[Federal Register Volume 64, Number 27 (Wednesday, February 10, 1999)]
[Notices]
[Pages 6692-6720]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-3098]


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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 tendency 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 15, 1999, through January 29, 1999. 
The last biweekly notice was published on January 27, 1999 (64 FR 
4152).

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

[[Page 6693]]

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 March 12, 1999, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspects(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 preset 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

[[Page 6694]]

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.

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: December 7, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) to permit a one-time only 
extension of the steam generator tube inspection interval for fuel 
cycle 14 and delete the requirement to have NRC staff concurrence of 
the steam generator examination program. Specifically, TS 4.13A.2.a 
would be revised with a footnote that states ``Examinations scheduled 
for 1999 only, shall be conducted during the 2000 Refueling Outage 
which will commence no later than June 3, 2000. The scheduled 
examinations will be completed prior to return to service from the 2000 
Refueling Outage.'' In addition, TS 4.13C.1 would be revised to state 
``The proposed steam generator examination program shall be submitted 
for NRC staff review at least 60 days prior to each scheduled 
examination.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 59.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not involve any physical modifications 
to the plant or modification in the methods of plant operation which 
could increase the probability or consequences of previously 
evaluated accidents. The proposed change permits an extension of the 
current steam generator tube inservice inspection cycle. This 
extension would allow the steam generator tube examinations to be 
conducted during the 2000 refueling outage which will commence no 
later than June 3, 2000. The basis for acceptance of this increase 
in the technical specification limit is the ``non-operating'' steam 
generator time between the last examination and the upcoming 
examination. Extending the steam generator ``operating'' duration by 
48 days would not significantly increase wear which might lead to 
tube failure. No appreciable steam generator tube wear or 
degradation is expected as a result of this extension. This change 
will not affect the scope, methodology, acceptance limits and 
corrective measures of the existing steam generator tube examination 
program. The probability and consequences of failure of the steam 
generators due to leaking or degraded tubes is not increased by the 
proposed change. Additionally the proposed administrative change to 
delete the requirement to receive NRC concurrence of the proposed 
steam generator examinations will have no bearing on the actual 
results of the steam generator examinations. Therefore, the 
probability and the consequence of a design basis accident are not 
being increased by the proposed change.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Plant systems and components will not be operated in a different 
manner as a result of the proposed Technical Specification change. 
The proposed change permits the upcoming steam generator tube 
examination to be conducted during the 2000 refueling outage that 
will commence no later than June 3, 2000. There are no plant 
modifications or changes in methods of operation. This extension is 
based upon the ``non-operating'' steam generator time between the 
last examination and the upcoming examination. Extending the steam 
generator ``operating'' duration by an additional 48 days would not 
significantly increase wear which might lead to tube failure. The 
proposed extension will not increase the probability of occurrence 
of a tube rupture, increase the probability or consequences of an 
accident, or create any new accident precursor. Additionally the 
proposed administrative change to delete the requirement to receive 
NRC concurrence of the proposed steam generator examinations will 
have no bearing on the actual results of the steam generator 
examinations. Therefore, the possibility of an accident of a 
different type than was previously evaluated in the safety analysis 
report is not created by the proposed change to the Technical 
Specification.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change to Technical specification section 4.13A.2.a 
will not reduce the margin of safety. This amendment involves an 
extension of the current steam generator tube inservice inspection 
cycle. The basis for acceptance of this increase in the technical 
specification limit is the ``non-operating'' steam generator time 
between the last examination and the upcoming examination. Extending 
the steam generator ``operating'' duration by an additional 48 days 
would not significantly increase wear which might lead to tube 
failure. No appreciable steam generator tube wear or degradation is 
expected as a result of this extension. Additionally the proposed 
administrative change to delete the requirement to receive NRC 
concurrence of the proposed steam generator examinations will have 
no bearing on the actual results of the steam generator 
examinations. Therefore, the accident analysis assumptions for 
design basis accidents are unaffected and the margin of safety is 
not decreased by the proposed Technical Specification change.

    [* * *]
    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: 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.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: April 30, 1998.
    Description of amendment request: The proposed amendment revises 
the definition of quadrant power tilt to clearly allow the use of 
either the incore detectors or the excore detectors for determining 
quadrant power tilt.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--Does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to the quadrant power tilt (QPT) definition 
will not alter any Safety Analysis Report (SAR) assumptions 
established and implemented by the technical specifications. The 
proposed change will allow the use of either the incore detectors or 
the excore power range detectors for determining QPT. This change is 
consistent with the improved Standard Technical Specifications (STS) 
which has been previously approved by the NRC. QPT measured by 
incore detectors provides a more accurate indication of reactor core 
power distribution than the value determined from the excore 
detectors. The accident prevention and mitigation features of the 
plant are not affected by this proposed amendment.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.

[[Page 6695]]

    Criterion 2--Does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The proposed change to the definition of QPT does not alter the 
ANO-1 SAR analysis or core operating limits report (COLR). The 
change will clearly permit the use of either the incore detectors or 
the excore detectors for monitoring QPT. The design and physical 
configuration of the plant are not affected by this change.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--Does not involve a significant reduction in the 
margin of safety.
    The proposed change to the QPT definition incorporates the 
improved TS definition contained in NUREG-1430. The revised 
definition allows the use of either the incore detectors or the 
excore power range detectors for determination of QPT. The change 
does not vary or affect any of the plant's operating parameters. The 
COLR currently specifies acceptable QPT limits based upon the 
measurement techniques. These limits are based upon the unique 
measurement characteristics of the incore and excore power range 
detectors and assure the measurement independent limit is not 
violated.

    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: August 6, 1998.
    Description of amendment request: The proposed amendment revises 
the minimum and the maximum concentration limits for the sodium 
hydroxide tank. The proposed change also revises the minimum specified 
tank volume to refer to the parameter used in the analysis with no 
allowance for instrument uncertainty and deletes the maximum specified 
tank volume.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Criterion 1--Does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Sodium hydroxide is not an accident initiator. It is, however, a 
contributor to the mitigation of the effects of a Loss-of-Coolant-
Accident (LOCA). The proposed change in NaOH tank concentration 
results in changing the expected post-LOCA reactor building sump pH. 
The reduction in the lower value of sump pH, from 8.5 to 7.0, is 
acceptable based on guidance contained in NUREG-0800, Standard 
Review Plan, Section 6.5.2, ``Containment Spray as a Fission Product 
Cleanup System Review Responsibilities,'' Revision 2, December 1988. 
This guidance allows the assumption of long-term iodine retention 
when the equilibrium sump pH, after mixing and dilution with the 
primary coolant and ECCS injection, is above 7.0. Although the 
change allows the volume of the NaOH tank to be maintained at a 
lower volume, the proposed minimum volume bounds the analyses of 
concern.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2--Does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Sodum hydroxide is added for iodine removal and for pH 
adjustment of the borated water in the reactor building sump 
following a LOCA. The proposed changes in NaOH tank concentration 
and volume introduce no new mode of plant operation.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3--Does not involve a significant reduction in the 
margin of safety.
    The proposed change in NaOH tank concentration results in 
changing the expected post-LOCA reactor building sump pH. This 
proposed change does involve an incremental reduction in the margin 
to safety since iodine retention is dependent on the pH of the sump/
spray solution. However, this reduction is not considered 
significant in that the effect of the change in sump pH, from 8.5 to 
7.0 has a relatively minor effect on iodine retention, as supported 
by Standard Review Plan (NUREG-0800), Section 6.5.2, Revision 2, 
dated December 1988. Although the change allows the volume of the 
NaOH tank to be maintained at a lower volume, the proposed minimum 
volume bounds the analyses of concern.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: John N. Hannon.

Entergy Operations, Inc., Et Al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi and Entergy 
Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, 
River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: January 12, 1999, superceding the 
amendment request in the letter of September 30, 1996, for both 
stations.
    Description of amendment request: The proposed amendment would add 
an additional required action to the Limiting Condition for Operation 
(LCO) 3.9.1, ``Refueling Equipment Interlocks,'' of the Technical 
Specifications for both stations. The additional action would allow an 
alternative to the current action for one or more inoperable refueling 
equipment interlocks. The current action is to ``suspend in-vessel fuel 
movement with equipment associated with the inoperable interlock(s).'' 
The alternative action proposed is to (1) insert a control rod 
withdrawal block, and (2) verify all control rods are fully inserted in 
core cells containing one or more fuel assemblies. The proposed 
amendment would also revise the Bases for the LCO 3.9.1 actions to 
describe the proposed alternative actions. The previous Federal 
Register notice of the amendment request in the superceded letter of 
September 30, 1996, was issued on June 16, 1996, (61 FR 31178), for 
Grand Gulf Nuclear Station (GGNS).
    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:

    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The refueling interlocks are explicitly assumed in the GGNS 
Updated Final Safety Analyses Report (UFSAR) and RBS Updated Safety 
Analyses Report (USAR) analysis of the control rod removal error or 
fuel loading error during refueling. This analysis evaluates the 
probability and consequences of control rod withdrawal during 
refueling. Criticality and, therefore, subsequent prompt reactivity 
excursions are prevented during

[[Page 6696]]

the insertion of fuel, provided all required control rods are fully 
inserted during the fuel insertion. The refueling interlocks 
accomplish this by preventing loading fuel into the core with any 
control rod withdrawn, or by preventing withdrawal of a rod from the 
core during fuel loading.
    When the refueling interlocks are inoperable the current method 
of preventing the insertion of fuel when a control rod is withdrawn 
is to prevent fuel movement. This method is currently required by 
the Technical Specifications. An alternate method to ensure that 
fuel is not loaded into a cell with the control rod withdrawn is to 
prevent control rods from being withdrawn and verify that all 
control rods required to be inserted are fully inserted. The 
proposed actions will require that a control rod block be placed in 
effect thereby ensuring that control rods are not subsequently 
inappropriately withdrawn. Additionally, following placing the 
control rod withdrawal block in effect, the proposed actions will 
require that all required control rods be verified to be fully 
inserted. This verification is in addition to the requirements to 
periodically verify control rod position by other Technical 
Specification requirements. These proposed actions will ensure that 
control rods are not withdrawn and cannot be inappropriately 
withdrawn because an electrical or hydraulic block to control rod 
withdrawal is in place. Like the current requirements the proposed 
actions will ensure that unacceptable operations are blocked (e.g., 
loading fuel into a cell with a control rod withdrawn except 
following the requirements of LCO 3.10.6, ``Multiple Control Rod 
Removal--Refueling,'' which is unaffected by this change).
    The proposed additional acceptable Required Actions provide an 
equivalent level of assurance that fuel will not be loaded into a 
core cell with a control rod withdrawn as the current Required 
Action or the Technical Specification Surveillance Requirement. 
Therefore, the proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The change in the Technical Specification requirements does not 
involve a change in plant design. The proposed requirements will 
continue to ensure that fuel is not loaded into the core when a 
control rod is withdrawn except following the requirements of LCO 
3.10.6, ``Multiple Control Rod Removal-Refueling,'' which is 
unaffected by this change.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    III. The proposed change does not involve a significant 
reduction in a margin of safety.
    As discussed in the Bases for the affected Technical 
Specification requirements, inadvertent criticality is prevented 
during the insertion of fuel provided all required control rods are 
fully inserted during the fuel insertion. The refueling interlocks 
function to support the refueling procedures by preventing control 
rod withdrawal during fuel movement and the inadvertent loading of 
fuel when a control rod is withdrawn.
    The proposed change will allow the refueling interlocks to be 
inoperable and fuel movement to continue only if a control rod 
withdrawal block is in effect and all required control rods are 
verified to be fully inserted. These proposed Required Actions 
provide an equivalent level of protection as the refueling 
interlocks by preventing a configuration which could lead to an 
inadvertent criticality event. The refueling procedures will 
continue to be supported by the proposed required actions because 
control rods cannot be withdrawn and as a result fuel cannot be 
inadvertently loaded when a control rod is withdrawn except 
following the requirements of LCO 3.10.6, ``Multiple Control Rod 
Removal--Refueling,'' which is unaffected by this change.
    Therefore, the proposed changes do not cause 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: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120, for Grand Gulf 
Nuclear Station, and Government Documents Department, Louisiana State 
University, Baton Rouge, LA 70803, for River Bend Station.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502, for 
Grand Gulf Nuclear Station, and Mark Wetterhahn, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005, for River Bend 
Station.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: November 22, 1998.
    Description of amendment request: The proposed amendment would 
revise the reactor thermal margin safety limit lines and flow rates 
stated in the technical specifications (TS). The amendment would also 
update the reference for dose conversion factors used in Dose 
Equivalent Iodine-131 calculations, and administrative changes to the 
criticality analysis uncertainty described in TS 5.6.1.a.1, update the 
analytical methods used in determining core operating limits listed in 
TS 6.9.1.11, and revise the TS bases for the steam generator pressure-
low trip setpoint.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Replacement of the St. Lucie Unit 1 steam generators in 1997 
resulted in an increase in RCS [reactor coolant system] flow. The 
proposed amendment would increase the values of design minimum 
reactor coolant flow and the low flow trip setpoint presently stated 
in the Technical Specifications (TS). These revisions are 
accompanied by a corresponding change to the Thermal Margin Safety 
Limit Lines of TS Figure 2.1-1. The RCS flow related revisions do 
not change the probability of any previously evaluated accident, as 
they do not impact any plant component, structure or system 
affecting the accident initiators. The proposed changes would 
continue to maintain adequate operational margin to TS limits for 
RCS flow and the low-flow trip setpoint.
    The proposed changes to the thyroid dose conversion factors from 
TID-14844 to ICRP-30, fuel storage TS 5.6.1.a.1, the list of 
analytical methods in TS 6.9.1.11, and the Bases for Steam Generator 
Pressure-Low trip setting have no relevance to the accident 
initiators, and thus do not affect the frequency of occurrence of 
previously analyzed transients. Additionally, there are no changes 
to any active plant component due to these proposed changes.
    The supporting evaluation of proposed TS changes demonstrates 
acceptable results for all the accidents previously analyzed, and it 
is concluded that the radiological consequences would remain within 
their established acceptance criteria when including the effects of 
increased RCS flow, increased low flow trip setpoint, and change to 
the thyroid dose conversion factors used in the determination of 
dose consequences. Proposed changes to the Bases for the Steam 
Generator Pressure-Low trip setpoint, fuel storage design features, 
and the list of analytical methods in TS 6.9.1.11 are administrative 
in nature and do not impact current safety analyses.
    Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    This proposed amendment revises limiting flow parameters to 
derive analysis benefits from increased RCS flow due to the 
replacement stream generators, while assuring safe plant operation 
commensurate with the proposed RCS flow and low flow

[[Page 6697]]

trip setpoint changes. These changes along with the proposed changes 
to the Bases for the Steam Generator Pressure-Low trip setpoint, 
dose conversion factors, the list of analytical methods in TS 
6.9.1.11, and the fuel storage design features do not require 
modifications to the plant configuration, systems or components 
which would create new failure modes. There would be no change in 
the modes of operation of the plant. The design functions of all the 
safety systems remain unchanged. Therefore, operation of the 
facility in accordance with the proposed amendment would not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment revises limiting flow parameters to 
derive analysis benefits from increased RCS flow due to the 
replacement steam generators, while assuring safe plant operation 
commensurate with the proposed design minimum RCS flow and low-flow 
trip setpoint changes. FPL has evaluated the impact of the proposed 
changes on available margin to the acceptance criteria for Specified 
Acceptable Fuel Design Limits (SAFDL), 10 CFR 50.46(b) requirements, 
primary and secondary over-pressurization, peak containment 
pressure, potential radioactive releases, and existing limiting 
conditions for operation. With the proposed changes to the design 
minimum RCS flow, low-flow trip setpoint, and dose conversion 
factors, FPL has concluded that there would be no adverse impact to 
the existing safety analyses. The proposed changes to the Bases for 
the Steam Generator Pressure-Low trip setpoint, the list of 
analytical methods in TS 6.9.1.11, and the fuel storage design 
features are administrative in nature. Therefore, operation of the 
facility in accordance with the proposed amendment would 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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Cecil O. Thomas.

