1 through 5. However, 
concern for non-ductile failure of the reactor vessel and flange 
applies at all times, not just in Modes 1 through 5. Therefore, the 
applicability will be expanded. Increasing the applicability of 
Technical Specification 3.4.9.1 will place additional restrictions on 
the plant. However, these additional restrictions will ensure the 
integrity of the RCS, in particular the reactor pressure vessel, is 
maintained. Therefore, the RCS will continue to function as designed.
    These more restrictive changes will not result in a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    LTOP PORV Setpoint Change. The required PORV actuation setpoint 
will be reduced from [less than or equal to] 450 psig to [less than or 
equal to] 415 psia (400 psig). The 50 psi setpoint reduction (pressure 
units have also been changed to agree with control room indication) 
will cause the PORVs to actuate earlier during an LTOP transient to 
prevent an RCS overpressurization. It is a more restrictive change that 
is consistent with the new LTOP analysis.
    PORV actuation at the proposed setpoint, in combination with the 
proposed mass input restrictions, will ensure the 10 CFR 50 Appendix G 
limits are not exceeded. Therefore, the proposed change will not result 
in a significant increase in the probability or consequences of an 
accident previously evaluated.
    RCP Start Criteria. The requirements to start the first reactor 
coolant pump (RCP), when RCS temperature is [less than or equal to] 275 
 deg.F, will be modified. The new criteria will ensure that starting an 
RCP will not result in an energy addition transient that could exceed 
the capability of the steam bubble in the pressurizer to mitigate the 
event. (No credit for PORV actuation during this energy addition 
transient was assumed in the new LTOP analysis.) This will ensure that 
the 10 CFR 50 Appendix G limits are not exceeded.
    The proposed RCP restrictions are consistent with the new LTOP 
analysis. This analysis has demonstrated that with the proposed 
restrictions the pressurizer will provide adequate protection for RCS 
overpressurization transients. Therefore, the proposed changes will not 
result in a significant increase in the probability or

[[Page 4317]]

consequences of an accident previously evaluated.
    Boron Dilution Analysis. The analysis of the boron dilution event 
contained in the Millstone Unit No. 2 FSAR [Final Safety Analysis 
Report] Section 14.4.6 assumes that dilution flow rate is limited to 88 
gpm in Modes 4, 5, and 6. Since the charging pumps are the assumed 
dilution source, no more than two charging pumps can be injecting for 
this assumption to remain valid. This results in a Technical 
Specification requirement that no more than two charging pumps can be 
operable when the RCS is in Mode 4 or below (< 300  deg.F). This 
requirement will be modified by replacing the word ``operable'' with 
``capable of injecting into the RCS.'' This more accurately addresses 
the boron dilution analysis restriction of limiting the dilution flow 
to two charging pumps since an inoperable pump can still inject into 
the RCS. This change is consistent with the boron dilution accident 
analysis. The boron dilution analysis further assumes that if this 
dilution flow rate restriction is met, there will be sufficient time 
for the operators to recognize and terminate the dilution before a 
complete loss of shutdown margin occurs. Operator action to restore 
shutdown margin by boration is not assumed.
    The proposed changes will not affect the current Technical 
Specification restriction that no more than two charging pumps can be 
capable of injecting into the RCS (operable) when the RCS is below 300 
deg.F. However, no corresponding action statement currently exists in 
Technical Specification 3.1.2.4 to provide guidance if this requirement 
is not met. The addition of the proposed action statement to Technical 
Specification 3.1.2.4 will require immediate action to correct this 
situation. This is consistent with other current Technical 
Specification requirements (Technical Specification 3.1.2.3 TSAS b and 
Technical Specification 3.5.3 TSAS c) that address excessive RCS makeup 
capacity. Therefore, the proposed changes will not result in a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    RCS Pressure/Temperature and Heatup/Cooldown Limit Changes. The 
proposed changes to the heatup and cooldown rates are a result of the 
new analysis of the RCS pressure/temperature and heatup/cooldown 
limits. These changes will provide flexibility during plant heatup and 
cooldown, and especially during equipment manipulations such as 
securing RCPs, swapping shutdown cooling (SDC) heat exchangers, and 
initiating SDC.
    Figure 3.4.2 will be replaced by two curves, Figures 3.4-2a and 
3.4-2b. Each figure will contain the minimum flange boltup temperature 
and the minimum temperature for criticality. The heatup figure (Figure 
3.4-2a) will also contain the inservice leak and hydrostatic testing 
limits. The temperature change limits will be contained in the new 
Table 3.4-2, instead of in the LCO [limiting condition for operation]. 
The new limits will use cold leg temperature instead of average 
temperature to determine when to change rates. There should be little 
difference between these two temperatures, and cold leg indication is 
directly available to the control room operators.
    The proposed curves and rates are based on indicated cold leg 
temperature. This parameter, which is the best available indication of 
reactor vessel downcomer temperature, will normally be monitored by 
using either RCS cold leg temperature indication or SDC return 
temperature. Plant conditions will determine which one is the 
appropriate indication to use. Actual RCS cold leg temperature will be 
used if any RCP is operating or natural circulation is occurring. 
Otherwise, SDC return temperature will be used.
    RCP restrictions, assumed in the development of the heatup and 
cooldown curves, will be added to the curves. Most of the RCP 
restrictions already exist either in other Technical Specifications or 
in plant procedures. Two new RCP restrictions will be added to 
Technical Specifications. The restriction of no more than three RCPs 
until RCS temperature is above 500  deg.F already exists in plant 
procedures, but it will be added to Technical Specifications. The 
restriction of no more than two RCPs when RCS temperature is below 200 
deg.F already exists in Technical Specifications (3.4.1.4). The RCP 
restriction of no RCPs below 150  deg.F during plant cooldown will be 
added to Technical Specifications. This restriction will have no effect 
on plant operations because RCPs will normally be secured when cooling 
down below 150  deg.F to minimize heat input.
    The inservice leak and hydrostatic testing temperature change limit 
currently specified in Technical Specification 3.4.9.1.c will be 
relocated to Table 3.4-2. The wording will be modified (clarification 
only) to specify the limit also applies for one hour prior to the start 
of inservice leak and hydrostatic testing. This is necessary since the 
development of the inservice leak and hydrostatic testing test curve 
assumes isothermal conditions. The wording will also be modified to 
specify the restrictions apply during testing above the heatup curve 
instead of above system design pressure. This type of testing is not 
performed above system design pressure.
    The 50  deg.F/hr cooldown rate and curve will normally be used when 
the RCS is <190  deg.F and an RCS vent of >2.2 in2 has been 
established. This curve and rate may also be used when RCS cold leg 
temperature is below 230  deg.F to demonstrate compliance with Appendix 
G limits when unanticipated temperature excursions occur.
    The current action statements of Technical Specification 3.4.9.1 
will be separated by Mode and will be modified. Similar changes will be 
made to the action statements of Technical Specification 3.4.9.2. A 
time limit of 72 hours will be placed on the performance of the 
engineering evaluation. If this evaluation is not performed in this 
time period, or the evaluation does not allow continued operation, the 
plant will be required to enter Mode 5 ([less than or equal to] 200 
deg.F), instead of the current requirement to be <200  deg.F. This 
slight relaxation will have no significant impact on plant operations 
because plant temperature is not normally maintained at the mode change 
temperature limit.
    The required RCS pressure will be reduced from 500 psia to 300 
psia. This is closer to the actual plant conditions established in Mode 
5. The required RCS pressure will be reduced from 500 psig to 500 psia 
for the Pressurizer, Technical Specification 3.4.9.2. The change in 
units is consistent with plant instrumentation. Establishing a lower 
RCS pressure is more conservative because it will result in less 
pressure stress on either the reactor vessel or the pressurizer.
    In other than Modes 1 through 4, immediate action will be required 
for limit restoration. Violation of these limits is typically more 
severe when the RCS is cold (<200  deg.F), therefore an immediate 
response is appropriate. A time limit of prior to entering Mode 4 will 
be placed on the performance of the engineering evaluation. This will 
prevent plant startup until the evaluation has determined that the RCS 
is acceptable for continued operation.
    The frequency of surveillance requirements 4.4.9.1.c, 4.4.9.2, 
4.10.3.1 will be increased from once per hour to once per 30 minutes. 
This more restrictive change will provide the plant operators with 
earlier indication that a limit may be exceeded, so that action can be 
taken to prevent exceeding the limit. The proposed changes to the RCS

