[Federal Register Volume 63, Number 106 (Wednesday, June 3, 1998)]
[Notices]
[Pages 30261-30271]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-14519]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from May 11, 1998, through May 21, 1998. The last 
biweekly notice was published on May 20, 1998 (63 FR 27757).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not; 
(1) Involve a significant increase in the probability or consequences 
of an accident previously evaluated; or (2) create the possibility of a 
new or different kind of accident from any accident previously 
evaluated; or (3) involve a significant reduction in a margin of 
safety. The basis for this proposed determination for each amendment 
request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public

[[Page 30262]]

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

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

    Date of amendment request: March 6, 1998.
    Description of amendment request: The proposed change will revise 
the H. B. Robinson, Unit 2 Technical Specifications to allow use of the 
Post Accident Monitoring (PAM) source range (SR) neutron flux detector 
as a compensatory measure in the event that one of the two required BF3 
detectors become inoperable while the plant is in MODE 6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change to Technical Specifications is only applicable 
during the refueling mode of operation (MODE 6). Neither the BF3 SR nor 
PAM neutron flux monitors provide an automatic initiation signal for 
the operation of plant systems or components but are only relied upon 
to provide indication of core reactivity. Since the proposed change to 
Technical Specifications does not alter the design or operation of 
plant equipment or systems, there is no change in the initiating 
mechanisms for

[[Page 30263]]

any accidents previously analyzed. Therefore this change does not 
involve a significant increase in the probability for an accident 
previously analyzed.
    The UFSAR [Updated Final Safety Analysis Report] identifies two 
accidents that credit the SR monitoring capability in MODE 6, the boron 
dilution accident and the fuel handling accident. No other accidents 
were found to rely on SR monitoring in MODE 6. The proposed change will 
continue to require BF3 SR visual indication of core reactivity in the 
control room and a BF3 SR neutron flux monitor audible indication in 
containment. This change will not result in a significant reduction in 
operator capability to detect unexpected changes in core reactivity and 
perform actions credited with termination of those events, therefore 
the proposed change does not involve a significant increase in the 
consequences of an accident previously analyzed.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change to Technical Specifications does not involve 
any physical alteration of plant systems, structures or components or 
changes in parameters governing plant operations. The proposed change 
will not result in a significant reduction in monitoring capability 
since two BF3 SR channels of SR visual indication in the control room 
and audible SR indication in the containment are required during core 
alterations and positive reactivity changes. The use of the PAM SR 
neutron flux monitor as a compensatory measure does not introduce any 
new accident initiation scenarios since the SR instruments are for 
monitoring and criticality assessment only and are not relied upon to 
initiate automatic accident mitigation measures. Therefore, this change 
does not create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The proposed change will maintain two BF3 SR monitoring means for 
visually monitoring core reactivity as currently discussed in the bases 
for the affected Technical Specifications. Audible indication provided 
by one BF3 SR neutron flux monitor will still be required and fulfilled 
by the remaining BF3 SR neutron flux monitor. The PAM SR neutron flux 
monitors use fission chambers as detectors which have a sensitivity of 
4 cps/neutron-volts (cps/nv) for thermal neutrons and 2 cps/nv for fast 
neutrons. The BF3 SR neutron flux monitors have a sensitivity of 9 cps/
nv. The PAM SR neutron flux monitor has comparable range and accuracy 
(i.e., range of 1E-01 cps to 1E+05 cps with an accuracy of 2% of full 
scale) to that of BF3 SR neutron flux monitor (i. e., range of 1E-00 
cps to 1E + 06 cps with an accuracy of 3% of full scale) which meets 
the Technical Specifications Section 3.9.2 Bases requirements of 6 
decades of indication and 5% accuracy. 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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: P. T. Kuo, Acting.

Commonwealth Edison Company, Docket No. 50-249, Dresden Nuclear Power 
Station, Unit 3, Grundy County, Illinois