Florida Power and Light Company, Et Al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: December 18, 1998.
    Description of amendment request: The proposed amendment would 
revise the St. Lucie Unit 2 Plant Technical Specifications (TS) Index 
Page III; TS 1.10, Dose Equivalent I-131; TS 2.1.1.2, Linear Heat Rate; 
Bases 2.1.1, Reactor Core; Bases Figure B2.1-1, Axial Power 
Distributions for Thermal Margin Safety Limits; Bases 2.2.1, Reactor 
Trip Setpoints (Variable Power Level-High); TS 3.1.1.1/4.1.1.1.1, 
Shutdown Margin--Tavg Greater Than 200  deg.F; TS 3/4.1.1.2, Shutdown 
Margin--Tavg Less Than or Equal to 200  deg.F; TS 3.1.2.2, Boration 
Systems Flow Paths--Operating; TS 3.1.2.4, Charging Pumps--Operating; 
TS 3.1.2.6, Boric Acid Makeup Pumps--Operating; TS 3.1.2.8, Borated 
Water Sources--Operating; Bases 3/4.1.1.1 and 3/4.1.1.2, Shutdown 
Margin; Bases 3/4.1.2, Boration Systems; and TS 6.9.1.11, Core 
Operating Limits Report (COLR). The core operating limits for shutdown 
margin will be relocated to the St. Lucie Unit 2 COLR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the license has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment involves changes to the dose conversion 
factors used in the thyroid dose calculations and the relocation of 
the SHUTDOWN MARGIN requirements for Modes 1 through 5 from TS to 
the Core Operating Limits Report (COLR). Additionally, the peak 
linear heat rate value corresponding to centerline melt is deleted 
from the TS. The deletion of this TS remains consistent with the 
requirements of 10 CFR 50.36. Bases Figure B2.1-1 is replaced with a 
new figure, consistent with the input assumptions of the safety 
analysis report.
    The proposed amendment addresses analytical methods changes such 
as the use of HERMIT code in one dimensional mode for spatial 
details, the rod bow penalty calculations using L2/I 
dependence discussed in CEN-289 (A)-P, CEAW methodology change for 
crediting the delta-T power trip, and the methodology for core 
designs containing Gadolinia-Urania burnable absorbers (CENPD-275-P, 
Revision 1-P, Supplement 1-P). None of these changes is a 
contributor to the initiation of previously evaluated accidents. The 
changes to TS bases and the COLR methodology changes have no impact 
on the accident initiators. Accordingly, the probability of an 
accident previously evaluated is not significantly increased.
    The proposed changes have been evaluated by Florida Power & 
Light (FPL) and Asea Brown Boveri--Combustion Engineering (ABB-CE). 
The safety analyses assumed bounding physics parameters, and satisfy 
all the applicable acceptance criteria. Although specification 
2.1.1.2 is deleted from TS, the safety analyses continue to meet the 
same centerline melt acceptance criteria as before and from which 
the peak linear heat rate value is derived. Additionally, the peak 
linear heat rate value (corresponding to the centerline melt) does 
not meet the criteria specified in 10 CFR 50.36 for safety limits.
    The changes to TS bases do not affect safety analysis results. 
The relocation of SHUTDOWN MARGIN requirements to COLR does not 
affect analysis results or consequences as the limits remain 
unchanged. Future changes to these limits will be controlled per 
Generic Letter 88-16 under the provisions of 10 CFR 50.59.
    The use of HERMITE code in one dimension, for space-time loss-
of-flow simulation, has been successfully applied for other ABB-CE 
plants. The use of HERMITE code in this mode, for St. Lucie Unit 2, 
is acceptable since there are no fundamental core and nuclear steam 
supply system (NSSS) differences between St. Lucie Unit 2 and these 
plants. The analyses presented in this submittal include the use of 
a supplement to the gadolinia-urania core design methodology topical 
report. The change in the rod bow penalty effects similar to that 
approved for another ABB-CE plant is justified for St. Lucie Unit 2 
based on a comparative analysis of factors influencing the rod bow. 
The change in the CEAW analysis method removes unnecessary 
conservatisms as compared to the previous analysis method. The 
validity of results and conclusions of this evaluation are 
contingent upon NRC approval of these revised methods.
    The radiological does consequences for applicable safety 
analyses, using the dose conversion factors from ICRP-30, Supplement 
to Part 1, satisfy the acceptance criteria established to ensure 
compliance with the 10 CFR 100 dose limits.
    The COLR methodology changes proposed to be listed in TS are 
those previously approved for CE plants with changes as described 
above. The use of these methodologies remains consistent with their 
applicability for safety analyses.
    Therefore, the proposed changes do not significantly increase 
the probability or consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment involves changes to the Technical 
Specifications for the dose conversion factors used in the thyroid 
dose calculations, the deletion of TS 2.1.1.2, the replacement of 
Bases Figure B2.1-1, and the relocation of SHUTDOWN MARGIN 
requirements to the COLR. Additionally, there are methodology 
changes related to the safety analyses reported in this submittal. 
The methodology changes include the use of HERMITE code in one 
dimensional mode for space-time loss-of-flow simulations, revised 
rod bow DNB penalty calculations, CEAW analysis methodology change 
including the use of delta-T power trip, and

[[Page 6698]]

a supplement to the methodology for core designs containing 
Gadolinia-Urania burnable absorbers (CENPD-275-P Revision I-P, 
Supplement I-P). None of these changes, including those of the TS 
bases, will affect the plant configuration and there will be no 
impact on any system performance.
    Therefore, this amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed changes to the Technical Specifications have been 
evaluated with respect to the safety analyses using either 
previously approved methodology or methodology currently under NRC 
review (CENPD-275-P, Revision I-P, Supplement
I-P). The use of HERMITE code in one-dimensional mode for spatial 
details, for space-time loss-of-flow simulation, provides more accurate 
data for thermal margin calculations and has been used for similar 
applications at other plants. The calculations of rod bow DNB penalty 
using L\2\/I dependence has been previously approved for another ABB-CE 
plant and is justified for St. Lucie Unit 2 based on an analysis of 
important factors influencing the rod bow. The CEAW methodology change 
showed acceptable analysis results after conservatively accounting for 
appropriate uncertainties.
    The safety analyses performed with this methodology used 
bounding physics parameters to allow flexibility for future cycles 
core designs. The revised Bases Figure B2.1-1 is consistent with the 
attached safety analysis report. Deleting TS 2.1.1.2 is justified 
since the specified limit does not meet any of the criteria of 10 
CFR 50.36, and the fuel centerline melt criteria applied to the 
Specified Acceptable Fuel Design Limit (SAFDL) is not changed. The 
setpoint analyses and safety analyses of all design basis accidents 
meet the applicable acceptance criteria with respect to the 
radiological consequences, SAFDLs, primary and secondary 
overpressurization, and 10 CFR 50.46 requirements. The proposed 
amendment, therefore, will not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Cecil O. Thomas.

Florida Power and Light Company, Et Al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: December 16, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 6.3, ``Unit Staff Qualifications,'' and 
add specific staff qualifications for a Multi-Discipline Supervisor 
position.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed changes are administrative in nature addressing 
personnel qualification issues. The Multi-Discipline Supervisor 
(MDS) position will be filled with personnel who are experienced in 
one or more technical disciplines (maintenance, operations, 
engineering, or other related technical discipline). Fundamental 
working knowledge of tasks being performed will be acquired through 
the MDS initial training program. The training concentrates on 
developing the skills and knowledge of an MDS to safely oversee 
tasks for multi-discipline work teams. Therefore, four years 
experience in any related technical discipline or disciplines 
combined with the MDS training program provide adequate technical 
knowledge for proper job oversight. These proposed changes will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because they do not affect 
assumptions contained in plant safety analyses, the physical design 
and/or operation of the plant, nor do they affect Technical 
Specifications that preserve safety analysis assumptions. Therefore, 
operation of either facility in accordance with its proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The changes being proposed are administrative in nature and do 
not affect assumptions contained in plant safety analyses the 
physical design and/or modes of plant operation defined in the 
facility operating license, or Technical Specifications that 
preserve safety analysis assumptions. These changes address 
qualification requirements for the MDS position. Since the proposed 
changes do not change the qualifications for those individuals 
responsible for the actual licensed operation of the facility, 
operation of the facility in accordance with the proposed amendments 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated. No new failure mode 
is introduced due to the administrative changes since the proposed 
changes do not involve the addition or modification of equipment nor 
do they alter the design or operation of affected plant systems, 
structures, or components. Therefore, operation of either facility 
in accordance with its proposed amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are unchanged by the proposed 
amendments. The proposed changes to add the MDS position have 
management and administrative controls associated with the required 
qualification requirements. The St. Lucie Unit 1 and Unit 2 
Technical Specifications will ensure that any individual filling the 
MDS position has the requisite education, experience, and training. 
The proposed changes do not alter the basis for any technical 
specification that is related to the establishment of, or the 
maintenance of, a nuclear safety margin. Therefore, operation of 
either facility in accordance with its proposed amendment would 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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Direct: Cecil O. Thomas.

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

    Date of amendment request: September 3, 1998.
    Description of amendment request: The amendment would revise 
Technical Specifications 3.4.A.10.e and 3.5.a.2.e to incorporate a 
Condensate Storage Tank level of greater than 35 feet.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 6699]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

    1. The proposed TS change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change does not alter the design or function of any 
structures, systems or components and does not affect any of the 
parameters or conditions that could contribute to initiation of any 
accidents.
    The proposed change eliminated an inconsistency between the 
noted tank level and required water volume and, thereby, ensures 
360,000 gallons of water are available for use. The proposed change 
does not affect the volume of water required to be available, the 
conditions under which it must be available nor the manner in which 
it will be used. Therefore, the proposed TS change does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Eliminating an inconsistency between the noted tank level and 
the required water volume does not alter the designs or function of 
any structures, systems or components. The proposed tank level 
requirement is within the design parameters of the tank and, as 
such, does not [ ] introduce any new mechanisms which could 
contribute to the creation of a new or different kind of accident 
than previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed change eliminates an inconsistency between the 
noted tank level and required water volume. The proposed change 
ensures that an adequate makeup source is available and, in 
addition, that sufficient water volume is available to support 
operation of the core spray system in the event of a reactor vessel 
leak. Therefore, the proposed TS 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: 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: William M. Dean.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1. (NMP1) Oswego County, New York

    Date of amendment request: December 30, 1998.
    Description of amendment request: The footnote of current Technical 
Specification (TS) Table 3.6.14-2, Radioactive Gaseous Effluent 
Monitoring Instrumentation, specifies that the requirement for the 
emergency condenser system to have one operable noble gas activity 
monitor per vent, is applicable during reactor power operating 
conditions. Note (h) of current TS Table 4.6.14-2 specifies that the 
requirement to perform a sensor check once per day of the emergency 
condenser system noble gas activity monitor is applicable during 
reactor power operating conditions. The proposed amendment would change 
the footnote of TS Table 3.6.14-2 and note (h) of TS Table 4.6.14-2 to 
extend the applicability of the channel operability and daily sensor 
check surveillance requirement from during reactor power operating 
conditions, to during power operation conditions and whenever the 
reactor coolant temperature is greater than 212  deg.F except for 
hydrostatic testing with the reactor not critical. The proposed changes 
would also correct a clerical error in TS 4.6.15.d. The clerical error 
cited an incorrect TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The operation of Nine Mile Point Unit 1, in accordance with 
the proposed amendment, will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes extend the application of operability and 
daily sensor check for the Emergency Condenser Vent Noble Gas 
Activity Monitors to include, in addition to power operations, the 
condition when reactor coolant temperature is greater than 212 
deg.F, except for hydrostatic testing. These changes will make the 
conditions for Emergency Condenser Vent Noble Gas Activity Monitor 
operability and daily sensor check surveillance performance 
consistent with the conditions for ECS [emergency cooling system] 
operability as indicated in LCO [Limiting Condition for Operation] 
3.1.3.a.
    The proposed changes to the Emergency Condenser Vent Noble Gas 
Activity Monitor operability and daily sensor check surveillance 
requirements will continue to provide assurance that the intent of 
the effluent monitoring requirements of 10 CFR 50 Appendix A, GDC 
[General Design Criterion] 64, is satisfied and the radiological 
effluents are maintained within the dose and dose rate limits 
specified in 10 CFR 50 Appendix I, 10 CFR 20, and the RETS 
[Radiological Effluent Technical Specifications]. The proposed 
changes will not effect the capability of the ECS to mitigate the 
consequences of an accident that results in a loss of feedwater or 
reactor isolation from the primary heat sink and aid the Core Spray 
System and Automatic Depressurization System in providing effective 
core cooling following non-limiting small breaks.
    The proposed changes also correct a clerical error in the 
Uranium Fuel Cycle effluent monitoring SR [surveillance 
requirement]. The proposed correction simply restores the SR to the 
form that existed before the error was introduced. The clerical 
error did not affect the ODCM [Offsite Dose Calculation Manual] 
implementing procedures or plant operation. Thus, the cumulative 
dose contribution from Uranium Fuel Cycle sources will continue to 
be maintained within the limits of 40 CFR 190 and the RETS.
    Based on the above analysis, the proposed changes do not result 
in any hardware changes or physical alteration of the plant, and the 
changes will have no impact on the design or function of any 
structure, system or component (SSC). As such, the SSC process 
variables, characteristics, and functional performance will be 
maintained consistent with the event initiator and the initial 
condition assumptions for the accident analyses. Moreover, the 
proposed changes will not eliminate any actions or adversely affect 
any SSCs required to prevent accidents or mitigate accident 
conditions, nor will the changes result in the degradation of any 
fission product barriers so as to increase the radiological 
consequences of an accident. It is, therefore, concluded that 
operation in accordance with the proposed amendment will not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The operation of Nine Mile Point Unit 1, 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 proposed changes do not result in any hardware changes or 
physical alteration of the plant, and the changes do not impact the 
design or function of any SSC. The proposed changes maintain the 
capability of the ECS to respond to accidents, including non-
limiting small breaks, consistent with the current analyses. In 
addition, the proposed changes provide continued assurance that the 
radiological dose and dose rates will be maintained within limits. 
The proposed changes do not alter the process variables, 
characteristics, or functional performance of any SSC, do not 
eliminate any requirements, and do not impose any new requirements 
which could introduce new equipment failure modes or create new 
credible accidents. It is, therefore, concluded that operation in 
accordance with proposed amendment 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 1, in accordance with 
the proposed

[[Page 6700]]

amendment, will not involve a significant reduction in a margin of 
safety.
    The proposed changes do not affect the capability of the ECS to 
mitigate consequences of an accident that results in a loss of 
feedwater or reactor isolation from the primary heat sink, or affect 
the capability of the ECS to aid the Core Spray System and the 
Automatic Depressurization System in providing effective core 
cooling following non-limiting small breaks. Thus, there will be no 
impact on the post-accident radioactive material release analyses or 
a reduction in the margin to the associated 10 CFR 100 dose limits. 
In addition, the proposed changes provide continued assurance that 
the intent of the effluent monitoring requirements of 10 CFR 50 
Appendix A, GDC 64, is satisfied and the dose and dose rates due to 
the radiological effluents are maintained within the limits 
specified in 10 CFR 50 Appendix I, 10 CFR 20, 40 CFR 190, and the 
RETS. Moreover, the proposed changes do not eliminate any 
requirements or responsibilities, nor impose new requirements or 
responsibilities, or alter any physical parameters which could 
reduce the margin to an acceptance limit. It is, therefore, 
concluded that operation in accordance with the proposed amendment 
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, Director.

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

    Date of amendment request: December 16, 1998.
    Description of amendment request: The proposed editorial and 
administrative changes to the Technical Specifications would either 
revise references and statements that are inaccurate or provide relief 
from administrative controls which provide insignificant safety 
benefit.
    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 design basis accidents are not affected by the proposed 
editorial and administrative changes. The proposed changes do not 
change the level of programmatic controls or the procedural details 
currently in place. The proposed changes do not revise the station 
design, the response of the station to transients nor the manner in 
which the station is operated, therefore, these changes have no 
adverse affect to the safe operation of the station. The proposed 
changes do not involve a significant increase in 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 previously analyzed.
    The proposed changes do not alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. There are no changes to the source term, 
containment isolation or radiological release assumptions used in 
evaluating the radiological consequences in the Seabrook Station 
UFSAR. Existing system and component redundancy is not being changed 
by the proposed changes. The proposed changes have no adverse affect 
on component or system interactions. The proposed changes are 
editorial and administrative in nature and do not change the level 
of programmatic controls and procedural details associated with the 
aforementioned technical specifications. Therefore, since there are 
no changes to the design assumptions, conditions, configuration of 
the facility, or the manner in which the plant is operated and 
surveilled, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    There are no changes being made to the Technical Specification 
safety limits or safety system settings that would adversely affect 
plant safety. The changes do not affect the operation of structures, 
systems or components nor do they introduce administrative changes 
to plant procedures that could affect operator response during 
normal, abnormal or emergency situations. Therefore, the proposed 
changes do 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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: William M. Dean.