[[Page 4318]]

pressure/temperature limits and temperature change rates are based on 
the new analysis. This analysis uses standard approved methods that 
ensure the margins of safety required by 10 CFR 50, Appendix G are 
maintained. The other changes discussed are more restrictive 
enhancements to Technical Specification requirements. Therefore, the 
proposed changes will not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    Other Changes. The scope of the action statement for Technical 
Specification 3.1.2.6 will be expanded to cover all three flowpaths 
identified in Technical Specification 3.1.2.2.a. The intent of the 
current wording is to address all flowpaths. These minor wording 
changes will meet this intent.
    Clarification will be added to SR 4.1.2.8.d to be consistent with 
SR 4.1.2.7.c. The clarification will allow the boric acid storage tank 
(BAST) temperature to be verified by checking the ambient air 
temperature.
    A note will be added to Technical Specification 3.4.3 TSAS a to 
allow the block valve(s) to be cycled during plant cooldown when the 
block valve(s) is(are) closed due to inoperable PORV(s). The footnote 
will allow the PORV block valve(s) to be cycled during a plant cooldown 
to prevent thermal binding. This will ensure the associated block 
valve(s) can be opened to allow the PORV(s) which is(are) inoperable, 
can be manually cycled if necessary. Therefore, the PORV block valve(s) 
will be able to function as designed.
    The wording of Technical Specification 3.4.3 Action Statement d 
will be revised to state what action should be performed, and to remove 
specific details on how to perform the required action. This does not 
change the intent of the action statement. Therefore, the pressurizer 
PORVs will continue to function as designed.
    An action statement will be added to Technical Specification 
3.4.9.3 to provide an exception to Technical Specification 3.0.4 
requirements. This is necessary to allow a plant cooldown to MODE 5 if 
one or both PORVs are inoperable. MODE 5 conditions may be necessary to 
repair the PORV(s).
    A footnote will be added to Technical Specification 3.5.3 to allow 
entry into Mode 4 without an operable high pressure safety injection 
pump. This new footnote will allow the plant to enter Mode 4 where this 
specification is applicable without any operable HPSI pumps. However, 
this condition will only be allowed for a very short time period, one 
hour. The proposed change to Technical Specification 3.4.9.3 will allow 
a HPSI pump to be operable above 190  deg.F. However, the 10  deg.F 
range before Mode 4 is reached may not allow sufficient time to ensure 
a HPSI pump is operable. Adding this note will provide the operating 
crew sufficient time to make an orderly transition into Mode 4. This 
condition will only be allowed for one hour, which is the same time 
allowed by the first part of TSAS a for an inoperable HPSI pump.
    The LTOP requirements currently contained in Technical 
Specifications 3.1.2.3 and 3.5.3 will be relocated to the LTOP 
Technical Specification 3.4.9.3. Relocating requirements within 
Technical Specifications will not change the technical content of the 
requirement.
    Various redundant or outdated Technical Specification requirements 
will be eliminated and references will be adjusted to reflect the 
proposed changes. Removal of redundant or outdated requirements from 
Technical Specifications and adjustments to references to other 
requirements will not impact any technical requirements.
    Minor wording changes have been made to many of the Technical 
Specifications contained in this license amendment request. These 
changes do not change any technical aspect of the Technical 
Specification affected. They are editorial changes only.
    The proposed changes do not alter the way any structure, system, or 
component functions. There will be no effect on equipment important to 
safety. Therefore, the proposed changes will not result in a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes have no effect on any of the design basis 
accidents previously evaluated. Therefore, the license amendment 
request does not impact the probability of an accident previously 
evaluated nor does it involve a significant increase in the 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes will not alter the plant configuration (no new 
or different type of equipment will be installed) or require any new or 
unusual operator actions. They do not alter the way any structure, 
system, or component functions and do not alter the manner in which the 
plant is operated. The proposed changes do not introduce any new 
failure modes. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will modify the LTOP requirements, RCS 
pressure/temperature limits, and the RCS heatup and cooldown limits. 
The majority of the proposed changes are being made as a result of the 
new pressure/temperature and LTOP analyses performed. The new pressure/
temperature curves and heatup and cooldown rates were developed in 
accordance with the requirements and methods described in 10 CFR 50 
Appendix G and are consistent with the criteria contained in the 
Standard Review Plan Section 5.3.2. The new LTOP mass input and RCP 
starting restrictions and LTOP PORV setpoints are consistent with the 
criteria contained in the Standard Review Plan Section 5.2.2. 
Additional changes have been proposed to correct various items 
identified during the review of the Millstone Unit No. 2 Technical 
Specifications. The proposed changes do not change the requirements to 
maintain RCS pressure and temperature within the requirements defined 
in Technical Specifications. This will ensure the integrity of the 
reactor vessel is maintained during all aspects of plant operation. 
Therefore, there is no significant effect on the probability or 
consequences of any accident previously evaluated and no significant 
impact on offsite doses associated with previously evaluated accidents. 
This License Amendment Request does not result in a reduction of the 
margin of safety as defined in the Bases for the Technical 
Specifications addressed by the proposed changes.
    The NRC has provided guidance concerning the application of 
standards in 10 CFR 50.92 by providing certain examples (March 6, 1986, 
51 FR 7751) of amendments that are considered not likely to involve an 
SHC [significant hazards consideration]. The changes proposed herein to 
correct terminology, numbering, references, and relocating requirements 
within Technical Specifications are enveloped by example (i), a purely 
administrative change to Technical Specifications. The more restrictive 
changes proposed herein that are based on the new analyses performed 
and the more restrictive enhancements are enveloped by example (ii), a 
change that constitutes an additional limitation, restriction, or 
control not presently included in Technical Specifications. All other 
changes proposed herein are not enveloped by a specific example.
    As described above, this License Amendment Request does not impact 
the probability of an accident previously evaluated, does not involve a 
significant

[[Page 4319]]

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 [Northeast Nuclear 
Energy Company] 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 Deputy Director: Phillip F. McKee.