    Date of amendment request: May 6, 1998.
    Description of amendment request: The proposed amendment would 
amend Technical Specification (TS) 4.6.E to allow a one-time extension 
of the 40-month requirement to pressure set test or replace all Main 
Steam Safety Valves (MSSVs) to a maximum interval of 60 months as 
currently allowed by the American Society of Mechanical Engineers 
(ASME) Boiler and Pressure Vessel Code (Code).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    The proposed changes request a one-time change to the surveillance 
requirement for the MSSVs. The surveillance interval between safety 
valve testing is not a precursor assumed in any previously analyzed 
accident. Therefore, the probability of a previously evaluated accident 
has not been increased.
    The proposed extension is consistent with the ASME Code requirement 
to test all valves within 60 months. The proposed changes are also 
consistent with NUREG-1433 and do not adversely affect existing plant 
safety margins or the reliability of the equipment assumed to operate 
in the safety analysis. Operating experience and superior materiel 
condition of the MSSVs support the expectation that they will continue 
to perform their intended function. Therefore, the consequences of a 
previously evaluated accident have not been increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    No new equipment is required, nor will the MSSVs be operated in a 
different manner during the period of the extended surveillance 
interval. The proposed change is consistent with NUREG-1433 
requirements for safety valve surveillance intervals as well as the 
ASME Code for requirements testing safety valves. Operating experience 
and superior materiel condition of the MSSVs support the expectation 
that they will continue to perform their intended function. Therefore, 
the possibility of a new or different accident has not been increased.
    3. Involve a significant reduction in the margin of safety because:
    The proposed amendment represents an extension to the current TS 
requirements, but would otherwise be provided generically by ASME Code. 
The proposed changes are also consistent with NUREG-1433, request a 
shorter total interval than previously granted by the Staff (Reference 
b), [J.F. Stang (NRC) to D. L. Farrar, SER dated October 8, 1996] and 
do not adversely affect existing plant safety margins or the 
reliability of the equipment assumed to operate in the safety analysis. 
The proposed changes have been evaluated and found to be acceptable for 
use at Dresden based on system safety analysis requirements and 
operational performance. The MSSV provisions continue to be adequately 
maintained during plant operation. The proposed changes to the MSSV 
surveillance interval do not significantly reduce existing plant safety 
margins since excellent materiel condition and acceptable surveillance 
test results support the expectation that no significant degradation 
will occur over the extended interval.
    The proposed changes are based on NRC accepted provisions at other 
operating plants that are applicable at

[[Page 30264]]

Dresden and maintain necessary levels of system or component 
reliability.
    The proposed amendment for Dresden will not reduce the availability 
of systems required to mitigate accident conditions; therefore, the 
proposed changes do 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: May 4, 1998.
    Description of amendment request: The proposed amendment would 
incorporate Technical Specifications Requirements for the protection 
systems for the new static VAR compensators being installed onsite to 
address degraded electrical grid voltage.
    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 changes addressed by this amendment request involve the 
addition of SVCs and their associated protection systems to the onsite 
circuit connections for the plant offsite electrical power sources, 
i.e., to the RAT and ERAT. As noted throughout this request, the 
addition of the SVCs will help to maintain voltage at the site for both 
of the offsite electrical power sources consistent with the ``capacity 
and capability'' requirements of GDC 17. Further, the regulating effect 
of the SVCs will compensate for the voltage drop that can occur without 
the SVCs when the plant trips off-line (and thus no longer supports 
grid voltage) during normal or accident conditions. This supports 
compliance with the GDC 17 requirement to minimize the probability of 
losing electric power from the offsite supplies as a result of, or 
coincident with, the loss of power from the offsite supplies as a 
result of, or coincident with, the loss of power generated by the 
nuclear power unit. Consequently, the likelihood of transferring to the 
onsite emergency power supplies (diesel generators) during an accident 
will be reduced. At the same time, as also addressed in this amendment 
request, incorporation of the SVCs into the CPS auxiliary power system 
requires consideration of failure modes that could be introduced by the 
SVCs wherein such failure modes could involve a significant increase in 
the probability or consequences of any accident previously evaluated.
    By supplying each of the SVCs with an enhanced protection system, 
consisting of dual, redundant protection subsystems, either of which 
will isolate the SVC from the bus (by automatically opening the SVC 
main circuit breakers) in response to postulated SVC failures or 
associated abnormal conditions, the potential for such conditions or 
failures to adversely affect the plant safety busses, the associated 
plant loads, or the onsite emergency electrical power sources is 
reduced to a very low probability. The protection systems designed for 
the SVCs include consideration of failure modes or abnormal conditions 
that may be postulated or expected to occur with some degree of 
probability for the offsite electrical sources or grid with or without 
the presence of the SVCs, (such as a sustained degraded voltage 
condition), as well as consideration of any new or other failure modes 
or abnormal conditions potentially introduced by the SVCs that would be 
less likely to occur in the offsite electrical network without the 
presence of the SVCs (such as the introduction of harmonics). The 
proposed change to the CPS Technical Specifications to incorporate 
requirements for the SVC protection systems will ensure that the SVC 
protection systems are adequately maintained in an operable condition 
to perform their intended function of protecting against such 
conditions or failure modes. Operable SVC protection systems will 
reduce the probability of an SVC failure event that leads to equipment 
damage and subsequent core damage to a level that makes such an event 
incredible.
    It should be noted that tripping of the SVCs in response to an SVC 
failure or abnormal condition does not result in a loss of power from 
the offsite sources. Thus, the probability of a loss of offsite power, 
which is an analyzed event in the plant safety analyses, will not be 
significantly increased by the SVC protection systems.
    As noted previously, the proposed change to the Technical 
Specifications to incorporate SVC protection system operability and 
testing requirements would ensure that plant safety systems or 
components are not electrically affected by the SVCs in an adverse 
manner. In addition, except where the SVCs are physically located and 
connected to the ERAT and RAT via bus ducts, plant safety-related 
structures and supporting systems would not be mechanically affected by 
the SVCs. Separation, clearance and related requirements to ensure no 
other interaction with the RAT, ERAT and offsite source connections, as 
well as for maintaining offsite source independence would be 
maintained. On this basis, the safety functions of systems for 
preventing or mitigating analyzed events or accidents would not be 
impacted by the SVCs.
    Based on the above, the proposed change to the Technical 
Specifications does not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    (2) In consideration of the potential adverse impacts that the SVCs 
may have on plant systems, structures or components, such impacts are 
primarily confined to potential electrical faults or abnormal 
conditions. As noted above, the SVCs have no mechanical impact on 
safety-related plant systems, structures or components. Thus, no new 
failure modes or precursors to potentially new and unanalyzed events 
would be introduced via any mechanical means.
    With respect to potential adverse electrical impacts, the potential 
electrical failure modes or abnormal conditions postulated for the SVCs 
include conditions or events that, although could be considered 
possible for the offsite sources (i.e., the grid), were not in fact 
considered credible and therefore previously evaluated for the offsite 
electrical sources. These conditions or events, such as the 
introduction of harmonics or excessive overvoltage or phase imbalance 
caused by an SVC failure, would have the potential to degrade plant 
safety-related equipment connected to the busses at the time of the SVC 
failure if no protection for such conditions was provided. However, 
enhanced protection systems are provided for the SVCs to ensure that 
such failures cannot damage plant equipment. As noted previously, the 
probability of an event involving an SVC failure that leads to 
equipment damage and subsequent core damage has been calculated to be 
1.5 x 10<SUP>-8</SUP>/year. This low probability makes such an event 
incredible just as comparable events that could be postulated for the 
offsite electrical power sources were not previously