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

    Date of amendment request: December 16, 1998.
    Description of amendment request: The proposed change would 
relocate Technical Specifications (TS) 3/4.7.10, ``Area Temperature 
Monitoring,'' and associated TS Table 3.7-3, to the Seabrook Station 
Technical Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. The proposed change does not alter or prevent the 
ability of structures, systems, or components (SSCs) to perform 
their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR). The proposed change is 
administrative in nature and does not decrease the effectiveness of 
programmatic controls or the procedural details of assuring 
operation of the facility in a safe manner.
    The provisions of TS 3/4.7.10 for area temperature monitoring of 
the referenced selected areas is neither part of an initial 
condition of a design basis accident or transient that either 
assumes the failure of or presents a challenge to the integrity of a 
fission product barrier, nor is area temperature monitoring relied 
upon as a primary success path to mitigate such events. The 
provisions for area temperature monitoring is not related to events 
that are considered frequent or dominant contributors to plant risk. 
Area temperature monitoring is not considered a design feature or an 
operating restriction that is an initial condition of a design basis 
accident or transient analysis, nor does it provide a function or 
actuate any accident mitigation feature in order to mitigate the 
consequences of a design basis accident or transient.
    Relocating TS 3/4.7.10 to the Technical Requirements Manual will 
still provide adequate controls for area temperature in those areas 
designated in TS Table 3.7-3. The relocated requirements of TS 3/
4.7.10 to the Technical Requirements Manual will continue to be 
administratively controlled in accordance with TS Section 6.0, 
``Administrative Controls.''
    The Seabrook Station Technical Requirements Manual is a 
licensee-controlled

[[Page 6701]]

document which contains certain technical requirements and is the 
implementing manual for the Technical Specification Improvement 
Program. Changes to these requirements are reviewed and approved in 
accordance with Seabrook Station Technical Specifications, Section 
6.7, and as outlined in the Technical Requirements Manual. 
Specifically, changes to the Technical Requirements require a 10 CFR 
50.59 safety evaluation and are reviewed and approved by the Station 
Operations Review Committee (SORC) and the Nuclear Safety Audit 
Review Committee (NSARC) prior to implementation.
    The proposed change will not degrade the ability of systems, 
structures and components important to safety to perform their 
safety function. The proposed change will not change the response of 
any system, structure or component important to safety as described 
in the Seabrook Station Updated Final Safety Analysis Report 
(UFSAR). Since the plant response to an accident will not change, 
there is no change in the potential for an increase in the 
consequences of an accident previously analyzed. As such, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed change does not alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. There are no changes to the source term, 
containment isolation or radiological release assumptions used in 
evaluating the radiological consequences in the Seabrook Station 
UFSAR. Existing system and component redundancy is not being changed 
by the proposed change. The proposed change has no adverse impact on 
component or system interactions. The proposed change will not 
adversely degrade the ability of systems, structures and components 
important to safety to perform their safety function nor change the 
response of any system, structure or component important to safety 
as described in the Seabrook Station Updated Final Safety Analysis 
Report (UFSAR). The proposed change is administrative in nature and 
does not change the level of programmatic controls and procedural 
details controls of assuring operation of the facility in a safe 
manner. Therefore, since there are no changes to the design 
assumptions, conditions, configuration of the facility, or the 
manner in which the plant is operated and surveilled, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously analyzed.
    Future changes to area temperature monitoring requirements will 
be reviewed and approved in accordance with Seabrook Station 
Technical Specifications, Section 6.7, and as outlined in the 
Technical Requirements Manual. Specifically, changes to the 
Technical Requirements require a 10 CFR 50.59 safety evaluation and 
are reviewed and approved by the Station Operations Review Committee 
(SORC) and the Nuclear Safety Audit Review Committee (NSARC) prior 
to implementation.
    Since the plant response to an accident will not change, there 
is no change in the potential for an increase in the consequences of 
an accident previously analyzed, nor can it create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    Relocation of the area temperature monitoring requirements to 
the Technical Requirements Manual will not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There is no adverse impact on equipment design or operation and 
there are no changes being made to the Technical Specification 
required safety limits or safety system settings that would 
adversely affect plant safety. The proposed change is administrative 
in nature and does not change the level of programmatic controls and 
procedural details associated with area temperature monitoring to 
ensure that environmentally qualified equipment will not be exposed 
to temperatures beyond that which they were originally qualified.
    Future changes to the area temperature monitoring requirements 
will be reviewed and approved in accordance with Seabrook Station 
Technical Specifications, Section 6.7, and as outlined in the 
Technical Requirements Manual. Specifically, changes to the 
Technical Requirements require a 10 CFR 50.59 safety evaluation and 
are reviewed and approved by the Station Operations Review Committee 
(SORC) and the Nuclear Safety Audit Review Committee (NSARC) prior 
to implementation.
    Relocation of the requirements contained in TS 3/4.7.10 to the 
Technical Requirements Manual 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 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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: William M. Dean.

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

    Date of amendment request: December 28, 1998.
    Description of amendment request: NNECO is proposing to change 
Technical Specification 2.2.1, ``Limiting Safety System Settings--
Reactor Trip Setpoints,'' and the associated Bases to reflect revised 
loss of normal feedwater (LONF) analyses. An additional Technical 
Specification Bases change to the floor value for the thermal margin 
low pressure reactor trip is also included. This proposed change is not 
related to the revised LONF analyses.
    NNECO is also seeking NRC approval to incorporate changes to the 
Millstone Unit No. 2 Final Safety Analysis Report (FSAR). The proposed 
changes to the FSAR, except the floor value for thermal margin low 
pressure reactor trip, are associated with the revised LONF analyses.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with 10 CFR 50.92, NNECO has reviewed the proposed 
changes and has concluded that they do not involve a Significant 
Hazards Consideration (SHC). The basis for this conclusion is that 
the three criteria of 10 CFR 50.92(c) are not compromised. The 
proposed changes do not involve an SHC because the changes would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The analysis of a loss of normal feedwater (LONF) event, as 
described in the Millstone Unit No. 2 FSAR Chapters 10 and 14, has 
been revised. Certain key assumptions have been changed to ensure 
acceptable analysis results. An evaluation of the LONF analyses 
changes, and associated Technical Specification changes will be 
presented. In addition, an evaluation of an additional non LONF 
analyses related Technical Specification Bases and FSAR change is 
included.
    LONF analyses changes. The LONF analyses, contained in FSAR 
Chapters 10 and 14, have been revised using a steam generator liquid 
inventory assumption, at the time of reactor trip on low steam 
generator water level, that is consistent with the design of the 
replacement steam generators. The revised Chapter 10 and 14 LONF 
analyses also incorporate a reduction in auxiliary feedwater 
delivery rates resulting from a recalculation of the Auxiliary 
Feedwater (AFW) System flows. The results of revised analyses 
indicate that the analytical limit for the low steam generator water 
level reactor trip must be raised to 43% narrow range level from the 
current 34% narrow range level. This will result in a change to the 
low steam generator water level reactor trip setpoint listed in 
Technical Specification 2.2.1.
    The revised Chapter 14 LONF analysis will now take credit for 
automatic initiation of the motor driven auxiliary feedwater (MDAFW) 
pumps. The current Chapter 14 LONF analysis assumes auxiliary 
feedwater flow will be initiated 10 minutes after the event. The 
Chapter 10 LONF analysis assumption of

[[Page 6702]]

automatic initiation of one MDAFW pump within 4 minutes, after the 
low steam generator level AFW actuation setpoint is reached, has not 
changed.
    To demonstrate that one MDAFW pump delivers sufficient flow to 
preclude steam generator dryout, the Chapter 10 LONF analysis will 
not take credit for the operation of the steam generator atmospheric 
dump valves, instead of the main steam safety valves as in the 
current analysis. This new assumption yields lower predicted steam 
generator pressures which result in an increase in the delivered AFW 
flows.
    LONF analyses related technical specification changes. The trip 
setpoint and allowable value for the low steam generator water level 
reactor trip will be changed to be consistent with the revised LONF 
analyses. The revised analyses assume an analytical limit of 43% 
narrow range level, instead of the current analytical limit of 34% 
narrow range level. The calculation of the trip setpoint, which 
includes instrument uncertainty, has determined that the trip 
setpoint should be changed from [greater than or equal to] 36.0% to 
[greater than or equal to] 48.5%.
    The increase in the low steam generator level Reactor Protection 
System (RPS) actuation setpoint from [greater than or equal to] 36% 
to [greater than or equal to] 48.5% will result in an increase in 
the probability of an RPS actuation on low steam generator water 
level since the difference between the proposed setpoint and the 
normal operating value of steam generator level will decrease. The 
proposed actuation setpoint is below the normal operating level of 
60 to 75%. Steam generator level is not expected to approach the 
actuation setpoint during normal operation. An unexpected plant 
event (e.g., loss of main feedwater or difficulty controlling steam 
generator level at low power levels) would be necessary for steam 
generator level to approach the actuation setpoint. To provide the 
operators with advance notice of the steam generator low level 
condition, the existing RPS low steam generator water level pretrip 
alarm setpoint will be changed to provide approximately the same 
margin between pretrip and trip as currently exists (5%). This will 
ensure that the pretrip alarm is received prior to reaching the 
actual record trip setpoint. Therefore, even though the proposed 
change will decrease the margin between the normal operating steam 
generator level and the RPS actuation setpoint, this change will not 
significantly impact the probability of an RPS actuation on low 
steam generator level during normal plant operations. In addition, 
the proposed setpoint and allowable value change will ensure a 
reactor trip signal is generated at, or before the analytical limit 
used in the revised LONF analysis is reached. Therefore, the RPS 
will continue to function as designed to mitigate the consequences 
of the design basis accidents.
    The basis for the steam generator level low reactor trip will be 
modified to be consistent with the revised LONF analyses. The 
discussion concerning available water inventory and time until 
auxiliary feedwater is required will be removed. The proposed change 
to the FSAR will include a discussion of the relationship between 
the LONF analysis and the need to automatically initiate auxiliary 
feedwater flow.
    Non LONF analyses related technical specification bases and FSAR 
change. This Technical Specification Bases and FSAR change is not 
related to the revised LONF analyses.
    The basis for the thermal margin low pressure (TMLP) reactor 
trip (Technical Specification 2.2.1 Bases) will be modified. The 
current basis states that the floor, or minimum value, for this trip 
function is set at 1850 psia pounds per square inch absolute]. This 
value will be changed to be consistent with instrument uncertainty 
calculations that have determined that the floor should be increased 
to 1865 psia. The increase in floor value is the result of greater 
instrument uncertainties when harsh containment environment 
conditions are included.
    The increase in the TMLP floor (from 1850 psia to 1865 psia) 
could result in an increase in the probability of an RPS actuation 
on thermal margin low pressure since the difference between the 
proposed floor setpoint and the normal operating value of 
pressurizer pressure will decrease. However, the proposed actuation 
setpoint is significantly below the normal operating pressure of 
approximately 2250 psia. Pressurizer pressure is not expected to 
approach the actuation setpoint during normal operation. A 
significant plant event (e.g., loss of primary coolant) would be 
necessary for a rapid pressure excursion to approach the actuation 
setpoint. Since the setpoint change is small, it will not adversely 
impact the probability of an RPS actuation on low pressurizer 
pressure during normal plant operations. In addition the proposed 
change to the floor value will ensure a reactor trip signal is 
generated at, or before the analytical limit used in the respective 
accident analyses is reached. Therefore, the RPS will continue to 
function as designed to mitigate the consequences of the design 
basis accidents.
    Conclusion. The results of the revised LONF analyses contained 
in FSAR Chapters 10 and 14 have concluded that the LONF event does 
not result in the violation of the Specified Acceptable Fuel Design 
Limits, that the peak pressurizer pressure does not exceed 110% of 
the design pressure, that liquid primary coolant is not expelled 
through the pressurizer safety valves, and that adequate cooling 
water is supplied by the AFW System to prevent steam generator 
dryout and allow a safe and orderly plant shutdown. By preventing 
steam generator dryout, sufficient removal of decay heat from the 
Reactor Coolant System (RCS) will occur, preventing excessive RCS 
heatup and pressurization. This will ensure the steam generator 
fatigue analysis remains valid, and excessive discharge of primary 
coolant through the pressurizer safety valves does not occur. 
Therefore, there will be no adverse effect on the consequences of a 
LONF event. This is consistent with the acceptance criteria 
contained in Standard Review Plan (SRP) 15.2.7, [``Loss of Normal 
Feedwater Flow,'' Rev. 1--July 1981]. (Millstone Unit No. 2 is not 
an SRP plant.)
    The proposed changes do not alter the way any structure, system, 
or component functions. The changes in actuation setpoints and 
equipment used in the LONF analyses affect equipment important to 
the mitigation of design basis accidents. These changes do not 
affect any equipment that can cause a design basis accident to 
occur. Therefore, the proposed changes do not affect the probability 
of occurrence of a previously evaluated accident.
    These proposed changes do not alter the way any structure, 
system, or component functions. There will be no adverse effect on 
any design basis accident previously evaluated, on any equipment 
important to safety, or on the radiological consequences of any 
design basis accident. Therefore, these proposed changes will not 
adversely affect the consequences of a previously evaluated 
accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Results of the proposed LONF analyses have demonstrated that the 
Specified Acceptable Fuel Design Limits are not violated, that the 
peak pressurizer and steam generator pressures do not exceed 110% of 
the design pressure, that liquid primary coolant is not expelled 
through the pressurizer safety valves, and that adequate cooling 
water is supplied by the AFW System to prevent steam generator 
dryout and allow a safe and orderly plant shutdown. Therefore, there 
are no new or different types of failures of systems or equipment 
important to safety which could cause a new or different type 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 revised FSAR Chapter 14 analysis has concluded that the 
steam generator low water level reactor trip setpoint does not 
provide sufficient water inventory in the steam generators at the 
time of the reactor trip such that auxiliary feedwater flow will not 
be required for 10 minutes. This contradicts the current Technical 
Specification Basis (Technical Specification 2.2.1) for the steam 
generator low water level reactor trip setpoint. Therefore, the 
revised analysis reduces the margin of safety as defined in the 
Bases of the Millstone Unit No. 2 Technical Specifications. However, 
with the proposed changes to increase the low steam generator water 
level reactor trip setpoint and taking credit for automatic AFW 
System actuation, it has been shown that operation of these systems 
can mitigate the LONF event, and ensure plant response is within the 
acceptance criteria. Results of the proposed LONF analyses have 
demonstrated that the Specified Acceptable Fuel Design Limits are 
not violated, that the peak pressurizer and

[[Page 6703]]

steam generator pressures do not exceed 110% of the design pressure, 
that liquid primary coolant is not expelled through the pressurizer 
safety valves, and that adequate cooling water is supplied by the 
AFW System to prevent steam generator dryout and allow a safe and 
orderly plant shutdown. Therefore, these proposed changes do not 
involve a significant reduction in a margin of safety.
    The proposed change to the floor value for the TMLP reactor trip 
function is the result of a revision to the instrument loop 
uncertainty and setpoint calculations. The proposed change to the 
Technical Specification Basis will incorporate the RPS TMLP floor 
setpoint change. This change to the TMLP floor will not adversely 
affect this function. The TMLP reactor trip function will still 
operate as designed. The RPS will continue to function as designed 
to mitigate the consequences of design basis accidents. Therefore, 
this proposed change does not involve a significant reduction in a 
margin of safety.
    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. The changes proposed herein are not enveloped by any 
specific example.
    As described above, this License Amendment Request does not 
impact the probability of an accident previously evaluated, does not 
involve a significant increase in the consequences of an accident 
previously evaluated, does not create the possibility of a new or 
different kind of accident from any accident previously evaluated, 
and does not result in a significant reduction in a margin of 
safety. Therefore, NNECO has concluded that the proposed changes do 
not involve an SHC.

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

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

    Date of amendment request: January 18, 1999.
    Description of amendment request: NNECO is proposing to change 
Technical Specification 3.6.1.2, ``Containment Systems--Containment 
Leakage.'' The Bases for this Technical Specification and the Final 
Safety Analysis Report (FSAR) will also be modified to address the 
proposed changes.
    The limit for secondary containment bypass leakage specified in 
Technical Specification 3.6.1.2.c will be reduced from less than 0.017 
La to less than 0.0072 La. This new limit is 
consistent with the value of secondary containment bypass leakage used 
in the revised off-site and control room dose calculations following a 
design basis loss-of-coolant accident (LOCA).
    Technical Specification 3.6.1.2.c will be modified by replacing 
``identified in Table 3.6-1 as'' with ``that are.'' This will allow 
Table 3.6-1 to be removed. The removal of this table from Technical 
Specifications and the proposed wording change are consistent with the 
guidance contained in Generic Letter (GL) 91-08. It is not necessary to 
maintain a list of the secondary containment bypass leakage paths in 
Technical Specifications. The Millstone Unit No. 2 FSAR (Section 5.3.4) 
provides the necessary information to determine the secondary 
containment bypass leakage paths that must be considered to ensure that 
the combined leakage rate limit contained in Technical Specification 
3.6.1.2.c is met.
    Technical Specification 3.6.1.2 Table 3.6-1, ``Secondary 
Containment Bypass Leakage Paths,'' will be removed and the phrase 
``This Page Intentionally Deleted'' will be added to Page 3/4 6-5.
    The Bases for Technical Specification 3.6.1.2 will be modified to 
indicate that the Millstone Unit No. 2 FSAR contains a list of the 
containment penetrations that have been identified as secondary 
containment bypass leakage paths.
    FSAR Section 5.3.4, ``Through-Line Leakage Evaluation,'' will be 
changed to include the additional secondary containment bypass leakage 
paths that have been identified. The criteria used to determine the 
secondary containment bypass leakage paths will be modified to be 
consistent with the criteria used in the evaluation that identified the 
additional leakage paths.
    The discussion of the use of a leakage rate of 11 cc/hr for the 
control room dose calculations will be modified. The revised control 
room dose calculations will assume a total secondary containment bypass 
leakage rate consistent with the proposed change to Technical 
Specification 3.6.1.2.
    As a result of these proposed changes, the calculated off-site and 
control room doses following a design basis LOCA will change. The 
calculated doses are specified in FSAR Section 14.8.4, ``Radiological 
Consequences of the Design Basis Accident.'' A revision to this section 
of the FSAR has been submitted to the NRC by the letter dated September 
28, 1998. This submittal will be revised to incorporate the proposed 
total secondary containment bypass leakage rate and the associated 
change to the calculated off-site and control room doses following a 
design basis LOCA.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    In accordance with 10 CFR 50.92, NNECO has reviewed the proposed 
changes and has concluded that they do not involve a significant 
hazards consideration (SHC). The basis for this conclusion is that 
the three criteria of 10 CFR 50.92(c) are not compromised. The 
proposed changes do not involve an SHC because the changes would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to lower the limit for secondary containment 
bypass leakage, as specified in Technical Specification 3.6.1.2.c, 
from [less than] 0.017 La to [less than] 0.072 
La will reduce the off-site doses associated with the 
design basis LOCA. The proposed change to raise the limit for 
secondary containment bypass leakage from 11 cc/hr to [less than] 
0.0072 La will increase the dose to the Control Room 
Operators following a design basis LOCA. However, the revised off-
site and control room dose calculations, using the proposed combined 
secondary containment bypass leakage limit, demonstrate that the 
limits of 10 CFR 100 and 10 CFR 50, Appendix A, General Design 
Criteria (GDC) 19 are met. In addition, these proposed changes will 
result in the use of the same limit for secondary containment bypass 
leakage when determining the radiological consequences of a design 
basis LOCA.
    The proposed wording change to Technical Specification 
3.6.1.2.c, and the associated removal of Table 3.6-1, will not 
change the requirement to verify total secondary containment bypass 
leakage is within the limit assumed in the determination of the 
radiological consequences of the design basis LOCA. Control of the 
penetrations that have been identified as secondary containment 
bypass leakage paths will be maintained by the process used to 
change the Millstone Unit No. 2 FSAR. This process ensures that 
appropriate changes to the FSAR are evaluated in accordance with 10 
CFR 50.59 to determine if NRC approval is required prior to 
implementing the change. This process also ensures that the NRC is 
informed of FSAR changes via regular