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

    Date of amendment request: December 8, 1997.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to resolve several compliance 
issues. The proposed changes would (1) correct the wording and the 
formula in TS Definition 1.18 ``Azimuthal Power Tilt--Tq;'' 
(2) correct the wording in TS 4.1.1.1.2 ``Reactivity Control Systems 
Shutdown Margin--Tavg [less than or equal to] 200  deg.F;'' 
(3) correct the mode applicability from Mode 3 to Modes 1 and 2 in TS 
3.1.3.4 ``Reactivity Control Systems--Rod Drop Time;'' (4) correct the 
terminology used to refer to the power dependent insertion limit alarm 
in TS 4.1.3.6 ``Reactivity Control Systems--Regulating CEA [Control 
Element Assembly] Insertion Limits;'' (5) add a footnote for Mode 4 
operability requirement clarification to TS 3.5.3 ``Emergency Core 
Cooling Systems, ECCS Subsystems--Tavg <300  deg.F;'' (6) 
correct the wording, frequency, and reference number for the 
surveillance requirements in TS 3.6.3.2 ``Containment Systems 
Containment Ventilation System;'' (7) correct the nomenclature used for 
the A.C. busses in TSs 3.8.2.1 and 3.8.2.1A ``Onsite Power Distribution 
Systems A.C. Distribution--Operating;'' (8) correct TS Bases by 
modifing the applicable sections to reflect the proposed changes; (9) 
delete the word ``original'' from the statement ``original design 
provision'' in Design Features Section--TSs 5.1.3 ``Flood Control,'' 
5.2.3 ``Penetrations,'' 5.3.2 ``Control Element Assemblies,'' and 5.7.1 
``Seismic Classification;'' and (10) delete Design Features Section--TS 
5.9 ``Shoreline Protection.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change in the wording and associated formula of 
Technical Specifications Definition 1.18 will ensure the calculated 
value of Azimuthal Power Tilt (Tq) used to verify compliance 
with Technical Specification 3.2.4 is associated with the quadrant of 
highest power production with respect to the average of the four 
quadrants, instead of the quadrant that deviates the most (increases or 
decreases) from the average of the four quadrants. This is consistent 
with the method by which power distribution factors are calculated and 
applied in the accident analysis and how the Core Power Distribution 
Monitoring System calculates Tq. The proposed change will 
not alter the way Tq is calculated by the Core Power 
Distribution Monitoring System, nor will it alter any of the power 
distribution assumptions used in the accident analysis. Therefore, this 
change will not significantly increase the probability or consequences 
of an accident previously evaluated.
    Surveillance Requirement (SR) 4.1.1.1.2 requires that the 
difference between predicted and measured core reactivity values be 
maintained within [plus or minus] 1.0% [delta]k/k, and that an 
adjustment be made between the measured and predicted core reactivity 
conditions prior to exceeding 60 EFPD [effective full power days] 
following a refueling outage. The proposed change will not affect the 
requirement to maintain predicted and measured core reactivity values 
within [plus or minus] 1.0% [delta]k/k. However, it will no longer be 
necessary to make an adjustment prior to exceeding 60 EFPD provided the 
[plus or minus] 1.0% [delta]k/k requirement is met. Historically, this 
difference has been small at Millstone Unit No. 2 (less than 
approximately [plus or minus] 1.0% [delta]k/k) and an adjustment has 
not been necessary to ensure the [plus or minus] 1.0% [delta]k/k 
requirement is met. The fact that no adjustment (normalization) will be 
necessary when reactivity differences are small will not affect the 
ability to identify reactivity anomalies. Therefore, this change will 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change to Technical Specification 3.1.3.4 will change 
the applicability from Mode 3 to Modes 1 and 2. This is necessary to 
allow performance of SR 4.1.3.4 at the conditions in the accident 
analysis, and also specified in the [Limiting] Condition [for] 
Operation (LCO). CEA [Control Element Assembly] drop time is important 
for the mitigation of accidents that are initiated while the reactor is 
critical. To ensure the CEA drop time assumed in the accident analysis 
is valid, it is necessary to verify CEA drop time with plant conditions 
consistent with those expected when the reactor is critical. This 
proposed change will allow this verification, and thereby ensure the 
CEAs will function as designed to mitigate design basis accidents. 
Therefore, this change will not significantly increase the probability 
or consequences of an accident previously evaluated.
    The proposed change to SR 4.1.3.6 will modify the terminology used 
to refer to the power dependent insertion limit (PDIL) alarm to agree 
with plant terminology. This change will not alter equipment operation 
or any technical aspect of the SR. The information added to the Bases 
will specify what equipment provides the PDIL alarm. These changes will 
eliminate any confusion with alarm terminology. Therefore, this change 
will not significantly increase the probability or consequences of an 
accident previously evaluated.
    Technical Specification 3.5.3 requires an operable flowpath capable 
of taking a suction from the refueling water storage tank (RWST) on a 
safety injection actuation signal (SIAS), and automatically 
transferring suction to the containment sump on a sump recirculation 
actuation signal (SRAS) in Mode 4. In Mode 4, the automatic SIAS 
generated by low pressurizer pressure and high containment pressure, 
and the automatic SRAS generated by low RWST level, are not required to 
be operable. Automatic actuation in Mode 4 is not required because 
adequate time is available for plant operators to evaluate plant 
conditions and respond by manually operating engineered safety

[[Page 4320]]

features components. Since the manual actuation (trip pushbuttons) 
portions of the safety injection and sump recirculation actuation 
signal generation are required to be operable in Mode 4, credit can be 
taken for remote manual operation to generate the SIAS and SRAS which 
will position all components to the required accident position. The 
proposed change to Technical Specification 3.5.3 will add a footnote 
(***) to explain how these requirements are met in Mode 4. This change 
will not reduce operability or surveillance requirements for the 
Emergency Core Cooling System (ECCS) subsystem required to be operable 
by Technical Specification 3.5.3. The ECCS will continue to function as 
designed to mitigate design basis accidents. Therefore, this change 
will not significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change to Technical Specification 3.6.3.2 will revise 
the wording of the LCO and SR by changing ``locked closed'' to ``sealed 
closed,'' and deleting the requirement to be electrically deactivated. 
The action statement will also be revised to reflect these proposed 
changes. These changes will not affect the requirement for the 
containment purge valves to be closed in Modes 1 through 4. Therefore, 
the proposed changes will not significantly increase the probability or 
consequences of an accident previously evaluated.
    The proposed change to SR 4.6.1.7 will change the surveillance 
frequency from ``prior to each reactor startup'' to ``at least once per 
31 days.'' This change, which will require the surveillance to be 
performed more often (assuming a normal plant startup sequence) will 
provide additional assurance that the containment purge valves are 
sealed closed. In addition, this change will ensure consistency between 
the SR and the applicability of this specification, and also with the 
requirements to verify containment integrity in accordance with 
Technical Specification 3.6.1.1. Therefore, the proposed change will 
not significantly increase the probability or consequences of an 
accident previously evaluated.
    The change in numbering of SR 4.6.1.7 to SR 4.6.3.2 is an 
administrative change only. It will not affect any technical aspect of 
the SR. Therefore, the proposed change will not significantly increase 
the probability or consequences of an accident previously evaluated.
    The proposed changes to Technical Specifications 3.8.2.1 and 
3.8.2.1A will modify the nomenclature used to refer to the vital A.C. 
buses to be consistent with the terminology used by Operations 
Department personnel and the nomenclature contained in their 
procedures. These changes will not alter equipment operation or any 
technical aspects of these specifications. These proposed changes are 
administrative changes only. The A.C. buses will continue to function 
as designed to mitigate design basis accidents. Therefore, these 
changes will not significantly increase the probability or consequences 
of an accident previously evaluated.
    The proposed changes to Technical Specifications 5.1.3, 5.2.3, 
5.3.2, and 5.7.1 will remove the word ``original.'' Reference to 
original design is not appropriate since these items can be changed by 
approved processes. However, these changes will still require the items 
addressed by these specifications to be designed and maintained in 
accordance with the Final Safety Analysis Report (FSAR). The proposed 
changes have no affect on the current approved plant design. Therefore, 
these changes will not significantly increase the probability or 
consequences of an accident previously evaluated.
    Technical Specification 5.9 will be deleted. The required 
provisions for shoreline protection have been completed, and this 
Technical Specification is no longer necessary. The removal of this 
outdated specification will not impact any current requirements. 
Therefore, this change will not significantly increase the probability 
or consequences of an accident previously evaluated.
    The proposed changes do not alter how any structure, system, or 
component functions. There will be no effect on equipment important to 
safety. The proposed changes have no effect on any of the design basis 
accidents previously evaluated. Therefore, this License Amendment 
Request does not impact the probability of an accident previously 
evaluated, nor does it involve a significant increase in the 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no new 
or different type of equipment will be installed) or require any new or 
unusual operator actions. They do not alter the way any structure, 
system, or component functions and do not alter the manner in which the 
plant is operated. The proposed changes do not introduce any new 
failure modes. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed change to the definition of Tq will make 
the Technical Specification definition consistent with the approved 
calculation methodology. This will ensure the core power distribution 
is consistent with accident analysis assumptions. The proposed change 
to the wording of SR 4.1.1.1.2 will not affect the acceptance criteria 
of [plus or minus] 1.0% [delta]k/k, which ensures the accident analysis 
accurately reflects core reactivity conditions. The proposed change in 
the applicability of Technical Specification 3.1.3.4 will allow 
verification of CEA drop time at plant conditions assumed in the 
accident analysis. This will ensure the CEAs will function as assumed. 
The proposed change to SR 4.1.3.6 will modify the terminology used to 
refer to the PDIL alarm to agree with plant terminology. This change 
will not alter equipment operation or any technical aspect of the SR. 
Adding the footnote to Technical Specification 3.5.3 will not change 
any technical aspects of this specification. One ECCS subsystem will be 
available for accident mitigation. The proposed change in wording of 
Technical Specification 3.6.3.2 will not affect the requirement for the 
containment purge valves to be closed in Modes 1 through 4. The 
proposed change in the frequency of performance for SR 4.6.1.7 will 
provide greater assurance that the containment purge valves are closed 
to prevent the potential release of radioactive material through these 
penetrations during accident conditions. The proposed changes in 
terminology in Technical Specifications 3.8.2.1 and 3.8.2.1A will not 
change any technical requirements for the equipment covered. The 
equipment will still function as assumed. Modifying the Bases of 
Technical Specifications are necessary to be consistent with the 
proposed changes will not change any requirements of these 
specifications. The modification to Technical Specifications 5.1.3, 
5.2.3, 5.3.2, and 5.7.1 will not affect the requirement to maintain 
these items in accordance with requirements contained in the FSAR. 
Deleting Technical Specification 5.9 will not affect any requirements 
since the requirements contained in this specification have already 
been completed.
    The proposed changes do not affect any of the assumptions used in 
the accident analysis, nor do they affect any operability requirements 
for equipment