[[Page 30265]]

considered credible and therefore were not considered to be design 
basis events. The calculated probability of 1.5 x 10<SUP>-8</SUP>/year 
for an SVC failure event involving core damage is an order of magnitude 
lower than the threshold probability criterion specified in Section 
2.2.3 of the Standard Review Plan (NUREG 0800) for design basis events 
involving an offsite hazard that can lead to core damage and 
radioactive release with dose consequences in excess of the limits 
specified in 10 CFR Part 100.
    The proposed change to the Technical Specifications incorporates 
requirements for maintaining operability of the SVC protection systems. 
On this basis and as described above, no new credible accidents that 
could be associated with the SVCs (i.e., failure of the SVCs) are thus 
introduced, so that the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) As noted previously, incorporation of the SVCs into the CPS 
auxiliary power system will support or regulate plant bus voltage for 
both of the offsite sources. Specifically, analysis has shown that the 
SVCs will recover reduced margin that has occurred or would occur in 
the future (without the SVCs) with respect to the voltage required for 
plant safety loads and the minimum expected offsite voltage, under 
normal and accident conditions (i.e., under steady-state and transient 
voltage conditions). This also means that the SVCs will enhance the 
capability and capacity of the offsite sources such that, when compared 
to the configuration of not having the SVCs, either source will be more 
likely to reset the safety bus degraded voltage relays in the event of 
an accident, thus permitting the preferred offsite sources to remain 
connected (and not causing a transfer to the diesel generators). These 
desirable results constitute a significant increase in the margin of 
safety with respect to voltage requirements for plant loads.
    Based on the above, IP has concluded that the proposed change to 
the Technical Specifications to support use of the SVCs and their 
protection systems 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, IL 62525.
    NRC Acting Project Director: Ronald R. Bellamy.