[[Page 6704]]

updates to the FSAR. The removal of Table 3.6-1 from Technical 
Specifications and the proposed wording change are consistent with 
the guidance contained in GL 91-08.
    The identification and addition of more secondary containment 
bypass leakage paths to the FSAR will have no impact on the 
calculated off-site and control room doses following a design basis 
LOCA since the combined leakage through all secondary containment 
bypass leakage paths is limited to the proposed value contained in 
Technical Specification 3.6.1.2. The addition of bypass leakage 
paths does not change the combined leakage limit, which is now used 
in the off-site and control room dose calculations.
    The Bases for Technical Specification 3.6.1.2 will be modified 
to indicate that the Millstone Unit No. 2 FSAR contains a list of 
the containment penetrations that have been identified as secondary 
containment bypass leakage paths.
    The proposed changes do not alter the way any structure, system, 
or component functions. These changes do not affect any equipment 
that can cause a design basis accident to occur. There will be no 
adverse effect on any design basis accident previously evaluated or 
on any equipment important to safety. The reduction in the allowable 
secondary containment bypass leakage limit will result in a decrease 
in the calculated off-site doses associated with the design basis 
LOCA. The use of the proposed secondary containment bypass leakage 
limit will increase the calculated doses to the Control Room 
Operators following a design basis LOCA. However, the calculated 
doses meet the criteria of 10 CFR 100 and GDC 19. Therefore, there 
will be no significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes 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. Also, the response of the plant and 
the operators following these accidents is essentially unaffected by 
the change. The criteria used by the plant operators to terminate 
containment spray following a design basis LOCA will change from 
containment pressure to either time or pressure, whichever requires 
longer operation. This will ensure that containment spray remains in 
operation long enough to achieve the assumed iodine decontamination. 
However, the operator action to terminate containment spray will 
remain the same. 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 lower the Technical Specification limit 
for secondary containment bypass leakage, to remove Table 3.6-1, and 
to add more secondary containment bypass leakage paths to the FSAR 
will have no adverse effect on equipment important to safety. The 
equipment will continue to function as assumed in the design basis 
accident analysis. These changes will ensure that the secondary 
containment bypass leakage paths are identified and tested to verify 
that the total secondary containment bypass leakage does not exceed 
the Technical Specification limit. This will ensure that the 
expected off-site and control room doses following a design basis 
LOCA are within the limits specified in 10 CFR 100 and GDC 19. 
Therefore, there will be no significant reduction in the margin of 
safety as defined in the Bases for the Technical Specification 
affected by these 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: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M, Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Project Director: William M. Dean.

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

    Date of amendment request: January 18, 1999.
    Description of amendment request: The proposed changes will remove 
the Technical Specification related to Hydrogen Purge System from the 
Millstone Unit No. 2 Technical Specifications. The proposed changes 
affect Technical Specifications 3/4.6.4.3, ``Containment Systems, 
Hydrogen Purge System.'' The Bases of the associated Technical 
Specification will be modified to address the proposed changes. The 
proposed changes will allow the licensee to downgrade the hydrogen 
purge system to a non-safety-related system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with 10 CFR 50.92, NNECO has reviewed the proposed 
changes and has concluded that they do not involve a Significant 
Hazards Consideration (SHC). The basis for this conclusion is that 
the three criteria of 10 CFR 50.92(c) are not compromised. The 
proposed changes do not involve an SHC because the changes would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The Hydrogen Purge System provides a backup means to manually 
control the hydrogen concentration in containment given the multiple 
failure of the redundant, Seismic Category I Hydrogen Recombiner 
System. The primary success path for hydr9gen control is the 
Hydrogen Recombiner System. The Hydrogen Recombiner System has 
redundant trains and is fully qualified to maintain hydrogen control 
following a design basis accident. FSAR [Final Safety Analysis 
Report] Section 14.8.3.5, ``Radiological Consequences of Purging'' 
is being removed from the FSAR since it is no longer required. Since 
the hydrogen recombiners are fully redundant, it is not necessary to 
postulate offsite doses for purge during a design basis accident. 
Thus, the deletion of consequences does not represent a change in 
the consequences of a design basis event. Therefore, this change 
will not significantly increase the probability or consequences of 
an accident previously evaluated.
    Revision of Index Page VII is an administrative change. The 
proposed change to Bases section 3/4.6.4 by deleting reference to 
``the purge system'' is required since Technical Specification 3/
4.6.4.3 is being removed. Therefore, these changes 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 purge system is a standby purge system which is not in 
service during normal operations as a hydrogen purge system (i.e., 
Charcoal Filter Heaters de-energized). Therefore, no new accident is 
created either by system unavailability or actuation. The FSAR will 
still address the use of the purge system as a backup to the 
recombiner system, Revision of Index Page VII is an administrative 
change. The proposed change to Bases section 3/4.6.4 by deleting 
reference to ``the purge system'' is required since Technical 
Specification 3/4.6.4.3 is being removed. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.

[[Page 6705]]

    3. Involve a significant reduction in a margin of safety.
    The margin of safety is defined in the Bases 3/4.6.4 which 
states that the ``hydrogen control systems are consistent with the 
recommendations of Regulatory Guide 1.7 * * * ''. Regulatory Guide 
1.7 describes methods that would be acceptable in meeting the 
standards for a combustible gas control system, 10 CFR 50.44, 
``Standards for combustible gas control systems in light-water-
cooled power reactors.'' Regulatory Guide 1.7 acknowledges that 
purging is a means of reducing the hydrogen concentration but it 
should not be the primary means because of the release of 
radioactivity to the environment. The regulatory guide does advise 
that there be an ``installed capability for a controlled purge of 
the containment atmosphere to aid in cleanup.'' Removal of the 
Hydrogen Purge System Technical Specification is consistent with 
Regulatory Guide 1.7. Additionally, the capability to purge is still 
documented in the FSAR. Revision of Index Page VII is an 
administrative change. The proposed change to Bases section 3/4.6.4 
by deleting reference to ``the purge system'' is required since 
Technical Specification 3/4.6.4.3 is being removed. Therefore, the 
proposed changes will not result in a significant reduction in the 
margin of safety as defined in the Bases for Technical 
Specifications covered in this License Amendment Request.

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

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

    Date of amendment request: January 18, 1999.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3/4.2.2 to be in accordance with 
NRC-approved Westinghouse methodologies for the heat flux hot channel 
factor--FQ(Z). In addition, the proposed amendment would 
make changes to the core operating limits and the analytical methods 
used to determine core operating limits contained in Section 6.9.1.6.a 
and b, respectively, by adding, modifying, or deleting references.
    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 signification hazards 
consideration, which is presented below:

    NNECO has reviewed the proposed revision in accordance with 10 
CFR 50.92 and has concluded that the revision does not involve any 
Significant Hazards Considerations (SHC). The basis for this 
conclusion is that the three criteria of 10 CFR 50.92(c) are not 
satisfied. The proposed Technical Specification revision does not 
involve an SHC because the revision would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    To determine any potential impact, the proposed changes to the 
TS are grouped into the following two categories.
    (a) Changes to Technical Specification 3/4.2.2 ``Heat Flux Hot 
Channel Factor--FQ(Z)''
    (b) Changes that are not related to the Heat Flux Hot Channel 
Factor TS, and are administrative in nature. These include defining 
a new core operating limit and deleting, re-numbering, updating and 
adding references to analytical methods used to determine core 
operating limits in TS 6.9.1.6 ``Core Operating Limit Report 
(COLR).['']
    With respect to item 1.a changes related to the Heat Flux Hot 
Channel Factor, FQ(Z), impact the initial conditions 
assumed in the accidents analyzed for MP3 [Millstone Unit 3]. These 
initial conditions are power distributions which are consistent with 
reactor operation as defined in the TS. The proposed changes to the 
Heat Flux Hot Channel Factor TS ensure that proper actions are taken 
to maintain peaking factors within the limits assumed in the MP3 
accident analysis. The proposed changes are consistent with the NRC 
approved Westinghouse methodology for FQ(Z) surveillance. 
Changes to the SURVEILLANCE and ACTION statements will not change 
the probability of occurrence of any analyzed accidents. 
Furthermore, the consequences of analyzed accidents will not change 
since the power distribution assumptions will not be challenged by 
reactor operation allowed by the Technical Specifications.
    With respect to item 1.b the administrative changes to the 
Technical Specifications do not affect existing or proposed Limiting 
Conditions for Operation (LCO) or SURVEILLANCE REQUIREMENTS. 
Therefore, there is no impact on the design basis accidents.
    Thus it is concluded that the proposed revision does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (a) Proposed changes to the Heat Flux Hot Channel Factor, TS 3/
4.2.2 ensure that proper actions are taken to maintain peaking 
factors within the limits assumed in the MP3 accident analysis. The 
proposed changes are consistent with the NRC approved Westinghouse 
methodology for FQ(Z) surveillance. Maintaining safety 
analysis assumptions on power distributions cannot be an initiating 
event for any design basis accidents and will not create the 
possibility of a different type of accident. Therefore the changes 
associated with the Heat Flux Hot Channel Factor limiting condition 
for operation do not represent a new unanalyzed accident.
    (b) Since the administrative changes do not affect plant 
operation, the potential for an unanalyzed accident is not created. 
No new failure modes are introduced.
    Thus, this proposed revision does not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Involve a significant reduction on the margin of safety.
    (a) The proposed changes ensure that FQ(Z), will 
remain within the safety analysis assumptions. The LCO limits and 
SURVEILLANCE REQUIREMENTS are not altered. Therefore, the impact on 
the consequences on the protective boundaries is unchanged. Meeting 
the intent of the NRC approved Westinghouse methodology for 
FQ(Z), SURVEILLANCE ensures that power distributions 
assumed in the accident analysis will not be challenged by reactor 
operations allowed by the Technical Specifications. Therefore, 
verification of no change in the margin of safety is encompassed by 
meeting the power distribution limits assumed in analyzed accidents.
    (b) Since the proposed changes do not affect the consequences of 
any accident previously analyzed, there is no reduction in the 
margin of safety.
    Thus it is concluded that the proposed revision does not involve 
a significant reduction in the margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve a Significant 
Hazard 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: 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.

[[Page 6706]]

    NRC Project Director: William M. Dean.

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

    Date of amendment request: December 31, 1998.
    Description of amendment request: The proposed amendment would 
revise the technical specification (TS) reactor pressure vessel (RPV) 
pressure-temperature (P-T) limit curves, delete completed RPV sample 
surveillance requirements, delete requirement to withdraw a specimen at 
next refueling outage, and remove the standby liquid control system 
(SBLC) relief valve setpoint. Associated administrative changes are 
also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    RPV P-T curve changes. It is proposed that P-T curves be revised 
to accommodate the shift in RTNDT determined using actual 
surveillance program data rather than generic data provided in 
Regulatory Guide [RG] 1.99 Revision 2 (Radiation Embrittlement of 
Reactor Vessel Materials). The new P-T curves will increase the 
margins provided in the P-T limit curves against non-ductile failure 
of the RPV. Regulatory Guide 1.99 Revision 2 encourages use of plant 
specific surveillance data as data becomes available.
    Eliminating prescriptive requirements to remove a RPV test 
specimen sample at three fourths service life will result in an 
overall improvement in the RPV surveillance program since the 
limited number of remaining surveillance samples will be removed at 
optimum intervals. Therefore, proposed changes will neither 
significantly increase the probability or the consequences of an 
accident previously evaluated.
    RPV surveillance requirements. Deleting completed, one time 
surveillance requirements [SRs] of SR section 4.6.B and 
incorporating a discussion of the results in the Bases is an 
administrative change and has no effect on probability or 
consequences of accidents.
    SBLC relief valve setpoint testing. The testing requirements of 
TS section 4.4.A.2.c are enveloped by the current testing performed 
by Monticello's IST [inservice test] Program, which implements ASME 
[American Society of Mechanical Engineers] Code Section XI, approved 
by 10 CFR 50.55a. The IST program requires all relief valves to be 
tested to their nameplate data setpoints. Any modification to a 
relief valve's nameplate data is controlled by the plant's 
configuration control process which would ensure the requirements of 
ASME Section XI are invoked as required by TS section 3.15. The IST 
program required by TS 4.15 ensures the SBLC relief valves will be 
properly tested for operability. Therefore, revising section 
4.4.A.2.c to remove specific setpoints does not increase the 
probability or consequences of an accident.
    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    RPV P-T curve change. Updated RPV P-T limit curves will not 
create the possibility of a new or different kind of accident nor 
alter operational standards. New limits continue a system of 
operating bounds which are in place to prevent damage to reactor 
vessels during normal operating conditions including hydrostatic 
pressure and leakage testing, and anticipated transients. The 
updated P-T curves incorporate the results of RPV surveillance 
specimen testing utilizing criteria defined in RG 1.99, Revision 2. 
No change is being made to the way the P-T limits provide plant 
protection. No new modes of operation are involved. The changes do 
not necessitate physical alteration of the plant.
    RPV surveillance requirements. Deleting completed, one time 
surveillance requirements of section 4.6B and incorporating a 
discussion of the results in the Bases is an administrative change 
and therefore has no effect on previously analyzed accidents.
    SBLC Relief Valve Setpoint Testing. The testing requirements of 
TS section 4.4.A.2.c are enveloped by the current testing performed 
by Monticello's IST Program, which implements ASME Code Section XI, 
approved by 10 CFR 50.55a. The IST program requires all relief 
valves to be tested to their nameplate data setpoints. Any 
modification to a relief valve's nameplate data is controlled by the 
plant's configuration control process which would ensure the 
requirements of ASME Section XI are invoked as required by TS 
section 3.15. The IST program required by TS 4.15 ensures the SBLC 
relief valves will be properly tested for operability. Therefore, 
revising section 4.4.A.2.c to remove specific setpoints does not 
create the possibility of a new or different kind of accident, from 
any accident previously analyzed.
    The proposed amendment will not involve a significant reduction 
in the margin of safety.
    RPV P-T curve change. The proposed RPV P-T curve changes are 
designed to maintain the recommended safety factors specified in the 
ASME Boiler and Pressure Vessel Code, Section III, Appendix G, and 
10 CFR Part 50, Appendix G. The revised curves are based on current 
NRC guidelines utilizing actual RPV surveillance program tests 
results. The proposed changes shift the curves in a slightly more 
conservative direction thus maintaining or increasing the previous 
margins of safety.
    RPV surveillance requirements. Deleting completed, one time 
surveillance requirements from Section 4.6.B and incorporating a 
discussion of the results in the Bases is an administrative change 
and has no effect on any margin of safety.
    SBLC relief valve setpoint testing. The testing requirements of 
TS section 4.4.A.2.c are enveloped by the current testing performed 
by Monticello's IST Program, which implements ASME Code Section XI, 
approved by 10 CFR 50.55a. The IST program requires all relief 
valves to be tested to their nameplate data setpoints. Any 
modification to a relief valve's nameplate data is controlled by the 
plant's configuration control process which would ensure the 
requirements of ASME Section XI are invoked as required by TS 
section 3.15. The IST program required by TS 4.15 ensures the SBLC 
relief valves will be properly tested for operability. Therefore, 
revising section 4.4.A.2.c to remove specific setpoints will not 
reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: January 4, 1999.
    Description of amendment request: PECO Energy Company (PECO Energy) 
is requesting Technical Specifications (TS) changes which will revise 
the Administrative Section of TS pertaining to controlled access to 
High Radiation Areas, and the reporting dates for the Annual 
Occupational Radiation Exposure Report and the Annual Radioactive 
Effluent Release Report.
    The specific TS changes are as follows:
    TS Section 6.12, 6.12.1, and 6.12.2 will be changed to: clarify 
requirements; incorporate additional monitoring options (to allow 
dosimetry and video monitoring) for entry into high radiation areas; 
add the requirement that all individuals entering a high radiation area 
have knowledge of the dose rates in the area; and add the requirement 
that locked high radiation controls apply to each individual entering 
the area.
    TS Sections 6.9.1.4, 6.9.1.5(a), and 6.9.1.8 will be changed to: 
support changes to the NRC reporting dates;

[[Page 6707]]

reference 10 CFR 20.2206; delete current reporting dates, and correct a 
typographical error.
    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 TS changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The changes are administrative in nature and do not impact the 
operation, physical configuration, or function of plant equipment or 
systems. The changes do not impact the initiators or assumptions, of 
analyzed events, nor do they impact mitigation of accidents on 
transient events. Therefore, these changes do not increase the 
probability of occurrence of consequences of an accident previously 
evaluated in the SAR [Safety Analysis Report].
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes are administrative in nature and do not 
alter plant configuration, require that new equipment be installed, 
alter assumptions made about accidents previously evaluated, or 
impact the operation or function of plant equipment. Therefore, 
these changes do not create the possibility of a new or different 
kind of accident than previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The proposed changes are administrative in nature and do not 
impact any safety assumptions, or potentially reduce any margin of 
safety as described in the LGS TS basis. The proposed changes have 
no impact on any safety analysis assumptions. Therefore, these 
changes do not involve any 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: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.
    Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
PA 19101.
    NRC Project Director: William M. Dean.