[[Page 4321]]

important to plant safety. Therefore, these proposed changes will not 
result in a significant reduction in the margin of safety as defined in 
the Bases for Technical Specifications covered in this License 
Amendment Request.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: December 12, 1997.
    Description of amendment request: The proposed amendment would 
revise the facility Technical Specifications (TSs) regarding normal 
working hours of plant staff to provide for shift duration of 12 hours. 
It would also revise the TSs to maintain existing ``once per shift'' 
surveillance requirements at 8-hour intervals.
    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:
    Does the proposed licensing amendment involve a significant 
increase in the probability or consequences of an accident previously 
evaluated?
    Response:
    Establishing operating personnel work hours at ``a normal 8 to 12 
hour day, nominal 40-hour week'' allows normal plant operations to be 
managed more effectively and does not adversely affect performance of 
operating personnel. Overtime remains controlled by site administrative 
procedures in accordance with NRC Policy Statement on working hours 
(Generic Letter 82-12). If 8 hour shifts are maintained in part or 
whole, then acceptable levels of performance from operating personnel 
is assured through effective control of shift turnovers and plant 
activities. No physical plant modifications are involved and none of 
the precursors of previously evaluated accidents are affected. 
Therefore, this change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Editorial changes clarify sections 6.2.2.6.b. and 6.2.2.6.c. 
without changing the intent or meaning. [...] Changes to sections 
4.5.F.3., 4.5.F.4., 4.5 Bases, and 4.7.A.7.a. do not change the intent 
or meaning of the Technical Specifications, do not change operating 
procedures, and are consistent with surveillance requirements. 
[Therefore, the proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.]
    Does the proposed license amendment create the possibility of a new 
or different kind of accident from any accident previously evaluated?
    Establishing operating personnel work hours at ``a normal 8 to 12 
hour day, nominal 40-hour week'' allows normal plant operation to be 
managed more effectively and does not adversely affect performance of 
operating personnel. If 8 hour shifts are maintained in part or whole, 
then acceptable levels of performance from operating personnel is 
assured through effective control of shift turnovers and plant 
activities. Overtime remains controlled by site administrative 
procedures in accordance with the NRC Policy Statement on working hours 
(Generic Letter 82-12). No physical modification of the plant is 
involved. As such, the change does not introduce any new failure modes 
or conditions that may create a new or different accident. Therefore, 
plant operation in accordance with the proposed amendment will not 
create the possibility of a new or different kind of accident from any 
previously evaluated.
    Editorial changes clarify sections 6.2.2.6.b. and 6.2.2.6.c. 
without changing the intent or meaning. [* * *] Changes to sections 
4.5.F.3., 4.5.F.4., 4.5 BASES, and 4.7.A.7.a. do not change the intent 
or meaning of the Technical Specifications or operating procedures. All 
previously performed functions are being maintained.
    Therefore, the changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Does the proposed amendment involve a significant reduction in a 
margin of safety?
    Establishing operating personnel work hours at ``a normal 8 to 12 
hour day, nominal 40-hour week'' allows normal plant operations to be 
managed more effectively and does not adversely affect performance of 
operating personnel. If 8 hour shifts are maintained in part or whole, 
then acceptable levels of performance from operating personnel is 
assured through effective control of shift turnovers and plant 
activities. Overtime remains controlled by site administrative 
procedures in accordance with the NRC Policy Statement on working hours 
(Generic Letter 82-12). The proposed change involves no physical 
modification of the plant, or alterations to any accident or transient 
analysis. [* * *] Therefore, the change does not involve any 
significant reduction in a margin of safety.
    Editorial changes clarify sections 6.2.2.6.b. and 6.2.2.6.c. 
without changing the intent or meaning. [* * *] Changes to sections 
4.5.F.3., 4.5.F.4., 4.5 BASES, 4.7.A.7.a. do not change the intent or 
meaning of the Technical Specifications or operating procedures.
    All previously performed functions are being maintained. Therefore, 
the changes do not involve any significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: December 19, 1997.
    Description of amendment request: The proposed amendment would 
revise the Hope Creek Generating Station (HCGS) Technical 
Specifications (TS) to incorporate changes that reflect the completion 
of the Salt Drift Monitoring Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 4322]]