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

    Date of amendment request: April 22, 1998.
    Description of amendment request: The proposed change would revise 
selected Technical Specification (TS) surveillance requirements to 
accommodate fuel cycles of up to 24 months for surveillances that are 
currently performed at each 18-month or other specified outage 
interval. Specifically, the following TS surveillance requirements 
would be revised by the proposed change: 4.1.2.2.b and c, ``Boration 
Systems Flow Paths--Operating;'' 4.3.3.5.2, ``Remote Shutdown System;'' 
4.4.3.2, ``Pressurizer;'' 4.4.4.1, ``Relief Valves;'' 4.4.6.2.2.a and 
b, ``Operational Leakage;'' 4.4.11.2, ``Reactor Coolant System Vents;'' 
4.5.1.1.d.1 and 2, ``Accumulators;'' 4.5.2.d, e, g.2), and h, 
``Emergency Core Cooling System (ECCS) Subsystems--Tavg Greater Than or 
Equal to 350 deg.F;'' 4.6.3.2, ``Containment Isolation Valves;'' and 
4.7.1.2.1.c, ``Auxiliary Feedwater System.'' In conjunction with the 
proposed change, components addressed in the following TS surveillance 
requirements have been evaluated to support an extension in frequency 
to accommodate fuel cycles of up to 24 months: 4.6.3.1 and 3, 
``Containment Isolation Valves;'' 4.7.1.2.2, ``Auxiliary Feedwater 
System;'' 4.7.1.5, ``Main Steam Line Isolation Valves;'' and 4.7.1.6, 
``Atmospheric Relief Valves.'' In addition, the proposed change would 
delete the restriction ``during shutdown'' in those TS surveillance 
requirements where this restriction is stated.
    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 have no adverse affect on 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 changes do 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 changes are administrative in nature and do not 
change the level of programmatic controls or the procedural details 
associated with aforementioned surveillance requirements.
    Changing the frequencies of the aforementioned surveillance 
requirements from at least once per 18 months to at least once per 
refueling interval does not change the basis for the frequencies. The 
frequencies were chosen because of the need to perform these 
verifications under the conditions that are normally found during a 
plant refueling outage, and to avoid the potential of an unplanned 
transient if these surveillances were conducted with the plant at 
power.
    Equipment performance over several operating cycles was evaluated 
to determine the impact of extending the surveillance intervals. This 
evaluation included a review of surveillance results, preventative 
maintenance records, and the frequency and type of corrective 
maintenance activities, and a failure mode analysis. The evaluations 
conclude that the subject SSCs are highly reliable, presently 
exhibiting no time dependent failure modes of significance, and that 
there is no indication that the proposed extension could cause 
deterioration in the condition or performance of the subject SSCs. 
There are no known mechanisms that would significantly degrade the 
performance of the evaluated equipment during normal plant operation. 
Although there have been generic or repetitive failures of some 
components in the past, which may have affected the ability of the SSCs 
to consistently and successfully perform their safety function, those 
items have been resolved through design changes and rework such that 
they have not recurred. There have been no repetitive failures or time 
dependent failures that were significant in nature which would have 
prevented the SSCs from performing their intended safety function.
    Deletion of the restriction ``during shutdown'' where this 
restriction is stated will permit performance of certain maintenance 
and testing activities during conditions or modes other than shutdown. 
North Atlantic

[[Page 30266]]

will ensure, through the implementation of administrative controls that 
proper regard to their effect on safe operation of the plant is given 
prior to conduct of a particular surveillance in a condition or mode 
other than shutdown.
    Since the proposed changes only affect the surveillance intervals 
for SSCs that are used to mitigate accidents, the changes do not affect 
the probability or consequence of a previously analyzed accident. While 
the proposed changes will lengthen the intervals between surveillances, 
the increase in intervals has been evaluated. Based on the reviews of 
the surveillance tests, inspections, and maintenance activities, it is 
concluded that there is no significant adverse impact on the 
reliability or availability of these SSCs.
    Since there are no changes to previous accident analyses, the 
radiological consequences associated with these analyses remain 
unchanged, therefore, 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 impact on component or 
system interactions. The proposed changes are administrative in nature 
and do not change the level of programmatic controls and procedural 
details associated with the aforementioned surveillance requirements. 
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 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 changes are administrative in nature and do 
not change the level of programmatic controls and procedural details 
associated with the aforementioned surveillance requirements.
    From the evaluations performed on the subject SSCs there are no 
indications that potential problems would be cycle-length dependent or 
that potential degradation would be significant for the time frame of 
interest and, therefore, increasing the surveillance interval to the 
bounding limit of 30 months (24 months plus 25%) will have little, if 
any, impact on safety.
    The proposed changes to the surveillance intervals are still 
consistent with the basis for the intervals and the intent and method 
of performing the surveillance is unchanged. Deletion of the 
restriction ``during shutdown'' where this restriction is stated will 
permit performance of certain maintenance and testing activities during 
conditions or modes other than shutdown. North Atlantic will ensure, 
through the implementation of appropriate administrative controls, that 
proper regard to their effect on safe operation of the plant is given 
prior to conduct of a particular surveillance in a condition or mode 
other than shutdown. In addition, use of the subject SSCs during normal 
plant operation, combined with their previous history of availability 
and reliability, provide assurance that the proposed changes will not 
affect the reliability of the subject SSCs. Thus, it is concluded that 
the subject SSCs would be available upon demand to mitigate the 
consequences of an accident and, therefore, there is no significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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: Cecil O. Thomas.