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

    Date of amendment request: December 28, 1998.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to permit an increase in the 
allowable leak rate for the main steam isolation valves (MSIVs) and to 
delete the MSIV Sealing System. The main steam drain lines and the main 
condenser would be utilized as an alternate MSIV leakage treatment 
method.
    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 to TS Section 3.6.1.2 do not involve a 
change to structures, components, or systems that would affect the 
probability of an accident previously evaluated in the Hope Creek 
Updated Final Safety Analysis Report (UFSAR).
    The proposed changes involve eliminating the Main Steam 
Isolation Valve (MSIV) Steam Sealing System requirements from the 
TS. As described in Section 6.7 of the UFSAR, the MSIV Steam Sealing 
System is manually initiated in about 20 minutes following a design 
basis Loss of Coolant Accident (LOCA). Since the MSIV Steam Sealing 
System is operated only after an accident has occurred, these 
proposed changes have no effect on the probability of an accident. 
Since MSIV leakage and operation of the MSIV Steam Sealing System 
are included in the radiological analysis for the design basis LOCA 
as described in Section 15.6.5 of the UFSAR, the proposed changes 
will not affect the precursors of other analyzed accidents. Analysis 
of the affects of the proposed changes do, however, result in 
acceptable radiological consequences for the design basis LOCA 
previously evaluated in Section 15.6.5 of the UFSAR.
    Hope Creek has an inherent MSIV leakage treatment capability as 
discussed below. [Public Service Electric and Gas Company] PSE&G 
proposes to use the drain lines associated with the main steam lines 
and main turbine condenser as an alternative to the guidance in 
Regulatory Guide 1.96, ``Design of Main Steam Isolation Valve 
Leakage Control System For Boiling Water Nuclear Power Plants,'' 
Revision 0, May 1975, for MSIV leakage treatment. If approved, PSE&G 
will incorporate this alternative method in the appropriate 
operational procedures and Emergency Operating Procedures.
    The Boiling Water Reactor Owner's Group (BWROG) has evaluated 
the availability of main steam system piping and main condenser 
alternate pathways for processing MSIV leakage, and has determined 
that the probability of a near coincident LOCA and a seismic event 
is much smaller than for other plant safety risks. Accordingly, this 
proposed MSIV leakage treatment pathway will be available during and 
after a LOCA. Nevertheless, the BWROG has also determined that the 
design requirements applied to the Hope Creek main steam system 
piping and main condenser contain substantial margin, based on the 
original design requirements.
    In order to further justify the capability of the main steam 
piping and main condenser alternate treatment pathway, the BWROG has 
reviewed limited earthquake experience data on the performance of 
non-seismically designed piping and condensers during past 
earthquakes. As summarized in General Electric (GE) Report, ``BWROG 
Report for Increasing MSIV Leakage Rate Limits and Elimination of 
Leakage Control Systems,'' NEDC-31858P, Revision 2, submitted to the 
[U.S. Nuclear Regulatory Commission] NRC by BWROG letter dated 
October 4, 1993, this study concluded that the possibility of a 
failure that could cause a loss of steam or condensate in Boiling 
Water Reactor (BWR) main steam piping or condensers in the event of 
a design basis (i.e., safe shutdown) earthquake is highly unlikely, 
and that such a failure would also be contrary to a large body of 
historical earthquake experience data, and thus unprecedented.
    PSE&G has performed a verification of seismic adequacy of the 
Hope Creek main stream piping and main condenser consistent with the 
guidelines discussed in NEDC-31858P, Revision 2, to provide 
reasonable assurance of the structural integrity of these 
components. This evaluation, ``Hope Creek Nuclear Plant Main Steam 
Isolation System Alternate Leakage Treatment Pathway Seismic 
Evaluation,'' clearly demonstrates that the MSIV leakage treatment 
drain pathway meets the intent of 10 CFR 100 Appendix A, with 
regards to seismic qualification. Except for the requirement to 
establish a proper flow path from the MSIVs to the condenser, the 
proposed method is passive and does not require any additional logic 
control and interlocks. The method proposed for MSIV leakage 
treatment is consistent with the philosophy of protection by 
multiple barriers used in containment design for limiting fission 
product release to the environment.
    A plant-specific radiological analysis has also been performed 
in accordance with NEDC-31858P, Revision 2, to assess the effects of 
the proposed increase to the allowable MSIV leakage rate in terms of 
Main Control Room (MCR) and off-site doses following a postulated 
design basis LOCA. This analysis utilizes the hold-up volumes of the 
main steam piping and condenser as an alternate method for treating 
the MSIV leakage. As discussed earlier, there is reasonable 
assurance that the main steam piping and condenser will remain 
intact following a design basis earthquake. The radiological 
analysis uses standard conservative assumptions for the radiological 
source term consistent with Regulatory Guide (RG) 1.3, ``Assumptions 
Used for Evaluating the Potential Radiological Consequences of a 
Loss-Of-Coolant Accident for Boiling Water Reactor,'' Revision 2, 
dated April 1974.
    The analysis results demonstrate that dose contributions from 
the proposed MSIV leakage rate limit of 200 scfh per steam line, not 
to exceed a total of 400 scfh for all four

[[Page 6708]]

main steam lines, and from the proposed deletion of the MSIV Steam 
Sealing System, result in an acceptable increase to the LOCA doses 
previously evaluated against the regulatory limits for the off-site 
doses and MCR doses contained in 10 CFR 100 and 10 CFR 50, Appendix 
A, General Design Criterion (GDC) 19, respectively. However, the 
calculation methodology for the revised dose exposures were 
performed in a manner that included more conservative design basis 
assumptions (e.g., inclusion of system response times, and increased 
allowable leakage rates) than in the existing Hope Creek licensing 
basis.
    The whole body doses at the low population zone (LPZ) outer 
boundary and MCR increase from about 0.2 rem to 0.6 rem and from 
0.04 rem to 0.09 rem, respectively. These increases are not 
significant since the revised doses are small fractions of the 
regulatory limits of 25 rem and 5 rem, respectively. The associated 
whole body dose at the exclusion area outer boundary (EAB) increases 
from about 1.3 rem to 2.6 rem, which is well within the regulatory 
limit of 25 rem. The revised thyroid dose at the LPZ outer boundary 
increases from about 18 rem to 36 rem, which is well within the 
regulatory limit of 300 rem. The revised thyroid dose at the EAB 
decreases from about 175 rem to 121 rem (due to plate out on the 
steam piping and condenser), which is within the regulatory limit of 
300 rem. However, the MCR thyroid dose increases from about 0.3 rem 
to 5.0 rem, which is well within the regulatory limit of 30 rem. 
Additionally, the MCR beta skin dose increases from about 0.9 rem to 
1.6, which is well within the regulatory limit of 30 rem.
    The resulting revised thyroid doses discussed above are 
dominated by the inorganic radioactive iodine fractions of the 
accident source term used in this analysis. More than 95% of the 
initial radioactive iodine inventory is assumed to be in the form of 
inorganic species in accordance with the guidance in Regulatory 
Guide 1.3. However, NUREG-1465, ``Accident Source Terms for Light-
Water Nuclear Power Plants,'' identifies that at least 95% of the 
iodine entering containment would be in the form of particulate 
iodine. Accordingly, the calculated doses discussed above are 
considered to be highly conservative relative to realistic 
radiological source terms resulting from a postulated LOCA.
    In summary, the proposed changes discussed above do not result 
in a significant increase in the radiological consequences of a LOCA 
when the same assumptions and methods specified in the UFSAR are 
used, recognizing that radiological consequences calculated in the 
UFSAR and for these proposed changes are significantly higher than 
those using more realistic assumptions and methods. Nevertheless, 
the calculated off-site and MCR doses resulting from a LOCA remain 
well below the regulatory limits. Although the revised LOCA doses 
are higher for low MSIV leakage rates, the effectiveness of the 
proposed alternate treatment method, even for leakage rates greater 
than the proposed increase in the MSIV allowable leak rate, ensures 
that off-site and MCR dose limits are not exceeded.
    The proposed change to TS Table 3.6.3-1 involves the deletion of 
MSIV Steam Sealing valves and associated main steam line drain 
valves from the list of primary containment isolation valves. This 
proposed change is consistent with the proposed deletion of the MSIV 
Steam Sealing System. The MSIV Steam Sealing System lines and main 
steam line drain valves that are connected to the main steam piping 
will be welded and/or capped closed to assure primary containment 
integrity is maintained. The welding and post weld examination 
procedures will be in accordance with American Society of Mechanical 
Engineers (ASME) Code, Section III requirements. These welds and/or 
caps will be periodically tested as part of the Containment 
Integrated Leak Rate Test (CILRT). This proposed change does not 
involve an increase in the probability of equipment malfunction 
previously evaluated in the UFSAR. This proposed change has no 
effect on the consequences of an accident since the MSIV Steam 
Sealing lines and associated main steam line drain valves will be 
welded an/or cap closed, thus assuring that the containment 
integrity, isolation, and leak test capability are not compromised.
    Therefore, as discussed above, the proposed changes do not 
involve a significant increase in the probability or consequences 
from 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.
    Although the proposed changes will introduce and take credit for 
a new level of operational performance for existing plant systems 
and components that have not been previously evaluated in the 
accident analysis, the affect on this equipment has been evaluated 
and found to provide an acceptable level of reliability that will 
provide the required level of protection. This conclusion is based 
on the evaluation performed in NEDC-31858P, Revision 2, and the 
seismic evaluation of the proposed MSIV leakage treatment pathway. 
Therefore, reliance on different equipment than previously assumed 
to mitigate the consequences of an accident does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The BWROG evaluated MSIV performance and concluded that MSIV 
leakage rates up to 200 scfh per line will not inhibit the 
capability and isolation performance of the MSIVs to effectively 
isolate the primary containment. Implementation of the proposed 
changes will not result in modifications that could adversely impact 
the operability of the MSIVs. The LOCA has been analyzed using the 
main steam piping and main condenser as a treatment method to 
process MSIV leakage at the proposed maximum rate of 200 scfh per 
main steam line, not to exceed 400 scfh total for all four main 
steam lines. Therefore, the proposed change to increase the allowed 
MSIV leakage rate does not create any new or different kind of 
accident from any accident previously evaluated.
    The proposed change to eliminate the MSIV Steam Sealing System 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated because the removal 
of the MSIV Steam Sealing System does not affect any of the 
remaining Hope Creek systems, and the LOCA has been re-analyzed 
using the proposed alternate method to process MSIV leakage. The 
associated proposed change to delete the MSIV Steam Sealing 
isolation valves and associated main steam line drain valves from TS 
Table 3.6.3-1 does not create the possibility of a new or different 
kind of accident, since the affected main steam piping will be 
welded and/or capped closed to assure that the primary containment 
integrity, isolation, and leak testing capability are not 
compromised.
    Therefore, as discussed above, the proposed changes do not 
create the possibility for any new or different kind of accident 
from any accident previously evaluated.
    (3) The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change to TS Section 3.6.1.2 to increase the MSIV 
allowable leakage does not involve a significant reduction in the 
margin of safety. As discussed in the current Bases for TS Section 
3/4.6.1.2, the allowable leak rate limit specified for the MSIVs is 
used to quantify a maximum amount of leakage assumed to bypass 
primary containment in the LOCA radiological analysis. Accordingly, 
results of the re-analysis supporting these proposed changes are 
evaluated against the dose limits contained in 10 CFR 100 for the 
off-site doses, and 10 CFR 50, Appendix A, GDC 19, for the MCR 
doses. As discussed above, sufficient margin relative to the 
regulatory limits is maintained even when assumptions and methods 
(e.g., RG 1.3) that are considered highly conservative relative to 
more realistic assumptions and methods, are used in the analysis.
    Results of the radiological analysis demonstrate that the 
proposed changes do not involve a significant reduction in the 
margin of safety. The whole body doses, in terms of margin of 
safety, are insignificantly reduced by 1.6% at the LPZ, 1.0% in the 
MCR, and by 5.2% at the EAB. The margin of safety for thyroid doses 
is reduced by 6.13% at the LPZ and 15.7% in the MCR, but is actually 
increased by 17.3% at the EAB. The margin of safety for beta dose is 
insignificantly reduced by 2.4% in the MCR. These reductions in the 
margin of safety are not significant since the revised calculated 
doses are highly conservative yet remain well below the regulatory 
limits, and therefore a substantial margin to the regulatory limits 
is maintained.
    Furthermore, while the proposed changes will result in a 
calculated reduction in the margin of safety, this reduction is not 
significant when considering the increased reliability and 
capability of the proposed MSIV leakage treatment system. The 
resulting revised thyroid doses discussed above are dominated by the 
inorganic radioactive iodine fractions of the accident source term 
used in this analysis. More than 95% of the initial radioactive 
iodine inventory is assumed to be in the form of inorganic species 
in accordance with the guidance in Regulatory Guide 1.3. However,

[[Page 6709]]

NUREG-1465, ``Accident Source Terms for Light-Water Nuclear Power 
Plants,'' identifies that at least 95% of the iodine entering 
containment would be in the form of particulate iodine. Accordingly, 
the calculated doses discussed above are considered to be highly 
conservative relative to realistic radiological source terms 
resulting from a postulated LOCA.
    The proposed change to eliminate the MSIV Steam Sealing System 
from TS does not reduce the margin of safety. In fact, the overall 
margin of safety is increased. The function of this system for MSIV 
leakage treatment will be replaced by alternate main steam drain 
lines and condenser equipment. This treatment method is effective in 
reducing the dose consequences of MSIV leakage over an expanded 
operating range compared to the capability of the MSIV Steam Sealing 
System and will, thereby, resolve the safety concern that the MSIV 
Steam Sealing System will not function at MSIV leakage rates higher 
than the Steam Sealing System's design capacity. Except for the 
requirement to establish a proper flow path from the MSIVs to the 
condenser, the proposed method is passive and does not require any 
new logic control and interlocks. This proposed method is consistent 
with the philosophy of protection by multiple barriers used in 
containment design for limiting fission product release to the 
environment. Furthermore, as previously identified, based on the 
evaluations discussed in NEDC-31858P, Revision 2, and the seismic 
evaluation performed for Hope Creek, the design of the MSIV leakage 
treatment pathway meets the intent of the 10 CFR 100, Appendix A, 
requirement for seismic qualification. Therefore, the proposed 
method is highly reliable and effective for MSIV leakage treatment.
    The revised calculated LOCA doses remain within the regulatory 
limits for the off-site and the MCR doses. Furthermore, the revised 
calculation shows that MSIV leakage rates greater than 200 scfh for 
all four main steam lines would not exceed the regulatory limits. 
Therefore, the proposed method maintains a margin of safety for 
mitigating the radiological consequences of MSIV leakage beyond the 
proposed TS leakage rate limit of 200 scfh per main steam line, not 
to exceed a total of 400 scfh for all four main steam lines.
    The proposed change to delete MSIV Steam Sealing valves from TS 
Table 3.6-3-1 [3.6-3-1] does not reduce the margin of safety. Welded 
and/or capped closure of the MSIV Steam Sealing lines assures that 
the primary containment integrity and leak testing capability are 
not compromised. These welds and/or caps will be periodically leak 
tested as part of the CILRT. Therefore, the proposed deletion of the 
MSIV Steam Sealing System isolation valves does not involve a 
reduction in a margin of safety.
    Accordingly, based on the above reasons, the proposed changes do 
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: 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: William M. Dean.