issue of no significant hazards consideration, which is presented 
below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The changes, which update the Terrestrial Ecology Monitoring 
Program status, are administrative in nature and in no way affect the 
initial conditions, assumptions, or conclusions of the Hope Creek 
Generating Station accident analyses. In addition, the proposed changes 
would not affect the operation or performance of any equipment assumed 
in the accident analyses. Based on the above information, we conclude 
that the proposed changes would not significantly increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    As previously stated, the proposed changes are administrative in 
nature and in no way impact or alter the configuration or operation of 
the facilities and create no new modes of operation. PSE&G therefore 
concludes that the proposed changes would not create the possibility of 
a new or different kind of accident.
    3. The proposed changes do not involve a significant reduction in a 
margin of safety.
    The changes are administrative in nature and in no way affect plant 
or equipment operation or the accident analysis. PSE&G therefore 
concludes that the proposed changes would not result in a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: November 14, 1997.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TSs) to provide surveillance 
requirements for the service water accumulator vessels. Specifically, 
surveillance requirements are provided for vessel level, pressure and 
temperature, and discharge valve response time. The surveillance 
requirements are included in TS 3/4.6.1.1 and 3/4.6.2.3, and the 
applicable Bases sections are expanded to provide supporting 
information.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes provide surveillance requirements for the 
Service Water [SW] accumulator tank level, pressure and temperature 
parameters and the discharge valve response time test. Supporting 
information is included in the Bases section of the applicable 
technical specifications. The SW accumulator tank and discharge valve 
design has been reviewed and approved by the NRC staff as documented in 
NRC Safety Evaluation Report (SER) dated June 19, 1997. The proposed 
surveillance requirements do not alter the design as reviewed by the 
NRC staff. The addition of tank parameter surveillance requirements to 
the technical specifications does not alter the physical plant 
arrangement or the installed monitoring instrumentation. The proposed 
addition of tank discharge valve response time surveillance 
requirements to the technical specifications does not alter the method 
of performing these surveillance requirements.
    Therefore the proposed changes do not increase the probability of 
an accident. The surveillance requirements provide additional controls 
for ensuring the SW accumulator tank and discharge valves will be 
maintained within the design parameters assumed in the safety analysis. 
This provides added assurance that the accumulator tanks and discharge 
valves will be capable of performing their required design function 
during accident conditions. There is no change to the performance 
requirements of these components in preventing two phase flow 
conditions and water column separation waterhammer vulnerabilities 
identified in GL [Generic Letter] 96-06. Therefore, the proposed 
changes do not involve an increase in the consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes provide surveillance requirements for Service 
Water Accumulator tank level, pressure and temperature and discharge 
valve time response. Supporting information is included in the Bases 
section of the applicable technical specifications. The SW accumulator 
tank and discharge valve design has been reviewed and approved by the 
NRC staff as documented in NRC Safety Evaluation Report (SER) dated 
June 19, 1997. The proposed surveillance requirements do not alter the 
plant configuration. Installed instrumentation will be used to 
accomplish the tank surveillance requirements. The current plant 
installation also provides for completion of the discharge valve 
response time surveillance utilizing test equipment in accordance with 
plant procedures and configurations. Therefore the performance of these 
surveillance requirements does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The Service Water Accumulator Vessels and discharge valves were 
installed to address the Generic Letter 96-06 issues of column 
separation waterhammer and two phase flow in the containment fan coil 
unit (CFCU) piping during an accident involving loss of offsite power. 
This design has been reviewed and approved by the NRC staff as 
documented in NRC Safety Evaluation Report (SER) dated June 19, 1997. 
The proposed surveillance requirements do not alter the design as 
reviewed by the NRC staff. By providing added assurance that these 
components are capable of performing their specified safety function as 
assumed in the safety analysis, the additional surveillance 
requirements assure system operability to further minimize the 
possibility of waterhammer and two phase flow in the CFCU piping during 
accident conditions. The proposal therefore minimizes the possibility 
of a new or different kind of accident from those previously evaluated 
accidents.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The additional surveillances provide added assurance that the 
margin of safety assumed in the containment integrity and containment 
cooling technical specification will be

[[Page 4323]]

maintained. The additional surveillance requirements further ensure 
that in the event the SW accumulator vessels are out of specification 
or the discharge valves do not meet their response time requirements, 
corrective actions will be completed in accordance with the existing 
containment integrity technical specification allowed outage time to 
restore containment integrity. The surveillance requirements further 
ensure that in the event the SW accumulator vessel or discharge valves 
do not meet these requirements, corrective actions will be completed in 
accordance with the containment cooling technical specification allowed 
outage time to restore the full complement of containment fan coil 
units to operability. Since the proposal maintains the margin of safety 
provided in the containment integrity and containment cooling technical 
specification, 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: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

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 amendment requests: September 16, 1997.
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 3.4.13, ``RCS Operational Leakage,'' TS 
5.5.2.11, ``Steam Generator (SG) Tube Surveillance Program,'' and TS 
5.7.2, ``Special Reports.'' The proposed change is to allow steam 
generator tube repair using ASEA Brown Boveri/Combustion Engineering 
(ABB/CE) leak tight sleeving as an alternative steam generator tube 
repair to plugging.
    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 supporting technical evaluation and safety evaluation for the 
ASEA Brown Boveri/Combustion Engineering (ABB/CE) leak tight sleeves 
demonstrate that the sleeve configuration will provide steam generator 
(SG) tube structural and leakage integrity under normal operating and 
accident conditions. The sleeve configurations have been designed and 
analyzed in accordance with the requirements of the ASME Code. 
Mechanical testing has shown that the sleeve and sleeve joints provide 
margin above acceptance limits. Ultrasonic Testing (UT) is used to 
verify the leak tightness of the weld above the tubesheet. Testing has 
demonstrated the leak tightness of the hardroll joint due to the 
reinforcing effect of the tubesheet. Tests have demonstrated that tube 
collapse will not occur due to postulated Loss of Coolant Accident 
(LOCA) loadings.
    A new, more conservative, Technical Specification (TS) SG tube 
leakage rate requirement is introduced by this change. Accident 
analysis assumptions remain unchanged in the event that significant 
leakage does occur from the sleeve joint or that the sleeve assembly 
ruptures. Any leakage through the sleeve assembly is fully bounded by 
the existing SG tube rupture analysis included in the San Onofre 
Nuclear Generating Station (SONGS) Updated Final Safety Analysis 
Report. Reactor coolant flow reduction from sleeving is addressed by a 
ratio of number of tubes sleeved to equal a plugged tube. The proposed 
sleeving repair process does not adversely impact any other previously 
evaluated design basis accidents.
    Therefore, proposed changes do not involve a significant increase 
in the probability or consequences of an accident.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Installation of the sleeves does not introduce any significant 
changes to the plant design basis. The use of a sleeve to span the area 
of degradation of the SG tube restores the structural and leakage 
integrity of the tubing to meet the original design bases. Stress and 
fatigue analysis of the sleeve assembly shows that the requirements of 
the ASME Code are met. Mechanical testing has demonstrated that margin 
exists above the design criteria. Any hypothetical accident as a result 
of any degradation in the sleeved tube would be bounded by the existing 
tube rupture accident analysis.
    Therefore, the operation of the facility in accordance with 
proposed changes does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The use of sleeves to repair degraded SG tubing has been 
demonstrated to maintain the integrity of the tube bundle commensurate 
with the requirements of the ASME Code and draft Regulatory Guide (RG) 
1.121 and to maintain the primary to secondary pressure boundary under 
normal and postulated accident conditions. The safety factors used in 
the verification of the strength of the sleeve assembly are consistent 
with the safety factors in the ASME Boiler and Pressure Vessel Code 
used in SG design. The operational and faulted condition stresses and 
cumulative usage factors are bounded by the ASME Code requirements. The 
sleeve assembly has been verified by testing to prevent both tube 
pullout and significant leakage during normal and postulated accident 
conditions. A test program was conducted to ensure the lower hardrolled 
joint design was leak tight and capable of withstanding the design 
loads. The primary coolant pressure boundary of the sleeve assembly 
will be periodically inspected by Non-Destructive Examination to 
identify sleeve degradation due to operation.
    Installation of the sleeves will decrease the number of tubes which 
must be taken out of service due to plugging. There is a small amount 
of primary coolant flow reduction due to the sleeve for which the 
equivalent sleeve to plug ratio is assigned based on sleeve length. The 
ratio is used to assess the final equivalent plugging percentage as an 
input to other safety analyses. The sleeve maintains the design basis 
requirements for the SG tubing.
    Therefore, operation of the facility with 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison

[[Page 4324]]

Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

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 amendment requests: October 17, 1997.
    Description of amendment requests: The licensee proposes to amend 
the licenses for SONGS Units 2 and 3 to revise the Final Safety 
Analysis Report (FSAR) to permit digital radiation monitor installation 
for both trains supplying the containment purge isolation signal, and 
permit digital radiation monitor installation for both trains supplying 
the control room isolation signal.
    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 is required to permit using digital radiation 
monitors as input to both trains of the Control Room Isolation Signal 
(CRIS), and to both trains of the Containment Purge Isolation Signal 
(CPIS). These changes will allow replacement of the remaining safety 
related obsolete radiation monitor equipment to address spare parts and 
equipment availability issues. The new containment airborne radiation 
digital monitor will have the same basic architecture as the existing 
analog system, and serves to perform the same function. In addition, 
the digital radiation monitors are expected to be more reliable than 
the existing equipment which is of an analog design.
    Furthermore, defense-in-depth equipment is available that either 
provides, or allows for, actions to mitigate the release of offsite and 
Control Room doses to within existing licensing limits based on 
realistic event input assumptions. Analyses show that if ``realistic'' 
input assumptions are utilized and reasonable operator actions are 
allowed, then acceptable dose consequences result both to the general 
public offsite, and to the Control Room operators.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change will permit upgrading the existing analog 
radiation monitors with upgraded digital radiation monitors. 
Replacement of an analog system to a predominantly digital system, uses 
software algorithms to perform the required functions. A satisfactory 
software verification and validation (V&V) report, including continued 
software change control procedures, provides assurance that a software 
common mode failure is not likely.
    In addition, the design, installation, testing, maintenance, and 
operation of the affected equipment will assure that no new or 
different kinds of accidents will be created. The ESFAS radiation 
monitors involved are portions of systems that respond to accidents. 
They can not, by their actions or inactions, create a new or different 
accident from any accident previously evaluated.
    Therefore, the proposed change will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The CRIS and CPIS Radiation Monitor Systems provide an accident 
mitigation function for offsite doses (10 CFR 100) and Control Room 
doses (10 CFR 50 Appendix A, General Design Criteria 19). A change in 
the margin of safety is introduced due to the possibility of a software 
common mode failure in redundant equipment simultaneously affecting 
equipment performing a different function.
    This change is not a significant reduction in the margin of safety, 
however, due to the following:
    (1) A probabilistic risk analysis has determined that the 
availability of the affected radiation monitors, including software, 
should be better than the existing equipment based on industry data to 
date,
    (2) The software V&V and preoperational testing to be performed 
will provide assurance of system operation, and
    (3) The combined occurrence of a software common mode failure that 
simultaneously causes failure of all available ESFAS radiation monitors 
concurrent with a design bases accident is very unlikely.
    In the unlikely event of a software common mode failure that causes 
all ESFAS radiation monitors to be inoperable concurrent with a design 
bases accident, analyses show that if ``realistic'' input assumptions 
are utilized and reasonable operator actions are allowed, then 
acceptable dose consequences result both to the general public offsite, 
and to the Control Room operators.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, Irvine, California 92713.
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

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 18, 1997.
    Description of amendment request: The proposed amendments would 
modify or delete obsolete conditions from the Unit 1 and Unit 2 
Operating Licenses. The changes are editorial or administrative in 
nature.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes either remove or modify provisions in the 
Plant Hatch Unit 1 and Unit 2 Operating Licenses that have been 
completed or are otherwise obsolete. Certain Surveillance Requirements 
(SRs) that were either added or modified at the time of Improved 
Technical Specifications (ITS) implementation were listed in the 
Operating Licenses with a schedule for performance. With the exception 
of Unit 1 SR 3.8.1.18, all SRs are deleted from the Operating Licenses, 
because they have since been performed according to schedule, and will 
henceforth be

[[Page 4325]]

performed in accordance with the Technical Specifications.
    A requirement for submittal of the Unit 1 inservice inspection plan 
for the recirculation and residual heat removal systems' piping is 
deleted due to completion of the activity.
    Two exemptions granted at Unit 2 startup are deleted due to 
completion of the required activities associated with the exemptions. 
These were seismic qualification demonstration for the Unit 2 reactor 
protection system power supply and completion of the long-term BWR 
[boiling water reactor] Owner's Group Mark I containment program.
    A requirement to conduct the Unit 2 Initial Test Program according 
to the requirements in Chapter 14 of the Final Safety Analysis Report 
without major changes is deleted due to completion of the activity. A 
condition relating to environmental protection is deleted from the Unit 
2 Operating License, since it was superseded by the Environmental 
Protection Plan (Nonradiological), Appendix B to the Operating 
Licenses. Attachment 2, Items To Be Completed Prior To Opening Main 
Steam Isolation Valves, is deleted due to completion of the activities.
    The proposed changes discussed above are strictly administrative/
editorial and do not affect the operation or function of any plant 
system, component, or structure. Therefore, the proposed changes do not 
increase the probability of occurrence or the consequences of a 
previously evaluated accident.
    2. The proposed changes do not create the possibility of a new and 
different type of accident from any previously evaluated.
    The proposed administrative/editorial changes do not alter the 
operation of any plant system or equipment and do not introduce a new 
mode of operation. Thus, the proposed changes cannot create a new 
accident initiating mechanism. Therefore, the proposed changes do not 
create the possibility of a new and different type of accident from any 
previously evaluated.
    3. The proposed changes do not involve a significant reduction in 
the margin of safety.
    Since the proposed changes are strictly administrative/editorial 
and do not involve any physical or procedural changes to the plant, the 
margin of safety, as defined in the bases for any Technical 
Specification is not affected by the proposed changes.
    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.

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

    Date of amendment request: December 17, 1997.
    Description of amendment request: The proposed amendment would 
extend the surveillance interval of the containment spray system nozzle 
air flow test from five years to ten years.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. Operation of the facility in accordance with the proposed 
amendment does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not result in any hardware changes. The 
Containment Spray system trains or nozzles are not assumed to be the 
initiators of any analyzed events. Extending the surveillance interval 
for performing the Containment Spray system nozzle air flow test from 
five to ten years does not represent a significant increase in the 
probability of an accident. The Containment Spray system nozzles are 
not precursors to any accident analyses.
    The Containment Spray system trains and nozzles function to 
mitigate the consequences of an analyzed event by providing spray flow 
to containment during an accident. The proposed change still provides 
assurance that the Containment Spray system nozzles will be maintained 
operable due to the passive nature of the design, the materials of 
construction, and the low-stress non-wetted environment. The extension 
of the surveillance interval does not significantly increase the 
probability or consequences of an accident since the nozzle will still 
be OPERABLE between surveillance tests.
    B. Operation of the facility in accordance with the proposed 
amendment does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed change does not necessitate a physical alteration of 
the plant or changes in parameters governing normal plant operation. No 
new or different types of equipment will be installed. The proposed 
change will still ensure Containment Spray system nozzle OPERABILITY is 
adequately maintained.
    C. Operation of the facility in accordance with the proposed 
amendment does not involve a significant reduction in a margin of 
safety.
    The increased interval between the Containment Spray system nozzle 
air flow test is acceptable due to the passive design of the nozzles 
and industry operating experience as detailed in NURG-1366. The 
increased interval is considered acceptable for maintaining nozzle 
OPERABILITY. The Containment Spray system, including the nozzles, will 
continue to provide their required safety function with the increase 
from five to ten years between inspections.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: John Hannon.