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: April 29, 1998.
    Description of amendment request: The proposed amendment would 
allow NNECO to revise the Updated Final Safety Analysis Report (UFSAR) 
for Millstone Unit 2 by deleting the diversity requirement for the two 
low-range pressurizer pressure transmitters, PT-103 and PT-103-1.
    NNECO proposes to replace PT-103 and PT-103-1 with transmitters 
that are more accurate in a post-accident environment to provide 
assurance that entry into shutdown cooling in a post-accident 
environment is not compromised and to provide relief for the reactor 
coolant system pressure/temperature curves. NNECO further indicates 
that only a single model series of Rosemount transmitters meet the 
revised design requirements and has specifically requested to delete 
the diversity requirement in the UFSAR, Section 4.3.8.2.3, 
``Pressurizer Pressure.'' NNECO has determined that this deviation from 
the current design basis constitutes an unreviewed safety question as 
defined in 10 CFR 50.59. Basis for proposed no significant hazards 
consideration determination: As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The replacement of the low-range pressurizer pressure transmitters 
with non-diverse transmitters will reduce the instrument uncertainties 
post--SBLOCA [small break loss-of-coolant accident] or MSLB [main 
steamline break]. The probability of a post-accident intersystem LOCA 
[loss-of-coolant accident] as a result of aligning the SDC [shutdown 
cooling] system to RCS [reactor coolant system] pressures beyond its 
design pressure is reduced due to the reduced uncertainties of low-
range pressurizer pressure signals to the SDC suction valve interlocks. 
The reduced uncertainty associated with the low-range pressurizer 
pressure transmitters in a harsh environment will not significantly 
reduce the probability of previously evaluated accidents relative to 
the use of the transmitter signal as an input variable to the ICC 
[inadequate core cooling] system, or to the other functions of LTOP 
[low temperature overpressure protection], SIT [safety injection tank] 
interlock, and Hot Shutdown Panel indication since these are functions 
not required in post-accident design bases. With respect to ICC, other 
parameters are available to the operator to determine adequate cooling 
of the core is taking place and

[[Page 30267]]

saturation conditions are being approached or are occurring.
    The loss of diversity in manufacturer and operating principle 
results in a small increase in the susceptibility of the replacement 
transmitters to common cause events that are primarily linked to 
internal failures of the transmitters verses failure that result from 
external events or conditions. To some extent, externally related 
common cause failures that can result from calibration or maintenance 
errors can be expected to also increase slightly because of commonality 
of procedures. Common cause failure increase is considered for the 
identified functions and all accidents. Because of the slight increase 
in the probability of common cause failures, the probability of 
exceeding RCS pressure/temperature (P/T) curves at temperatures [less 
than] 275 deg.F is slightly increased (assuming a common cause failure 
of the replacement transmitters that would result in indicating a 
pressure lower than actual RCS pressure). Also, the potential for 
exceeding ASME Section III, Appendix G, pressure/temperature limit 
curves on cooldown and heatup is also slightly increased due to the 
slight increase in potential for common cause failure. This small 
increase in the potential for common cause failure will not 
significantly affect the probability of previously evaluated accidents. 
The reasons for this are:

--Exceedance of P/T limit curves does not in and of itself result in an 
accident initiator.
--Internally caused common cause failures are not expected to have a 
significant impact on the overall common cause potential of the 
transmitters. Typically, the majority of common cause potential is due 
to external reasons. Further, many times simultaneous internal failures 
of instrumentation can be recognized by direct comparison at which time 
alternative means can be sought, if available. Unless failure of the 
replacement transmitters was simultaneous and resulted in consistent 
output signals, transmitter failure would likely be recognized before 
requiring LTOP.
--The narrow-range pressurizer pressure loops (1500-2500 psia [pounds 
per square inch absolute]) are fully qualified Class 1E with 
transmitters manufactured by a different vendor. They provide a check 
against the low-range pressurizer pressure transmitters in the 
overlapped range of 1500-1600 psia.
--The non-class 1E wide-range pressurizer pressure loop (0-3000 psig 
[pounds per square inch gauge]), although not environmentally 
qualified, utilizes a qualified type transmitter manufactured by a 
different vendor and has been demonstrated to be reliable. It provides 
a check against the low-range pressurizer pressure transmitters in the 
overlapped range of 0-1600 psia.