Public Service Electric & Gas Company, Docket Nos. 50-272, Salem 
Nuclear Generating Station, Unit No. 1, Salem County, New Jersey

    Date of amendment request: January 15, 1999.
    Description of amendment request: The proposed amendment would 
allow a one-time extension of the Technical Specification (TS) 
surveillance interval to the end of fuel cycle 13 for certain TS 
surveillance requirements (SRs). Specifically, (1) SR 4.3.2.1.3 
requires the instrumentation response time and sequence testing of each 
engineered safety features actuation system (ESFAS) function at least 
once per 18 months, (2) SRs 4.8.2.3.2.f and 4.8.2.5.2.d require that 
the 125 volt DC and the 28 volt DC distribution system batteries, 
respectively, be capacity service tested at least once per 18 months, 
during shutdown, (3) SR 4.8.3.1.a.1.a and 4.8.3.1.a.1.b require a 
channel calibration and integrated system functional test for one 4.16 
kilovolt reactor coolant pump circuit at least once per 18 months such 
that all circuits are tested at least once per 72 months, (4) SR 
4.1.2.2.c requires testing to verify that each automatic valve in the 
reactivity control system flow path actuate on a safety injection (SI) 
test signal at least once per 18 months during shutdown, (50 SRs 
4.3.1.1, Table 4.3-1, 4.3.2.1.1, Table 4.3-2, 4.3.3.5, Table 4.3-6, and 
4.3.3.7, Table 4.3-11 require, in part, the channel calibration of 
pressurizer water level, pressurizer water level-high, and containment 
water level-wide range, the manual solid-state protection system (SSPS) 
functional input check, and the ESFAS manual initiation channel 
functional test every 18 months, (6) SR 4.5.1.d requires testing to 
verify each accumulator isolation valve opens automatically on an SI 
test signal at least once per 18 months, (7) SR 4.5.2.e.1 requires 
testing to verify that each automatic valve in the emergency core 
cooling system (ECCS) flow path actuates on an SI test signal at least 
once per 18 months, (8) SR 4.7.6.1.d.2 requires the control room 
emergency air conditioning system to automatically actuate in the 
pressurization mode on an SI test signal or control room intake high 
radiation test signal at least once per 18 months, (9) SR 4.7.10.b 
requires each automatic valve in the chilled water loop to actuate on 
an SI signal at least once per 18 months. Further, SR 4.8.1.1.2.d.7 
requires a test to verify that each emergency diesel generator operates 
for at least 24 hours every 18 months, and SR 4.8.2.5.2.c.2 requires 
that the 125 volt DC battery connections be verified clean, tight, and 
coated with anti-corrosion material at least once per 18 months. 
Because of the length of the last outage and delays in restart, the SRs 
will be overdue prior to reaching the next refueling outage (1R13). The 
SRs are to be completed during the 1R13 outage, prior to returning the 
unit to Mode 4 (hot shutdown) upon outage completion. The proposed 
amendment also make some administrative and editorial changes on some 
of the pages that will be affected by above SR interval extensions.
    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:

4.3.2.1.3 (Instrumentation, Engineered Safety Feature Actuation System 
Instrumentation)

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Deferral of the surveillance requirement does not involve any 
physical changes to the plant nor does it change the way the plant 
is operated. Thus the proposal does not increase the probability of 
an accident previously evaluated.
    The SEC [safeguard equipment control] automatic self-test 
feature, the monthly functional surveillance testing and the 
positive surveillance testing history provide sufficient assurance 
of the operability of the system. These features also provide 
assurance that a degraded condition, if it did occur, would be 
detected.
    Thus, it is reasonable to conclude that this proposal represents 
no significant increase in the consequences of an accident 
previously analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident form any accident previously 
evaluated.
    Deferral of the surveillance requirement does not involve any 
physical changes to the plant nor does it change the way the plant 
is operated.
    Thus, it can be concluded that deferring the surveillance 
requirement to the refueling outage cannot create the possibility of 
a different kind of accident from any accident previously evaluated.

[[Page 6710]]

    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Deferral of the surveillance requirement does not involve any 
physical changes to the plant nor does it change the way the plant 
is operated. The self-test feature and the monthly functional 
testing will provide reasonable assurance that the SECs will remain 
operable during the few weeks of deferral to the refueling outage. 
Also the ability to detect a degraded condition in the SEC will not 
be affected during the deferral period.
    Therefore, the plant's response to accident conditions during 
the period of deferral will not be affected.
    Thus, it can be reasonably concluded that this proposal to amend 
the Salem Unit 1 Technical Specifications, on a one-time basis, to 
defer surveillance requirement 4.3.2.1.3 does not involve a 
significant reduction in any margin of safety.

4.8.2.3.2.f, (Electrical Power Systems, 125 Volt D.C. Distribution), 
and 4.8.2.5.2.c.2 and 4.8.2.5.2.d (Electrical Power Systems, 28 Volt 
D.C. Distribution)

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The deferral of the battery service tests to the refueling 
outage does not involve any physical changes to the power plant or 
to the manner in which the power plant is operated. Therefore, the 
probability of an accident previously evaluated is not increased.
    Weekly and quarterly testing and performance monitoring by the 
system manager along with the current condition of the batteries 
(past test results demonstrating above 100% capacity) provide 
assurance that battery condition and performance will not 
deteriorate during the deferral period. Other positive industry 
experience for similar batteries on 24 month cycles also support 
this assurance. Therefore, the consequences of a loss of power 
accident will not be increased due to the deferral of the 
surveillance requirements.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The deferral of the battery service tests to the refueling 
outage does not involve any physical changes to the power plant or 
to the manner in which the power plant is operated. No new failure 
mechanisms will be introduced by the surveillance deferral. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The deferral of the battery service tests to the refueling 
outage does not involve any physical changes to the power plant or 
to the manner in which the power plant is operated. Continuing 
weekly and quarterly testing and performance monitoring along with 
the current condition of the batteries provides assurance that 
battery condition and performance will be acceptable during the 
deferral period and that any degradation that may occur will be 
detected. Therefore, the plant's response to accident conditions 
during the period of deferral will not be affected.
    Thus, it can be reasonably concluded that this proposal to amend 
the Salem Unit 1 Technical Specifications, on a one-time basis, to 
defer surveillance requirements 4.8.2.3.2.f, 4.8.2.5.2.c.2 and 
4.8.2.5.2.d does not involve a significant reduction in any margin 
of safety.

4.8.3.1.a.l.a, 4.8.3.1.a.l.b (Electric Power Systems, Electrical 
Equipment Protective Devices)

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The deferral of inspection, calibration and meggering of 1A, 1B, 
1C 460VAC transformer relays and current transformers (CT's); and 
inspection, calibration and meggering of 1F 4KV Bus Overload Relays 
to the refueling outage does not involve any physical changes to the 
power plant or to the manner in which the power plant is operated. 
Therefore, the probability of an accident previously evaluated is 
not increased.
    The condition of the equipment as found for the three most 
recent completed surveillances (i.e. no failures or equipment 
problems found, no repair actions required, and test results 
satisfactory in all cases) provides assurance that equipment 
condition and performance will be acceptable during the deferral 
period. The subject equipment has performed well over the past 
several years and has demonstrated satisfactory stability and 
reliability. The plant's response to accident conditions during the 
period of deferral will not be affected. Therefore, the consequences 
of an accident previously evaluated will not be increased due to the 
deferral of the surveillance requirements.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The deferral of inspection, calibration and meggering of 1A, 1B, 
1C 460VAC transformer relays and current transformers (CT's); and 
inspection, calibration and meggering of 1F 4KV Bus Overload Relays 
to the refueling outage does not involve any physical changes to the 
power plant or to the manner in which the power plant is operated. 
No new failure mechanisms will be introduced by the surveillance 
deferral. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The deferral of inspection, calibration and meggering of 1A, 1B, 
1C 460VAC transformer relays and current transformers (CT's); and 
inspection, calibration and meggering of 1F 4KV Bus Overload Relays 
to the refueling outage does not involve any physical changes to the 
power plant or to the manner in which the power plant is operated. 
The results of previous tests which demonstrate the reliable and 
stable operation of the equipment over recent years provides 
assurance that the equipment will operate as designed during the 
deferral period. The plant's response to accident conditions during 
the period of deferral will not be affected.
    Thus, it can be reasonably concluded that this proposal to amend 
the Salem Unit 1 Technical Specifications, on a one-time basis, to 
defer surveillance requirements 4.8.3.1.a.l.a and 4.8.3.1.a.l.b does 
not involve a significant reduction in any margin of safety.

4.1.2.2.c (Reactivity Control Systems, Flow Paths--Operating), 
4.3.1.1.1, Table 4.3-1 (Reactor Trip System Instrumentation--
Surveillance Requirements); 4.3.2.1.1, Table 4.3-2 (Engineered Safety 
Feature Actuation System Instrumentation--Surveillance Requirements); 
4.5.1.d (Emergency Core Cooling Systems, Accumulators); 4.5.2.e.1 
(Emergency Core Cooling Systems, ECCS Subsystems--Tave [greater than or 
equal to] 350  deg.F); 4.7.6.1.d.2 (Plan Systems, Control Room 
Emergency Air Conditioning System); and 4.7.10.b (Plant Systems, 
Chilled Water System--Auxiliary Building Subsystem)

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The deferral of the Manual Safety Injection (SI) surveillance 
test to the refueling outage does not involve any physical changes 
to the power plant or to the manner in which the power plant is 
operated. Therefore, the probability of an accident previously 
evaluated is not increased.
    Other surveillance testing provides assurance that the equipment 
will be reliable during the short deferral period. This testing, in 
conjunction with successful previous SI test results assure that the 
equipment will function properly during the short deferral period. 
Therefore, the consequences of an accident previously evaluated will 
not be increased due to the deferral of the surveillance 
requirements.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The deferral of the Manual Safety Injection (SI) surveillance 
test to the refueling outage does not involve any physical changes 
to the power plant or to the manner in which the power plant is 
operated. No new failure mechanisms will be introduced by the 
surveillance deferral. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The deferral of the Manual Safety Injection (SI) surveillance 
test to the refueling outage does not involve any physical changes 
to the power plant or to the manner in which the power plant is 
operated. Other surveillance testing in conjunction with successful 
previous SI test results provides assurance that the equipment will 
be reliable during the short deferral period. The plant's response 
to accident conditions during the period of deferral will not be 
affected.
    Thus, it can be concluded that this proposal to amend the Salem 
Unit 1 Technical Specifications, on a one-time basis, to defer 
surveillance requirements 4.1.2.2.c; 4.3.1.1.1, Table 4.3-1; 
4.3.2.1.1, Table 4.3-2;

[[Page 6711]]

4.5.1.d; 4.5.2.e.1; 4.7.6.1.d.2; and 4.7.10.b does not involve a 
significant reduction in any margin of safety.

4.8.1.1.2.d.7 (Electrical Power Systems, A.C. Power Sources) Diesel 
Generator 24 Hour Endurance Run)

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Deferral of performance of the diesel generator 24 hour 
endurance runs to 1R13 does not involve any physical changes to the 
power plant or to the manner in which the power plant is operated. 
Therefore, the probability of an accident previously evaluated is 
not increased.
    Based of the favorable history for previous endurance runs for 
the six Sale Unit 1 & 2 emergency diesel generators, continued 
normal monthly surveillance testing and the trending of engine and 
generator parameters, diesel generator operability can be assured 
during the deferral period. Therefore, the consequences of an 
accident previously evaluated will not be increased due to the 
deferral of the surveillance requirements.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Deferral of performance of the diesel generator 24 hour 
endurance runs to 1R13. does not involve any physical changes to the 
power plant or to the manner in which the power plant is operated. 
No new failure mechanisms will be introduced by the surveillance 
deferral. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Deferral of performance of the diesel generator 24 hour 
endurance runs to 1R13 does not involve any physical changes to the 
power plant or other manner in which the power plant is operated. 
Satisfactory endurance run history, other surveillance testing and 
performance monitoring assures diesel generator operability during 
the deferral period.
    The plant's response to accident conditions during the period of 
deferral will not be affected.
    Thus, it can be conducted that this proposal to amend the Salem 
Unit 1 Technical Specifications, on a one-time basis, to defer 
surveillance requirement 4.8.1.1.2.d.7 does not involve a 
significant reduction in any margin of safety.

4.3.1.1.1, Table 4.3-1 (Reactor Trip System Instrumentation-
Surveillance Requirements); 4.3.3.5, Table 4.3-6 (Remote Shutdown 
Monitoring Instrumentation Surveillance Requirements); 4.3.3.7, Table 
4.3-11 (Surveillance Requirements for Accident Monitoring 
Instrumentation)

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Deferral of calibration of Pressurizer Level Channel 1, and the 
Containment Sump Level devices to 1R13 does not involve any physical 
changes to the power plant or to the manner in which the power plant 
is operated. Therefore, the probability of an accident previously 
evaluated is not increased.
    Review of trends of the level channels during the current 
operating cycle and continued monitoring of the channels provides 
reasonable assurance that the channels will perform their design 
function during the deferral period. Therefore, the consequences of 
an accident previously evaluated will not be increased due to the 
deferral of the surveillance requirements.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Deferral of calibration of Pressurizer Level Channel 1, and the 
Containment Sump Level devices to 1R13 does not involve any physical 
changes to the power plant or to the manner in which the power plant 
is operated. No new failure mechanisms will be introduced by the 
surveillance deferral. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Deferral of calibration of Pressurizer Level Channel 1, and the 
Containment Sump Level devices to 1R13 does not involve any physical 
changes to the power plant or to the manner in which the power plant 
is operated. Review of trends of the level channels during the 
current operating cycle and continued monitoring provides reasonable 
assurance that the channels will perform their design function 
during the deferral period. There will be no effect on the response 
to accident conditions during the period of deferral.
    Thus, it can be concluded that this proposal to amend the Salem 
Unit 1 Technical Specifications, one a one-time basis, to defer 
surveillance requirements 4.3.1.1.1, Table 4.3-1, item 11; 4.3.3.5, 
Table 4.3-6, item 2; and 4.3.3.7, Table 4.3-11, items 4 and 17 does 
not involve a significant reduction in any margin of safety.

Administrative and Editorial Change

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes are administrative or editorial and do not 
involve any physical changes to the plant. The administrative 
changes and editorial changes do not delete any existing 
surveillance requirements or delete any requirements from the 
Limiting Condition for Operations (LCOs) or Action Statements and 
therefore do not reduce the actions that are currently taken to 
demonstrate operability of plant structures, systems, or components 
(SSCs). The additional surveillance requirement that is being added 
including the new surveillance corrects a past administrative error 
and should have been incorporated within the Tech Specs as part of 
an approved Amendment. This change will provide additional assurance 
that SSCs perform their intended safety functions. Surveillance 
testing has been and is currently being performed for the 
surveillance requirement that should have been incorporated and is 
now administratively being added to the Tech Specs. Since these 
changes do not modify any SSCs or reduce the current requirements 
for demonstrating operability of these SSCs, the proposed changes to 
the Tech Specs do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the Tech Specs are administrative and 
editorial corrections that do not affect the ability of the plant 
systems to meet their current Tech Spec requirements or design basis 
functions. There is no reduction in the current surveillance 
requirements required to demonstrate the operability of plant SSCs. 
These changes also do not involve any physical changes to plant 
SSCs. Therefore the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes are administrative and editorial 
corrections that do not affect the ability of plant SSCs to perform 
their design basis accident functions. There is no reduction in the 
current surveillance requirements required to demonstrate the 
operability of plant SSCs. Therefore, the proposed changes do 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: 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: William M. Dean.

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 4, 1998 as modified December 
7, 1998.
    Description of amendment requests: The proposed amendment would 
modify the Technical Specifications (TS) to increase the allowed as-
found pressurizer safety valve setpoint tolerance from +/-1 percent to 
+3/-2 percent.