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

    Date of amendment request: December 31, 1997.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications 2.1 (Safety Limits), 2.2 (Limiting 
Safety System Settings), and 3/4.2.5 (Departure from Nucleate Boiling 
Parameters) by including alternate operating criteria to allow 
continued plant operation with a reduced measured reactor coolant 
system flow rate, if necessary.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 4326]]

consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
evaluated?
    The affected Reactor Protection System functions will continue to 
provide their current safety function under alternate operating 
criteria for reduced measured Reactor Coolant System flow conditions. 
The OT Delta-T [Overtemperature Delta-T], OP Delta-T [Overpower Delta-
T], and f(Delta-I) [a function of the indicated difference between top 
and bottom detectors of the power-range neutron ion chambers] safety-
analysis reactor trip setpoints have been recalculated to appropriately 
reflect the reduced flow conditions. In doing so, the difference, or 
margins, between the nominal and maximum values of the reference trip 
setpoints (i.e., K1, and K4 for the OT Delta-T and the OP Delta-T 
setpoints, respectively) have been maintained so that the Total 
Allowance remains unchanged and, therefore, the instrument accuracy 
uncertainties are unaffected.
    Furthermore, implementation of the provisions for reduced measure 
Reactor Coolant System flow under alternate operating criteria for the 
South Texas Project Technical Specifications does not increase the 
probability or consequences of an accident previously evaluated in the 
UFSAR [Updated Final Safety Analysis Report]. This change cannot 
directly initiate an accident. The consequences of accidents previously 
evaluated in the UFSAR are unaffected by this proposed change because 
no change to any equipment response or accident mitigation scenario has 
resulted. There are no additional challenges to fission product barrier 
integrity. Therefore, the probability of an accident previously 
evaluated has not been increased.
    (2) Does the proposed license amendment create the possibility of a 
new or different kind of accident from any accident previously 
evaluated?
    No new failure mechanisms or accident scenarios or limiting single 
failures are introduced as a result of this proposed change. Operation 
of the plant will be consistent with that previously modeled. All of 
the accident analyses previously evaluated in the UFSAR for South Texas 
Project Units 1 and 2 have been evaluated to support alternate 
operating condition with a 3 percent reduction in the minimum measured 
Reactor Coolant System flow. The new nominal Reactor Coolant System 
operating conditions supported by these evaluations have been 
determined. Revised Core Thermal Safety Limits have been established 
and will be incorporated into the Technical Specifications for the 3 
percent Reactor Coolant System measured flow reduction; and, the OT 
Delta-T and OP Delta-T setpoints are re-calculated based on the new 
Safety Analysis Limits, appropriate for the reduced flow operation. 
These reactor protection system functions affected by the change in 
operating conditions will, therefore, continue to provide an 
appropriate response equivalent to current safety analysis modeling. 
The proposed Technical Specification amendment does not challenge the 
performance or integrity of safety-related systems. The possibility of 
a new or different kind of accident, therefore, is not created.
    (3) Does the proposed amendment involve a significant reduction in 
a margin of safety?
    The modification will have no effect on the availability, 
operability, or performance of the South Texas Project safety-related 
systems and components. This is based on: the evaluation performed of 
all accidents previously evaluated in the UFSAR for operation of South 
Texas Project Units 1 and 2 at reduced Reactor Coolant System flow 
conditions; establishment of revised Core Thermal Safety Limits that 
are reflected in the proposed Technical Specification applicable for 
the 3 percent Reactor Coolant System flow reduction; and, the 
appropriately re-calculated OT Delta-T and OP Delta-T setpoints, also 
applicable for these reduced flow conditions. Allowing provision for 
these alternate operating criteria does not prevent inspections or 
surveillance required by the Technical Specifications. The margin of 
safety associated with the acceptance criteria for any accident is 
unchanged, and therefore, the proposed modification will not reduce the 
margin of safety as defined in the Bases of the South Texas Project 
Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J.M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: December 23, 1997.
    Description of amendment request: The license amendment request 
proposes changes to technical specification surveillances to remove the 
requirements related to accelerated testing of the standby emergency 
diesel generators, consistent with the recommendations in NRC Generic 
Letter 94-01, ``Removal of Accelerated Testing and Special Reporting 
Requirements for Emergency Diesel Generators.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not significantly increase the probability 
of occurrence of a previously evaluated accident because the standby 
diesel generators (including the High Pressure Core Spray [HPCS] diesel 
generator) are not initiators of previously evaluated accidents. The 
standby diesel generators mitigate the consequences of previously 
evaluated accidents involving a loss of offsite power. The Perry 
Nuclear Power Plant (PNPP) program developed to meet the Maintenance 
Rule (10 CFR 50.65) will continue to ensure the diesel generators 
perform their function when called upon. The change to the surveillance 
frequency does not affect the design of the diesel generators, the 
operational characteristics of the diesel generators, the interfaces 
between the diesel generators and other plant systems, the function, or 
the reliability of the diesel generators. Thus, the diesel generators 
will be capable of performing their accident mitigation function, there 
is no impact to the radiological consequences of any accident analysis, 
and the probability and consequences of previously evaluated accidents 
are not increased by this activity.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed activity involves a change to the frequency for 
specific technical specification surveillance requirements. No physical 
or

[[Page 4327]]

operational changes to the diesel generators or supporting systems are 
made by this activity. Since the proposed changes do not involve a 
change to the plant design or operation and thus no new system 
interactions are created by this change, these changes do not produce 
any parameters or conditions that could contribute to the initiation of 
accidents different from those already evaluated in the Updated Safety 
Analysis Report. The proposed changes only address the methods used to 
ensure diesel generator reliability. Thus, the proposed amendment does 
not create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes involve the methods used to ensure diesel 
generator performance and reliability. No changes, other than to 
frequency, are made to Technical Specification Surveillance 
Requirements 3.8.1.2 and 3.8.1.3. The NRC, in Generic Letter 94-01, has 
acknowledged the acceptability of the use of the Maintenance Rule 
program for the diesel generators to ensure diesel generator 
performance in lieu of accelerated testing. These proposed changes do 
not involve a change to the plant design or operation, and thus do not 
affect the design of the diesel generators, the operational 
characteristics of the diesel generator, the interfaces between the 
diesel generators and other plant systems, or the function or 
reliability of the diesel generators. Because the diesel generator 
performance and reliability will continue to be ensured by the diesel 
generator program to meet the Maintenance Rule, the proposed changes do 
not result in a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Richard P. Savio.

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

    Date of amendment request: December 23, 1997.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 4.4.5, ``Reactor Coolant 
System--Steam Generators--Surveillance Requirements (SRs).'' SR 4.4.5.8 
would be modified to provide flexibility in the scheduling of steam 
generator inspections during refueling outages.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power Station, 
Unit No. 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no change is being made to any 
accident initiator. No previously analyzed accident scenario is 
changed, and initiating conditions and assumptions remain as previously 
analyzed. The proposed change to Technical Specification (TS) 
Surveillance Requirement (SR) 4.4.5.8, to allow performance of required 
visual inspections of the secured internal auxiliary feedwater header, 
header to shroud attachment welds, and the external header thermal 
sleeves during the third period of the ten-year Inservice Inspection 
Interval, does not affect any Updated Safety Analysis Report (USAR) 
accident initiators. These inspections will continue to take place at a 
prescribed time interval scheduled similar to American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI 
components. Therefore, it can be concluded that the proposed change 
does not involve a significant increase in the probability of an 
accident previously evaluated.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed change does not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The proposed change does not 
alter the source term, containment isolation or allowable radiological 
releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
change does not alter the way the plant is operated, and no new or 
different failure modes have been defined for any plant system or 
component important to safety, nor has any limiting single failure been 
identified as a result of the proposed changes.
    These inspections were established to ensure that there are no new 
failure mechanisms resulting from these components. These inspections 
will continue to take place in the third period of each inservice 
inspection interval. No new or different types of failures or accident 
initiators are introduced by the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because visual inspections will be performed on a prescribed frequency 
that is consistent with the schedules established for ASME Code 
components in accordance with ASME Code Section XI.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.