    The probability of an intersystem LOCA will not increase, due to 
the multiple means of determining that RCS pressure is beyond the SDC 
system design pressure and the multiple failures that would have to 
occur. All other functions evaluated (ICC, Hot Shutdown Panel and the 
SIT interlock) would not increase the probability of a previously 
evaluated accident.
    Because the function and output of the replacement transmitters is 
the same as the existing transmitters and the transmitter failure types 
has not changed, the radiological consequences of previously evaluated 
accidents are not affected by the proposed change. The exception to 
this occurs when considering consequences of accidents that result in a 
harsh environment inside Containment and requiring SDC for long term 
cooling. In these cases, (SBLOCAs and MSLBs inside containment) given 
that transmitter accuracy is improved, the ability of getting onto SDC 
improves. This allows getting onto SDC more consistently. By doing so, 
a reduction in the radiological consequences of these accidents may be 
improved.
    Therefore, there is 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 function and output of the replacement transmitters are the 
same as the existing transmitters, and the transmitter failure types 
have not changed.
    Therefore, the change 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.
    Improvement of the transmitters' ability to provide an accurate 
interpretation of RCS pressures in the operating range of 0-1600 psia 
(post-accident harsh environment in Containment) results in a positive 
benefit in the ability to control cooldown rates, establish SDC, and 
operate at the proper RCS pressures. However, the Margin of Safety is 
not impacted when the original transmitters/uncertainty calculations 
are compared to the proposed replacement transmitters/uncertainty 
calculations with regard to RCS P/T curves.
    Therefore, the change will not involve a reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: June 25, 1997.
    Description of amendment request: The proposed amendment would 
change the Indian Point 3 Technical Specifications to allow the use of 
zirconium alloy or stainless steel filler rods in fuel assemblies to 
replace failed or damaged fuel rods.
    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:
    Consistent with the criteria of 10 CFR 50.92, the enclosed 
application is judged to involve no significant hazards based on the 
following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
analyzed?
    Response:
    The proposed changes modify the technical specification only to the 
extent that the reconstitution is recognized as acceptable under 
limited circumstances. Reconstitution is limited to substitution of 
zirconium alloy or stainless steel filler rods, and must be in 
accordance with approved applications of fuel rod configurations. 
Although

[[Page 30268]]

these changes permit reconstitution to occur without the need for a 
specific technical specification change, use of an approved methodology 
is required prior to its application. Since the changes will allow 
substitution of filler rods for leaking, potentially leaking rods or 
damaged rods, the changes may actually reduce the radiological 
consequences of an accident. It is noted that the specific changes 
requested in this letter have previously been found acceptable by the 
NRC in GL [Generic Letter] 90-02, Supplement 1. For these reasons, we 
conclude that the changes will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (2) Does the proposed license amendment create the possibility of a 
new or different kind of accident from any previously evaluated?
    Response:
    The proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because they will only affect the assembly configuration and can only 
be implemented if demonstrated to meet current plant requirements in 
accordance with an NRC-approved methodology. The other aspects of plant 
design, operation limitations, and responses to events will remain 
unchanged. It is noted that the changes have previously been determined 
acceptable by the NRC in GL 90-02, Supplement 1.
    (3) Does the proposed amendment involve a significant reduction in 
a margin of safety?
    Response:
    The proposed change will not involve a reduction in a margin of 
safety because the changes can only be implemented if demonstrated to 
meet current plant requirements in accordance with an NRC-approved 
methodology. It is noted that the changes have previously been 
determined acceptable by the NRC in GL 90-02, Supplement 1.
    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 10601.
    Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: April 24, 1998.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Section 3/4.3.1.1, ``Reactor 
Protection System Instrumentation,'' TS Section
3/4.3.2.1, ``Safety Features Actuation System Instrumentation,'' TS 
Section 3/4.3.2.2, ``Steam and Feedwater Rupture Control System 
Instrumentation,'' and the associated TS bases. The TS tables of 
response time limits would be relocated to the Davis-Besse Technical 
Requirements Manual. Other changes in these TS sections consistent with 
the relocation are also proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the 
proposed changes and determined that a significant hazards 
consideration does not exist because operation of the Davis-Besse 
Nuclear Power Station, Unit Number 1, in accordance with these changes 
would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiator, conditions 
or assumptions are affected by the proposed changes to Technical 
Specification (TS) 3/4.3.1.1, Reactor Protection System (RPS) 
Instrumentation, TS 3/4.3.2.1, Safety Features Actuation System (SFAS) 
Instrumentation, and TS 3/4.3.2.2, Steam and Feedwater Rupture Control 
System (SFRCS) Instrumentation and the associated TS Bases to relocate 
their tables of response time limits to the Technical Requirements 
Manual (TRM) of the DBNPS Updated Safety Analysis Report (USAR).
    The RPS, SFAS and SFRCS response time limits and surveillance 
intervals currently prescribed in the TS are not changed under the 
proposed License Amendment. The RPS, SFAS and SFRCS will continue to 
function in the manner described in the DBNPS USAR. Therefore, the 
performance of these protection systems will remain within the bounds 
of the USAR accident analysis.
    Under the proposed changes, the response time limits of the RPS, 
SFAS and SFRCS would continue to be tested in accordance with plant 
procedures in the same manner as in the past. The specific RPS, SFAS 
and SFRCS tables of response time limits will be relocated and remain 
controlled by the TRM of the DBNPS USAR following the guidance of the 
NRC's Generic Letter (GL) 93-08, ``Relocation of Technical 
Specification Tables of Instrument Response Time Limits,'' dated 
December 29, 1993. Any change to the relocated tables for response time 
limits will be subject to review and evaluation under Section 50.59, 
``Changes, Tests, and Experiments,'' of Title 10 of the Code of Federal 
Regulation (10 CFR) prior to any changes being made.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected by the proposed changes. As described above, 
the changes are consistent with the guidance of NRC GL 93-08. The 
proposed changes administratively relocate response time tables and do 
not alter the source term, containment isolation, or allowable 
releases. The proposed changes, therefore, will not increase the 
radiological consequences of a previously evaluated accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new accident 
initiators or assumptions are introduced by the proposed changes, which 
involve only the administrative relocation of response time limit 
tables. No new accident scenarios, transient precursors, failure 
mechanisms, or limiting failures are introduced as a result of the 
proposed changes. As described above, the changes are consistent with 
the guidance of NRC GL 93-08. The proposed changes do not alter any 
accident scenarios and future changes to the response time limits will 
be subject to the regulatory requirements in 10 CFR 50.59.
    3. Not involve a significant reduction in a margin of safety 
because the proposed changes only administratively relocate the 
response time tables from the TS to the USAR TRM, and do not reduce or 
adversely affect the capabilities of any plant structures, systems or 
components. No response times will be changed by this amendment 
request. Future changes to the response time limits will be subject to 
the regulatory requirements of 10 CFR 50.59. Accordingly, there is not 
a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 30269]]