[[Page 6712]]

    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.
    All Updated Final Safety Analysis Report (UFSAR) Chapter 15 
events have been evaluated to determine the impact of the increases 
in as found Pressurizer Safety Valve (PSV) tolerance from +1% and 
-1% to +3% and -2%. The events that result in challenging the 
opening of the PSVs are Loss of Condenser Vacuum With and Without 
Single Failure, Loss of Normal Feedwater Flow, Feedwater System Pipe 
Breaks, Total Loss of Reactor Coolant System (RCS) Flow, 
Uncontrolled Control Element Assembly (CEA) Withdrawal, CEA 
Ejection, Chemical and Volume Control System (CVCS) Malfunction With 
and Without Single Failure, Inadvertent Emergency Core Cooling 
System (ECCS) Actuation With and Without Single Failure, and 
Inadvertent Opening of a PSV. Of these, the limiting events are the 
Loss of Condenser Vacuum (LOCV), Loss of Condenser Vacuum With a 
Concurrent Single Failure of an Active Component (LOCVsf), CVCS 
Malfunction, CVCS Malfunction With a Concurrent Single Failure of an 
Active Component, and Feedwater System Pipe Breaks. These limiting 
events have been reanalyzed for the wider PSV tolerance. For all the 
reanalyzed events it is assumed that plant operation is maintained 
at a maximum pressurizer level of 57%. For the CVCS Malfunction With 
and Without Single Failure Events and the Inadvertent ECCS Actuation 
With and Without Single Failure Events, it is also assumed that the 
operator can respond within 15 minutes to mitigate the event.
    The change in as found PSV tolerance from -1% to -2% results in 
the earlier opening of the PSVs for the analyzed events. To 
compensate for this earlier opening of the PSVs the high pressurizer 
pressure trip analysis setpoint was reduced from 2437 psia (non-
harsh environment) and 2450 (harsh environment) to 2410 psia (non-
harsh environment) and 2434 (harsh environment). These setpoint 
changes insure that the high pressurizer pressure trip is actuated 
sufficiently early before the opening of the PSVs such that no 
liquid is released through the PSVs. Therefore, the change to the 
PSV negative tolerance does not result in a significant increase in 
the probability or consequences of any previously evaluated 
accident.
    The change in PSV as found tolerance from +1% to +3% results in 
later opening of the PSVs for the analyzed events. The PSV actuation 
to mitigate the consequences of the analyzed accidents are thus 
delayed. However, the lowering of the high pressurizer pressure trip 
setpoint, as discussed above, mitigates the increase in peak primary 
pressure and assures that no liquid is released through the PSVs. 
Therefore, this change to the PSV positive tolerance does not result 
in a significant increase in the probability or consequences of any 
previously analyzed design basis event.
    There are no other changes to the plant equipment or operation 
which could create an increase in the probability or consequences of 
any event previously evaluated.
    Therefore, operation in accordance with this proposed change 
will not involve a significant increase in the probability or 
consequences of any previously evaluated accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Operation in accordance with this proposed change will not 
involve any change to plant equipment or operation which could 
create a new or different kind of accident. The as-left PSV 
tolerance will continue to remain at +/-1%. The change in as-found 
tolerance of the PSVs to -2% and +3% will not introduce the 
possibility of a new or different kind of accident because 
evaluation of the design basis events shows that no water is 
expected to be released through the PSVs.
    There are no other changes to the plant equipment or operation 
which could create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, this 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.
    Operation in accordance with this proposed change will not 
change the manner in which safety limits, limiting safety settings, 
or limiting conditions for operation are determined. The acceptance 
criteria for all of the events reanalyzed include an appropriate 
margin of safety.
    There are no changes to the acceptance criteria nor are the 
acceptance criteria exceeded for these events assuming plant 
operation at a maximum pressurizer level of 57% and operator 
response time of 15 minutes for the CVCS Malfunction With and 
Without Single Failure Events and the Inadvertent ECCS Actuation 
With and Without Single Failure Events.
    Therefore, this 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 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, P.O. Box 19557, Irvine, California 92713.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, P.O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

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

    Date of application for amendments: January 15, 1999 (TS 98-09).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Technical Specification (TS) requirements by 
relocating Section 3.3.3.3, ``Seismic Instrumentation,'' to the SQN 
Technical Requirements Manual (TRM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    TVA has concluded that operation of SQN Units 1 and 2, in 
accordance with the proposed change to the TS, does not involve a 
significant hazards consideration. TVA's conclusion is based on its 
evaluation, in accordance with 10 CFR 50.91(a)(1), of the three 
standards set forth in 10 CFR 50.92(c).
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    A. The proposed revision to the TS relocates the requirements 
for SQN seismic instrumentation without changing the current 
requirements. TVA does not consider the instrumentation to be the 
source of any accident; therefore, this administrative relocation of 
the requirements will not increase the possibility of an accident. 
The capability of the seismic instrumentation will continue to 
provide the same function of data collection. Changes to the 
relocated requirements will be processed, in accordance with 10 CFR 
50.59, to ensure the seismic instrumentation functions will be 
properly maintained. Therefore, the proposed relocation of the 
seismic instrumentation requirements will not increase the 
consequences of an accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The SQN seismic instrumentation is used to record data for use 
in evaluating the effect of a seismic event. This instrumentation is 
not associated with accident mitigation or previously evaluated 
accidents and would not be the initiator of any new or different 
kind of accident. The proposed change does not alter the current 
functions of SQN's seismic instrumentation; therefore, this proposed 
change will not create the possibility of a new or different kind of 
accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.

[[Page 6713]]

    The requirements for SQN's seismic instrumentation are unchanged 
by the proposed relocation of the requirements to the SQN TRM. The 
function of the seismic instrumentation and SRs to ensure 
operability of the instrumentation remains unchanged. Any future 
changes to these requirements will be evaluated, in accordance with 
10 CFR 50.59, to ensure acceptability and NRC review as required. 
Accordingly, the proposed change will not result in a reduction in a 
margin of safety.

    The NRC 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: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Project Director: Cecil O. Thomas.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: April 23, 1998.
    Description of amendment request: The amendment request proposes 
changes to the existing requirements for the RHR Service Water (RHRSW), 
Station Service Water (SSW) and Alternate Cooling Tower Systems (ACS) 
as identified in Technical Specifications (TS) 4.5.C and 3/4.5.D.
    Specifically, the changes proposed are as follows:
    (1) Specifications 3.5.D.3 and 4.5.D.3: This requirement is revised 
to delete the existing allowance for 7 days of operation after both SSW 
subsystems are made or found to be inoperable.
    (2) Specification 4.5.C.1 and Specification 4.5.D.1: These 
requirements have been revised to relocate testing information related 
to pump flow and pressure testing characteristics for the RHRSW and SSW 
Systems, respectively, to the TRM.
    (3) Specifications 3.5.D.1, 3.5.D.2, 3.5.D.3, 4.5.D.2, 4.5.D.3 and 
associated Bases: All reference to SSW ``subsystem'' has been replaced 
by ``essential equipment cooling loop'' to more accurately reflect 
VYNPS design and operation. In addition, certain operability 
clarifications have been made to the Bases relative to affected 
Specifications.
    (4) Bases for Specifications 3.5.D: The Bases have been revised to 
omit statements which imply that the ACS could provide adequate heat 
removal following a postulated accident. Other Bases additions have 
been made which include certain operability clarifications relative to 
affected Specifications.
    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:
    For change No. 1:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change deletes the existing allowance for 7 days of 
operation after both Station Service Water (SSW) subsystems are made 
or found to be inoperable. At least one subsystem of the SSW System 
is required to be operable to mitigate the consequences of a design 
basis accident. Therefore, with both subsystems inoperable, the unit 
is required to shut down. Current Technical Specifications (TS) 
erroneously allow 7 days of operation after both SSW subsystems are 
made or found to be inoperable before requiring that the reactor be 
placed in cold shutdown within 24 hours. This allowance is 
incorrectly based on the assumption that the Alternate Cooling Tower 
System (ACS) is able to fulfill the post-accident heat removal 
requirements when both SSW Subsystems are made or found to be 
inoperable. Since the ACS is not capable of fulfilling this backup 
role, the allowance for seven days of operation with both SSW 
Subsystems inoperable is removed, and a requirement to shutdown the 
unit is provided in its place. This proposed change deletes the 
allowance for 7 days of operation in this condition, and instead 
requires an orderly shutdown to be initiated and the reactor to be 
placed in cold shutdown within 24 hours. Since the same amount of 
time is allowed to conduct the required shutdown, this change will 
not significantly increase the consequences of any previously 
analyzed accident. In addition, the SSW system is not considered to 
be the initiator of any previously analyzed accident. Therefore, 
this change will not significantly increase the probability or 
consequences of any previously analyzed accident.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This change will not physically alter the plant (no new or 
different types of equipment will be installed). The changes in 
methods governing normal plant operation are consistent with the 
current safety analysis assumptions. Therefore, this change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    This change deletes the existing allowance for 7 days of 
operation after both SSW subsystems are made or found to be 
inoperable. At least one subsystem of the SSW System is required to 
be operable to mitigate the consequences of a design basis accident. 
Therefore, with both subsystems inoperable, the unit is required to 
be shut down. Current TS requirements erroneously allow 7 days of 
operation after both the SSW subsystems are made or found to be 
inoperable before requiring that the reactor be placed in cold 
shutdown within 24 hours. This allowance is incorrectly based on the 
assumption that the ACS is able to fulfill the post-accident heat 
removal requirements when both SSW Subsystems are inoperable. Since 
the ACS is not capable of fulfilling this backup role, the allowance 
for seven days of operation with both SSW Subsystems inoperable is 
removed, and a requirement to shutdown the unit within 24 hours is 
provided in its place. Therefore, elimination of the allowance for 7 
days of operation with both SSW Subsystems inoperable does not 
involve a significant reduction in a margin of safety.
    For change No. 2:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change relocates testing information details for 
the Residual Heat Removal Service Water (RHRSW) and Station Service 
Water (SSW) systems, respectively, to the Technical Requirements 
Manual (TRM) under the control of 10 CFR 50.59. These controls are 
adequate to ensure the required testing is performed to verify 
operability. As such, these relocated details are not required to be 
in the Technical Specifications to provide adequate protection of 
the public health and safety. Changes to these relocated 
requirements in the TRM will be controlled by 10 CFR 50.59. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change will not impose or eliminate any requirements and adequate 
control of the information will be maintained. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change will not reduce a margin of safety because 
the simple relocation of testing details from the TS to the TRM has 
no impact on any safety analyses assumptions. Since any future 
changes to these requirements will be evaluated per the requirements 
of 10 CFR 50.59, no reduction in a margin of safety will be allowed. 
Therefore, this change does not involve a significant reduction in 
the margin of safety.
    For change No. 3:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change proposes to revise the wording of Station Service 
Water (SSW) Specifications to replace ``subsystem'' with

[[Page 6714]]

``essential equipment cooling loop'' to more accurately reflect 
VYNPS design and operation. At least two SSW pumps and one essential 
equipment cooling loop of the SSW System are required to be operable 
to mitigate the consequences of a design basis accident. Since this 
proposed change represents no change to existing requirements, this 
change will not significantly increase the consequences of any 
previously analyzed accident. In addition, SSW is not considered to 
be the initiator of any previously analyzed accident. Therefore, 
this change will not significantly increase the probability or 
consequences of any previously analyzed accident.
    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 a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in methods governing normal plant operation. The proposed 
change will not impose or eliminate any requirements and adequate 
control of existing requirements will be maintained. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change continues to provide the previous margin of 
safety regarding the capability to remove post-accident heat loads. 
At least two SSW pumps and one essential equipment cooling loop will 
be required to be operable or the unit will be required to be 
shutdown within 24 hours. Since this is the same basis both before 
and after the change, 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 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Project Director: William M. Dean.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: November 2, 1998.
    Description of amendment request: The licensee proposed to modify 
the Technical Specifications to more clearly describe the Emergency 
Core Cooling System Actuation Instrumentation--Low Pressure Coolant 
Injection (LPCI) System A/B Residual Heat Removal (RHR) Pump Start time 
delay requirements and the Core Spray System A/B Pump Start time delay 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    Change #1: Deletion of the O second time delay for first RHR 
pump (A/D) start.
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated:
    The proposed change does not involve a change to the plant 
design or operation. The instantaneous relays installed under 
corrective actions of LER 96-027 were evaluated as being equivalent 
in meeting the plant design of a 0 second time delay (instantaneous 
start) and an improvement on the minimum 500 millisecond time delay 
relays previously installed. The intent is to get LPCI flow started 
as soon as possible within the limits of the emergency bus power 
supply. The instantaneous start provides for a faster flow 
initiation. The proposed change does not affect any of the 
parameters or conditions that contribute to initiation of any 
accidents previously evaluated. Therefore, the proposed change 
cannot increase the probability of an accident previously evaluated.
    The proposed change does not involve a change in the operation 
of the relay controlling the initial RHR pump start on a [loss of 
coolant accident] LOCA with normal AC power not available. The 
instantaneous logic sequence relay functions to start the initial 
RHR Pump within 35 milliseconds of re-energization of the associated 
Emergency Bus. This start time is consistent with the plant safety 
analysis and [emergency diesel generator] EDG load analysis, 
therefore, the proposed change does not significantly increase the 
consequences of any accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated:
    This proposed change will not involve any physical changes to 
plant structures, systems or components (SSC), or the manner in 
which these SSCs are operated or maintained. Deletion of the 0 
second Time Delay Trip Function and associated calibration 
requirement will not affect initial RHR pump starting on a LOCA 
signal with normal AC power not available. The instantaneous logic 
sequence relay will still be tested under the Trip System Logic 
Functional Test at a frequency of once per operating cycle. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This proposed change to delete the 0 second Time Delay Trip 
Function and associated calibration requirement will not change 
operation of the initial RHR Pump start on a LOCA signal with normal 
power not available. The instantaneous logic sequence relay will 
function to initiate RHR Pump A/D start within 35 milliseconds of 
re-energization of the associated Emergency Bus, therefore, water 
will be delivered as designed. This RHR Pump start time is within 
the assumptions of the LOCA safety analysis of record. Therefore, 
this change does not involve a significant reduction in a margin of 
safety.
    Change #2: Addition of a 3 second lower limit to the trip level 
setting for the second RHR pump (B/C) start time delay trip 
function.
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated:
    The proposed change does not involve a change to the plant 
design or operation. The proposed change is more restrictive than 
existing Technical Specifications for this function. The proposed 
change limits the low value Trip Level Setting of the time delay 
relay and thus provides for EDG recovery from the initial RHR Pump 
(A/D) start. As a result, the proposed change does not affect any of 
the parameters or conditions that contribute to initiation of any 
accidents previously evaluated. The equipment will still start 
within the assumptions of the LOCA safety analysis of record. Thus, 
the proposed change cannot increase the probability of an accident 
previously evaluated.
    The proposed change ensures that the EDG has sufficient time to 
recover from the loading of the first RHR pump (A/D) prior to the 
loading of the second RHR pump (B/C). This load sequencing is 
experienced during a LOCA with normal AC power not available, thus 
providing increased reliability. Therefore, the proposed change will 
not result in a significant change in the consequences of any 
accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated:
    This proposed change will not involve any physical changes to 
plant systems, structures or components (SSC), or the manner in 
which these SSCs are intended to be operated or maintained. Addition 
of the 3 second lower limit on the second RHR Pump (B/C) Start Time 
Delay Function will ensure that, on a LOCA signal with normal AC 
power not available, the EDG voltage and frequency will adequately 
recover prior to the second RHR pump start. The instantaneous logic 
sequence relay will still be tested under the Trip System Logic 
Functional Test each Operating Cycle. Therefore, this change will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This proposed change to include a 3 second lower limit to the 
second RHR Pump Start Time Delay Trip Function will not change 
operation of the second RHR Pump

[[Page 6715]]

start on a LOCA signal (without normal power available). The 
proposed change will ensure sufficient time is available for the EDG 
to recover from the initial RHR Pump (A/D) start. The proposed 
second RHR Pump Start Time Delay Trip Level Setting of 3 [less than 
or equal to] t [less than or equal to] 5 seconds is within the 
assumptions of the LOCA evaluation and analysis of FSAR Sections 6.5 
and 8.5. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    Change #3: Addition of an 8 second lower limit to the trip level 
setting for the core spray pump (A/B) start time delay trip 
function.
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated:
    The proposed change does not involve a change to the plant 
design or operation. The proposed change is more restrictive than 
existing Technical Specifications for this function. The proposed 
change limits the low value Trip Level Setting of the time delay 
relay and thus provides for EDG recovery following the RHR B/C Pump 
start. As a result, the proposed change does not affect any of the 
parameters or conditions that contribute to initiation of any 
accidents previously evaluated. The equipment will still start 
within the assumptions of the LOCA analysis of record. Thus, the 
proposed change cannot increase the probability of an accident 
previously evaluated.
    The proposed change ensures that the EDG has sufficient time to 
recover following the loading of the B/C RHR pump and prior to the 
loading of the associated Core Spray pump. This load sequencing is 
experienced during a LOCA without normal power available, thus 
providing increased reliability. Therefore, the proposed change will 
not result in a significant change in the consequences of any 
accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated:
    This proposed change will not involve any physical changes to 
plant structures, systems or components (SSC), or the manner in 
which these SSCs are intended to be operated or maintained. Addition 
of the 8 second lower limit on the Core Spray Pump Start Time Delay 
Trip Function will ensure that, on a LOCA signal (with normal power 
not available) the EDG voltage and frequency will adequately recover 
prior to the Core Spray pump start. The Core Spray instantaneous 
logic sequence relays (normal AC available) and the CS Pump Start 
Time Delay relays will still be tested under the Trip System Logic 
Functional Test each Operating Cycle. Therefore, this change will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    This proposed change to include an 8 second lower limit to the 
Core Spray Pump Start Time Delay Trip Function will not change 
operation of the Core Spray Pump start on a LOCA signal with normal 
AC power not available. The proposed change will ensure sufficient 
time is available for the EDG to recover from the previous RHR Pump 
start. The proposed Core Spray Pump Start Time Delay Trip Level 
Setting of 8 [less than or equal to] t [less than or equal to] 10 
seconds is within the assumptions of the LOCA evaluation and 
analysis of FSAR Sections 6.5 and 8.5. Therefore, this 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Project Director: William M. Dean.