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

    Date of amendment request: December 23, 1997.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 1.0, ``Definitions,'' to 
clarify the meaning of core alteration; would relocate TS Section 3/
4.9.5, ``Refueling Operations--Communications,'' and the associated 
bases to the Technical Requirements Manual; and would add TS Section 
3.0.6 and the associated bases to address the return to service of 
inoperable equipment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards

[[Page 4328]]

consideration, which is presented below:
    The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit Number 1, in accordance with these changes 
would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because the probability of previously 
analyzed accidents is not affected by the criteria in the core 
alteration definition (Technical Specification (TS) 1.12). Nor do these 
changes, the proposed relocation of the refueling communications TS 3/
4.9.5 and Bases to the DBNPS Updated Safety Analysis Report (USAR) 
Technical Requirements Manual (TRM), or the proposed addition of new TS 
3.0.6 and Bases regarding return to service of inoperable equipment, 
affect any accident initiator, or assumption made in any safety 
analysis. The proposed changes are administrative in nature and are 
consistent with NUREG-1430, Revision 1, ``Standard Technical 
Specifications, Babcock and Wilcox Plants,'' dated April 1995, as 
modified by a pending NUREG-1430 change approved by the NRC, Technical 
Specification Task Force (TSTF) Standard Technical Specification Change 
Traveler Number 165.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
affect accident conditions or assumptions used in evaluating the 
radiological consequences of an accident. The proposed changes do not 
significantly alter the source term, containment isolation, or 
allowable radiological releases.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the proposed 
changes do not change the way the plant is operated. No new or 
different types of failures or accident initiators are introduced by 
the proposed changes.
    3. Not involve a significant reduction in a margin of safety 
because no inputs into the calculation of any Technical Specification 
Safety Limit, Limiting Safety System Settings, Technical Specification 
Limiting Condition for Operation, or other previously defined margins 
for any structure, system, or component important to safety are being 
affected by the proposed changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH.
    Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis & 
Brockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Acting Project Director: Richard P. Savio.

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

    Date of amendment request: December 18, 1997.
    Description of amendment request: By letter dated May 15, 1997, the 
licensee submitted a License Termination Plan. The NRC previously 
published a notice dated August 14, 1997, in the Federal Register (62 
FR 43559) advising of receipt of the Plan. The proposed request is for 
a license amendment approving the Plan for the Yankee Nuclear Power 
Station.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. Accident analyses are 
included in the approved Decommissioning Plan and incorporated into the 
FSAR. All decommissioning and fuel storage activities described in the 
License Termination Plan are consistent with those in the approved 
Decommissioning Plan. No systems, structures, or components that could 
initiate or be required to mitigate the consequences of an accident are 
affected by the proposed change in any way not previously evaluated in 
the approved Decommissioning Plan. Therefore, the proposed change is 
administrative in nature and does not involve an 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. Accident analyses are included 
in the approved Decommissioning Plan and are incorporated into the 
FSAR. All decommissioning and fuel storage activities described in the 
License Termination Plan are consistent with those in the approved 
Decommissioning Plan. The proposed change does not affect plant 
systems, structures, or components in any way not previously evaluated 
in the approved Decommissioning Plan, and no new or different failure 
modes will be created. Therefore, the proposed change is administrative 
in nature and does not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety. Approval 
of the License Termination Plan by license amendment is administrative 
in nature since all decommissioning and fuel storage activities 
described in the License Termination Plan are consistent with those in 
the approved Decommissioning Plan. Therefore, the proposed change does 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Greenfield Community College, 
1 College Drive, Greenfield, Massachusetts 01301.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One 
International Place, Boston, Massachusetts 02110-2624.
    NRC Project Director: Seymour H. Weiss.

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

[[Page 4329]]

published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action, see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of application for amendments: October 2, 1997.
    Brief description of amendments: The amendment changes the Calvert 
Cliffs Unit 1 Technical Specification Requirements 4.8.1.1.2.a.5, 
4.8.1.1.2.d.4, and 4.8.1.1.2.d.5. Baltimore Gas and Electric Company is 
planning to modify existing 1B emergency diesel generator (EDG) to 
increase its rated continuous capacity from 2700 kW to 3000 kW by 
increasing the mechanical capacity of the engine. The change revises 
the above surveillance requirements to reflect the new electrical 
capacity of 1B EDG.
    Date of issuance: January 5, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 224.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59913).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated January 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of amendment request: September 25, 1997.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by modifying the Limiting Condition for Operation 
(LCO) 3.6.1.2 (Containment Leakage), the associated action, and 
Surveillance Requirement (SR) 4.6.1.2 for Waterford Steam Electric 
Station, Unit 3 (Waterford 3). The air lock door seal leakage rate 
acceptance criteria in TS 6.15 is being changed from 0.01La 
to 0.005La. TS 6.15 is also being modified to make the terms 
used in the Containment Leakage Rate Testing Program consistent with 
terms used in the TS.
    Date of issuance: January 15, 1998.
    Effective date: January 15, 1998.
    Amendment No.: 138.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54872).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 15, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of application for amendment: October 10, 1997.
    Brief description of amendment: The amendment revises the Oyster 
Creek Nuclear Generating Station (OCNGS) operating license and 
technical specifications to reflect the registered trade name of ``GPU 
Energy'' under which the owner of OCNGS now does business and to 
reflect the change of the legal name of the operator of OCNGS from GPU 
Nuclear Corporation to GPU Nuclear, Inc. In addition, two minor 
editorial corrections associated with the name change are included in 
the amendment.
    Date of issuance: January 14, 1998.
    Effective date: As of the date of issuance, with full 
implementation within 30 days.
    Amendment No.: 194.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59915). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated January 14, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

    Date of application for amendments: October 8, 1997, and October 
21, 1997.
    Brief description of amendments: The amendments increase both the 
minimum required ice mass per ice basket and the total minimum required 
ice mass in the ice condenser, and change the bases for the technical 
specifications.
    Date of issuance: January 2, 1998.
    Effective date: January 2, 1998, with full implementation within 45 
days.
    Amendment Nos.: 220 and 204.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54863).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085

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

    Date of application for amendment: October 15, 1997.
    Brief description of amendment: Technical Specification 
Surveillances 4.1.2.3.1, 4.1.2.4.1, 4.5.2, 4.6.2.1, and 4.6.2.2 require 
the recirculation spray, quench spray, residual heat removal, 
centrifugal charging, and safety injection pumps to be tested on a 
periodic basis and after modifications that alter subsystem flow 
characteristics. The amendment replaces the specific surveillance pump 
pressure with a statement that the test be conducted in accordance with 
Specification 4.0.5, Inservice Testing Program. The

[[Page 4330]]

amendment also decreases the required individual safety injection and 
centrifugal charging pump injection line flow rates, increases the 
allowed individual safety injection pump runout flow rate, and makes 
editorial changes to the surveillances.
    Date of issuance: December 24, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 155.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59918).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 24, 1997.
    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 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.

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

    Date of application for amendments: November 4, 1997.
    Brief description of amendments: These amendments revise Technical 
Specification 3/4.8.1 on the emergency diesel generators to (1) delete 
the 18-month surveillance requirements 4.8.1.1.2.d.1 and (2) eliminate 
the accelerated testing requirement of Table 4.8-1.
    Date of issuance: January 8, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 203 and 185.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 3, 1997 (62 FR 
63982).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendments: November 21, 1997 (TS 97-05).
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) to allow a one-time provision for testing 
power-operated relief valves in Mode 5.
    Date of issuance: January 13, 1998.
    Effective date: January 13, 1998.
    Amendment No.: 230.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the TS.
    Date of initial notice in Federal Register: December 1, 1997 (62 FR 
63565).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

    Dated at Rockville, Maryland, this 21st day of January 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-1904 Filed 1-27-98; 8:45 am]
BILLING CODE 7590-01-P