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

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

    Date of amendment request: May 14, 1998.
    Description of amendment request: The proposed Technical 
Specification (TS) amendment would redefine the parent tube pressure 
boundary location for Westinghouse mechanical hybrid expansion joint 
(HEJ) steam generator (SG) tube sleeves. The proposed amendment would 
change the parent tube pressure boundary definition from a minimum 
required interference lip to a minimum required length of non-degraded 
hardroll engagement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change was reviewed in accordance with the provisions 
of 10 CFR 50.92 to show no significant hazards exist. The proposed 
change will not:
    (1) Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    Mechanical testing shows inherent structural integrity of the HEJ 
upper joint such that the requirements of RG 1.121 are met even for 360 
degree, 100 percent throughwall parent tube indications (PTIs). 
Structural test results are documented in WCAP-15050. Based on the test 
data, the structural recommendations of RG 1.121 are satisfied when 
there is a minimum length of non-degraded hardroll which measures 0.92 
inch (plus an allowance for NDE measurement uncertainty) or more from 
the bottom of the hardroll upper transition (HRUT), as measured on the 
inside of the sleeve. Based on the structural integrity of the HEJ 
upper joint, it can be concluded that application of the revised parent 
tube pressure boundary will not result in a significant increase in the 
probability of an accident previously evaluated.
    A conservatively bounding primary-to-secondary steam line break 
(SLB) leak rate of one gpm will be applied to the calculation for 
postulated SLB leakage. This leak rate encompasses all HEJs left 
inservice with PTIs located outside the revised parent tube pressure 
boundary. This one gpm is based on a normal operating leakage limit of 
150 gpd. This leak rate is based on tests and analysis documented in 
WCAP-15050. Application of this leak rate to the postulated leakage 
calculation will ensure primary-to-secondary leakage will not exceed 
the current maximum allowable during a SLB event. Maintenance of the 
current maximum allowable primary-to-secondary leak rate during a SLB 
event ensures off-site doses will not exceed a small fraction of 10 CFR 
100 and control room doses will not exceed GDC-19 criteria. Therefore, 
it can be concluded that the application of the revised parent tube 
pressure boundary will not increase the consequences of an accident 
previously evaluated.
    (2) Create the possibility of a new or different kind of accident 
from any previously evaluated.
    Implementation of the revised parent tube pressure boundary will 
not introduce a change to the design basis or operation of the plant. 
The configuration of the currently installed sleeves is not physically 
changed. As with the initial installation of the sleeves and previous 
changes to the parent tube pressure boundary for HEJs, implementation 
of the revised parent tube pressure boundary does not interact with 
other portions of the reactor coolant system. Neither the sleeve design 
nor the implementation of the revised parent tube pressure boundary 
affects any other component or location of the tube outside of the 
immediate repaired area. Mechanical testing of representative specimens 
supports the conclusions that the joint retains structural integrity 
consistent with RG 1.121 and leakage integrity with regards to 10 CFR 
100 and GDC-19. Any hypothetical accident as a result of potential PTIs 
is bounded by the existing steam generator tube rupture analysis. 
Therefore, application of the revised parent tube pressure boundary 
will not create the possibility of a new or different kind of accident 
from any previously evaluated.
    (3) Involve a significant reduction in the margin of safety.
    The safety factors used in establishment of the HEJ sleeved tube 
pressure boundary are consistent with safety factors in the ASME Boiler 
and Pressure Vessel Code used in the SG design. Based on the sleeve-to-
tube geometry, it is unrealistic to consider that application of the 
revised parent tube pressure boundary could result in single tube leak 
rates exceeding the normal makeup capacity during normal operating 
conditions. The parent tube pressure boundary developed in WCAP-15050 
has been developed using the methodology of RG 1.121. The performance 
characteristics of postulated degraded parent of HEJ sleeve/tube joints 
have been verified through testing to retain structural integrity and 
preclude significant leakage during both normal operating and SLB 
conditions. The existing off-site and control room dose evaluation 
performed for KNPP established a faulted loop primary-to-secondary leak 
rate of 12.85 gpm. Combined leakage from all sources including the 
assumed leak rate for the voltage based repair criteria and for HEJs 
with PTIs that are left inservice will not exceed 12.85 gpm in the 
faulted loop. Maintenance of this limit will ensure off-site doses will 
not exceed a small fraction of the 10 CFR 100 guidelines nor will it 
exceed the GDC-19 criteria for control room dose. Therefore, the 
application of the revised parent tube pressure boundary will not 
result in a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    Acting NRC Project Director: Ronald R. Bellamy.