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

    Date of amendment request: September 24, 1998.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to modify the testing 
requirements for the reactor trip bypass breakers. The current TS 
require the bypass breakers to be tested ``prior to being placed in 
service.'' The proposed changes will allow the bypass breakers to be 
tested immediately after placing the breaker in service, but prior to 
commencing Reactor Protection System testing or maintenance.
    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 and testing of the reactor trip bypass breakers 
does not increase the probability of an accident or malfunction of 
equipment important to safety previously evaluated in the safety 
analysis report.
    The testing sequence will continue to ensure that the reactor 
trip system will be operable to mitigate the consequences of any 
unsafe or improper reactor operation during steady state or 
transient power operations. During the short period of time the 
breaker is closed before the undervoltage trip device test, the 
operability of the breaker is established based on satisfactory 
breaker testing conducted during the previous surveillance interval. 
Although the breaker is placed in service before it is tested, the 
breaker is tested as soon as practicable to verify operability prior 
to performing testing of the reactor trip system or required 
maintenance. Therefore, the proposed test sequence does not 
significantly increase the probability of occurrence or the 
consequences of any previously analyzed accident.
    (b) The proposed Technical Specifications do not create the 
possibility of an accident or malfunction of a different type than 
any evaluated previously in the safety analysis report.
    The proposed test sequence change does not alter the actual test 
performed to establish operability of the reactor trip bypass 
breakers. The bypass breakers will be proven operable prior to 
reactor trip system testing or required maintenance. During the 
short period of time the breaker is closed before the undervoltage 
trip device test, the operability of the breaker is established 
based on satisfactory breaker testing conducted during the previous 
surveillance interval. Although the breaker is placed in service 
before it is tested, the breaker is tested as soon as practicable to 
verify operability prior to performing testing of the reactor trip 
system or required maintenance. Therefore, it is concluded that no 
new or different kind of accident or malfunction from any previously 
evaluated has been created.
    (c) The proposed Technical Specifications change does not result 
in a significant reduction in margin of safety.
    The proposed change in the reactor trip bypass breaker test 
sequence provides assurance that the reactor trip system remains 
operable during normal operations or during reactor trip system 
testing and required maintenance to mitigate the consequences of any 
unsafe or improper reactor operation. Therefore, the proposed change 
in the test sequence for the reactor trip bypass breaker does not 
significantly reduce the margin of safety.
    This analysis demonstrate that the proposed amendment to the 
Surry Units 1 and 2 Technical Specifications does not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident, does not create the possibility of a 
new or different kind of accident and 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg. Virginia 23185.
    Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams, 
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia 
23219.

[[Page 6716]]

    NRC Project Director: Herbert N. Berkow.

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

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

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

    Date of amendment request: December 7, 1998.
    Description of amendment request: The proposed amendments would 
correct the lube oil inventory requirement from a range of 575-600 
gallons to a range of 375-400 gallons.
    Date of publication of individual notice in Federal Register: 
December 2, 1998 (63 FR 66591).
    Expiration date of individual notice: January 4, 1999.
    Local Public Document Room location: York County Library, 138 East 
Street, Rock Hill, South Carolina.

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

    Date of amendment request: January 4, 1999.
    Description of amendment request: The proposed amendment would 
change Technical Specifications (TSs) 3.5.2, ``Emergency Core Cooling 
Systems--ECCS Subsystems-Tavg [greater than or equal to] 300 [degrees 
Fahrenheit];'' 3.6.2.1, ``Containment Systems--Depressurization and 
Cooling Systems--Containment Spray and Cooling Systems;'' 3.7.1.2, 
``Plant Systems--Auxiliary Feedwater Pumps;'' 3.7.3.1, ``Plant 
Systems--Reactor Building Closed Cooling Water System;'' and 3.7.4.1, 
``Plant Systems--Service Water System.'' Changes to the acceptance 
criteria contained in these TSs are necessary based on revised 
hydraulic analyses and related accident analyses. Also, the bases of 
the associated TSs will be modified to address the proposed changes.
    Date of publication of individual notice in Federal Register: 
January 14, 1999 (64 FR 2523).
    Expiration date of individual notice: February 16, 1999.
    Local Public Document Room: Learning Resources Center, Three Rivers 
Community-Technical College, 574 New London Turnpike, Norwich, 
Connecticut, or the Waterford Public Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, Connecticut.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendments: December 7, 1998.
    Brief description of amendments: The amendments revise Technical 
Specification Section 3.8.3 to correct the lube oil inventory 
requirement from a range of 575-600 gallons to a range of 375-400 
gallons.
    Date of issuance: January 15, 1999.
    Effective date: As of the date of issuance to be implemented 
concurrently with implementation of Amendment Nos. 173 (Unit 1) and 165 
(Unit 2).
    Amendment Nos.: 175--Unit 1; 167--Unit 2.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69328).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of amendment request: June 29, 1998.
    Brief description of amendment: The amendment revises the 
Applicability of Technical Specification (TS) 3.4.2, ``Reactor Coolant 
System--Safety Valves--Shutdown.'' An associated action is also revised 
and a footnote is removed. The amendment also revises TS 3.4.12, 
``Reactor Coolant System--Overpressure Protection,'' allowing safety 
injection tanks to remain unisolated if they are pressurized to less 
than 300 psig and making some editorial changes. In addition, affected 
index and Bases pages are revised.
    Date of issuance: January 19, 1999.
    Effective date: The license amendment is effective as of its date 
of issuance with full implementation within 60 days.
    Amendment No.: 199.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998, (63 
FR 56243).
    The Commission's related evaluation of the amendments is contained 
in a

[[Page 6717]]

Safety Evaluation dated January 19, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of application for amendment: June 29, 1998.
    Brief description of amendment: The amendment approves a change to 
the Technical Specifications (TS) Table 3.3-4, ``Engineered Safety 
Feature Actuation System Instrumentation,'' to provide a range of 
acceptable values for the 4 KV buss loss of voltage relays rather than 
a single value as currently recorded in the TS. In addition minor 
changes were made to the trip time delay.
    Date of issuance: January 26, 1999.
    Effective date: The license amendment is effective as of its date 
of issuance and shall be implemented prior to the facility's restart 
from refueling outage 2R13.
    Amendment No.: 200.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56244).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 
50-458, River Bend Station, Unit 1, West Feliciana Parish, 
Louisiana

    Date of amendment request: September 22, 1998.
    Brief description of amendment: The amendment deleted license 
conditions associated with the River Bend Station (RBS) Transamerica 
Delaval, Inc. (TDI) emergency diesel generators (EDGs), which 
prescribed various inspection requirements following an EDG overload 
condition. The License Conditions were originally issued following the 
publication of NUREG 1216, which called for extensive periodic engine 
tear-downs as the major part of a maintenance and surveillance program 
for TDI engines. The removal of the aforementioned license conditions 
is consistent with the NRC's approval of Generic Topical Report TDI-
EDG-001-A ``Basis for Modification to Inspection Requirements for 
Transamerica Delaval, Inc., Emergency Diesel Generators''. EOI will 
continue to inspect and maintains its EDGs in accordance with Technical 
Requirements Manual (TRM) surveillance requirement TSR 3.8.1.21. 
Periodicity of planned inspections and maintenance are based upon the 
manufacturer's recommendations for standby service.
    Date of issuance: January 27, 1999.
    Effective date: January 27, 1999.
    Amendment No.: 102.
    Facility Operating License No. NPF-47: The amendment revised the 
operating license.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59592).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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

    Date of application for amendment: August 24, 1998, as supplemented 
November 20, 1998.
    Brief description of amendment: The amendment approves operator 
action for meeting the ``ready-to-load'' requirement for the Division 3 
diesel generator.
    Date of issuance: January 19, 1999.
    Effective date: January 19, 1999.
    Amendment No.: 119.
    Facility Operating License No. NPF-62: The amendment authorized 
revision of the Updated Safety Analysis Report.
    Date of initial notice in Federal Register. September 10, 1998 (63 
FR 48529).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 19, 1999.
    No significant hazards consideration comments received: No.
     Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.

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

    Date of application for amendment: July 31, 1998.
    Brief description of amendment: The amendment clarifies 
requirements for diesel generator start voltage and frequency.
    Date of issuance: January 20, 1999.
    Effective date: January 20, 1999.
    Amendment No.: 120.
    Facility Operating License No. NPF-62: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53949).

    The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 20, 1999.
    No significant hazards consideration comments received: No.

    Local Public Document Room location: The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.

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

    Date of amendment request: March 2, 1998.
    Description of amendment request: The amendment changes the 
Technical Specifications by eliminating the emergency diesel generator 
accelerated testing and special reporting requirements of TS 
4.8.1.1.2a, 4.8.1.1.3, Table 4.8-1 and 4.8.1.2 in accordance with 
Generic Letter 94-01.
    Date of issuance: January 21, 1999.
    Effective date: As of its date of issuance, to be implemented 
within 60 days.
    Amendment No.: 59.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19971). The Commission received comments which were addressed in the 
staff's Safety Evaluation dated January 21, 1999.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 21, 1999.
    No significant hazards consideration comments received: Yes.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

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

    Date of application for amendment: September 28, 1998.
    Brief description of amendment: The amendment approves the 
previously implemented revision to the Final Safety Analysis Report 
(FSAR) Section 8.7.3.1 that changed certain electrical separation 
requirements from 12 inches to 6 inches. The FSAR change was

[[Page 6718]]

previously implemented following an erroneous 10 CFR 50.59 evaluation.
    Date of issuance: January 20, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 224.
    Facility Operating License No. DPR-65: Amendment revised the FSAR.
    Date of initial notice in Federal Register: November 4, 1998 (63 FR 
59593).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 20, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: November 10, 1998.
    Brief description of amendment: The amendment changes Technical 
Specifications 3.3.1.1, ``Reactor Protective Instrumentation,'' and 
3.3.2.1, ``Engineered Safety Feature Actuation System 
Instrumentation,'' to restrict the time a reactor protection or 
engineered safety feature actuation channel can be in the bypass 
position for 48 hours, from an indefinite period of time.
    Date of issuance: January 27, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 225.
    Facility Operating License No. DPR-65: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69343).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

PECO Energy Company, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: September 14, 1998.
    Brief description of amendment: This amendment revised Limerick 
Generating Station, Unit 2, Technical Specification (TS) Table 
4.4.6.1.3-1, ``Reactor Vessel Material Surveillance Program--Withdrawal 
Schedule.'' The revision changed the schedule for withdrawing the first 
surveillance capsule from 8 Effective Full Power Years (EFPY) to 15 
EFPY, and the second surveillance capsule from 20 EFPY to 30 EFPY. A 
revision to the TS Surveillance Requirement (SR) has also been made. 
This revision removed the reference to flux wire removal and analysis 
that was originally required following the first cycle of operation. TS 
SR 4.4.6.1.4 was changed to refer to the flux wires that are located 
within the surveillance capsules, which will be removed and analyzed in 
accordance with the surveillance capsule removal schedule located in 
Table 4.4.6.1.3-1.
    Date of issuance: January 12, 1999.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment No.: 94.
    Facility Operating License No. NPF-85: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 21, 1998 (63 FR 
56253).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: October 15, 1998.
    Brief description of amendments: These amendments add a new 
Technical Specification (TS) section and TS Bases section to 
incorporate a special test exception to allow reactor coolant 
temperatures greater than 200  deg.F but less than or equal to 212 
deg.F during inservice testing and hydrostatic testing.
    Date of issuance: January 12, 1999.
    Effective date: Both units, as of the date of issuance, to be 
implemented within 30 days.
    Amendment Nos.: 133 and 95.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64120).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 12, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

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

    Date of application for amendments: May 10, 1996, as supplemented 
on March 19 and August 29, 1997.
    Brief description of amendments: This amendment incorporates into 
the Technical Specifications the Margin Recovery portion of the Fuel 
Upgrade Margin Recovery Program and supports increased steam generator 
plugging, improved fuel reliability, reduced fuel costs, longer fuel 
cycles, reduced spent fuel storage, and enhanced reactor safety. In a 
letter dated November 26, 1997, the Commission issued the amendment for 
Salem Unit 1.
    Date of issuance: January 8, 1999.
    Effective Date: January 8, 1999.
    Amendment Nos.: 197.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications and/or License.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34898).
    The March 19, and August 29, 1997, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
January 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

South Carolina Electric & Gas Company, South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of application for amendment: July 1, 1998.
    Brief description of amendment: The amendment revises Virgil C. 
Summer Nuclear Station Technical Specification Surveillance Requirement 
(SR) 4.7.7.e to

[[Page 6719]]

remove the ``during shutdown'' condition from the specified test 
interval. The amendment also makes administrative changes to SR 
4.7.7.g, and BASES 3/4.2.2 and 3/4.2.3 to correct typographical errors.
    Date of issuance: January 27, 1999.
    Effective date: January 27, 1999.
    Amendment No.: 141.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53955).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.

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

    Date of application for amendments: September 3, 1998, as 
supplemented by letter dated December 8, 1998.
    Brief Description of amendments: The amendments change the Vogtle 
Electric Generating Plant, Units 1 and 2 Technical Specifications to: 
(1) Support the replacement of the Nuclear Instrumentation System 
Source Range and Intermediate Range Channels and Post-Accident Neutron 
Flux Monitoring System, and (2) delete the requirement for performing 
response time testing of the source range channels and power range 
detector plateau voltage determinations.
    Date of issuance: January 22, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1--104; Unit 2--82.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: October 7, 1998 (63 FR 
53957).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 22, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

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: October 29, 1998.
    Brief description of amendments: Relocates the Technical 
Specification 3/4.3.4 requirements for Turbine Overspeed Protection to 
the Technical Requirements Manual.
    Date of issuance: January 21, 1999.
    Effective date: The license amendment is effective as of its date 
of issuance, to be implemented within 30 days of issuance.
    Amendment Nos.: Unit 1--Amendment No. 101; Unit 2--Amendment No. 
88.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998, (63 
FR 69347). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 21, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit No. 2, and Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit No. 3, Limestone County, Alabama

    Date of amendment request: March 3, 1998 as supplemented November 
13, and December 15, 1998.
    Description of amendment request: The amendments revise the 
pressure-temperature limit curves in the Technical Specifications (TS) 
for BFN Units 2 and 3 to 16 and 20 effective full power years, 
respectively.
    Date of issuance: January 15, 1999.
    Effective date: January 15, 1999.
    Amendment Nos.: 257 and 217.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the TS.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19979). The licensee's letters of November 13, and December 15, 1998, 
did not expand the scope of the application or affect the staff's 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated January 15, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611.

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

    Date of application for amendment: June 26, 1998, as supplemented 
November 6, 1998. (TS 98-06).
    Brief description of amendment: The amendment authorizes the 
deletion of the power range neutron flux high negative rate reactor 
trip function based on the analysis provided in Westinghouse Electric 
Corporation WCAP-11394-A, ``Methodology for the Analysis of the Dropped 
Rod Event.''
    Date of issuance: January 15, 1999.
    Effective date: January 15, 1999.
    Amendment No.: 18.
    Facility Operating License No. NPF-90: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: July 29, 1998 (63 FR 
40562). The November 5, 1998, letter contained clarifying information 
that did not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 15, 1998.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: November 11, 1998.
    Brief description of amendments: Revises core safety limit curve 
and Overtemperature N-16 reactor trip setpoints based on analysis of 
the core configuration and expected operation for the CPSES Unit 2, 
Cycle 5. The changes apply equally to CPSES Units 1 and 2 licenses 
since the Technical Specifications are combined.
    Date of issuance: January 29, 1999.
    Effective date: The license amendment is effective as of its date 
of issuance, to be implemented within 90 days of issuance.
    Amendment Nos.: Unit 1--Amendment No. 63; Unit 2--Amendment No. 49.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71974).

[[Page 6720]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: February 24, 1998, as 
supplemented by letters dated May 27, June 25, August 25, September 3, 
November 3, and December 4, 1998.
    Brief description of amendment: The amendment revised the technical 
specifications to allow an increase in the Callaway Plant, Unit 1 spent 
fuel pool storage capacity from 1344 fuel assemblies to 2363 fuel 
assemblies. The amendment also revises the technical specifications to 
allow storage of an additional 279 fuel assemblies in the cask loading 
pit.
    Date of issuance: January 19, 1999.
    Effective date: January 19, 1999, to be fully implemented no later 
than December 31, 1999, except that the racks in the cask loading pit 
may be installed at a future time after the completion of the next 
refueling outage.
    Amendment No.: 129.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 13, 1998 (63 FR 
37598).
    The June 25, August 25, September 3, November 3, and December 4, 
1998, supplemental letters provided additional clarifying information 
that did not change the staff's original no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 19, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Elmer Ellis Library, 
University of Missouri, Columbia Missouri 65201.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: November 3, 1998.
    Brief description of amendment: The amendment makes administrative 
changes to the Technical Specifications to correct errors, add 
consistency within the Technical Specifications, and make nomenclature 
changes to support and enhance usability of the Technical 
Specifications.
    Date of Issuance: January 5, 1999.
    Effective date: January 5, 1999, to be implemented within 30 days.
    Amendment No.: 164.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66605).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 5, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of application for amendment: December 11, 1998.
    Brief description of amendment: The amendment revises the Technical 
Specifications to allow manual containment isolation valves to be 
opened intermittently under administrative controls.
    Date of Issuance: January 19, 1999.
    Effective date: January 19, 1999, to be implemented within 30 days.
    Amendment No.: 165.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1998 (63 
FR 70168).
    The Commission's related evaluation of this amendments is contained 
in a Safety Evaluation dated January 19, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: December 17, 1998, as 
supplemented by letter dated January 21, 1999.
    Brief description of amendment: The amendment revised Technical 
Specification Surveillance Requirement 3.8.1.8 to remove the 
restriction on testing of the manual transfer between the startup and 
backup offsite power sources while in Mode 1 or 2.
    Date of issuance: January 27, 1999.
    Effective date: January 27, 1999, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 156.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1998 (63 
FR 70807).
    The January 21, 1999, supplemental letter provided additional 
clarifying information and did not change the staff's original no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

    Dated at Rockville, Maryland, this 3rd day of February 1999.

    For the Nuclear Regulatory Commission.
John N. Hannon,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-3098 Filed 2-9-99; 8:45 am]
BILLING CODE 7590-01-M