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

[[Page 30270]]

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

Carolina Power & Light Company, et al., Docket No. 50-325, Brunswick 
Steam Electric Plant, Unit 1, Brunswick County, North Carolina

    Date of application for amendments: February 23, 1998, as 
supplemented March 27, 1998.
    Brief description of amendment: The amendment modifies the values 
for the safety limit for the Minimum Critical Power Ratio (SLMCPR) in 
the TS and the associated action statement for Cycle 12 operation only. 
A reference in TS 6.9.3.2.c is also revised.
    Date of issuance: May 11, 1998.
    Effective date: May 11, 1998.
    Amendment No.: 194.
    Facility Operating License No. DPR-71: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1998 (63 FR 
17900).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 11, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: September 26, 1997, as 
supplemented on April 7, 1998, and May 1, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications to upgrade the ventilation filter testing 
program to the current industry standards and specify that the 
auxiliary electric equipment room is required to be habitable during 
design bases accidents. Revisions related to drywell and suppression 
chamber purge and the editorial changes requested in the September 26, 
1997, application were approved and issued under Amendment Nos. 125 and 
110 dated April 27, 1998.
    Date of issuance: May 13, 1998.
    Effective date: Immediately, to be implemented prior to restart 
from L1F35 for Unit 1 and prior to restart from the current outage for 
Unit 2.
    Amendment Nos.: 126 and 111.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61840). The April 7 and May 1, 1998, submittals provided additional 
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 
May 13, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

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

    Date of application for amendment: July 8, 1994, as supplemented 
August 13, 1996, and February 12, 1998.
    Brief description of amendment: The amendment revises Technical 
Specifications Sections 3.7 and 3.3.E to clarify offsite power 
availability requirement, revise emergency diesel generator fuel oil 
availability requirements and specify the configuration requirements 
for removing Component Cooling Pump 22 from service.
    Date of issuance: May 8, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 196.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42336).
    The August 13, 1996, and February 12, 1998, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of application for amendment: January 18, 1996, as 
supplemented October 1, 1997, and January 29 and April 27, 1998.
    Brief description of amendment: The amendment revises the technical 
specifications regarding inspection requirements for the reactor 
coolant pump (RCP) flywheels. The staff denied a portion of the 
amendment request regarding application to the flywheel testing program 
of the Surveillance Requirement 4.0.2 allowance for surveillance 
interval extension of up to 25%. A separate Notice of Partial Denial of 
Amendment to Facility Operating License and Opportunity for Hearing has 
been published in the Federal Register.
    Date of issuance: May 15, 1998.
    Effective date: May 15, 1998.
    Amendment No.: 182.
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59915).
    The January 29 and April 27, 1998, letters provided additional 
clarifying information that was within the scope of the original 
Federal Register notice and did not change the staff's initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 15, 1998.
    No significant hazards consideration comments received: No.

[[Page 30271]]

    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendment: December 19, 1997, as 
supplemented March 6, 1998.
    Brief description of amendment: This amendment changes the wording 
of Section 4.2.2, ``Terrestrial Ecology Monitoring,'' of the 
Environmental Protection Plan to include completion of the Salt Drift 
Monitoring Program.
    Date of issuance: May 8, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 111.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4321) The March 6, 1998, supplement provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

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

    Date of application for amendments: January 22, 1998, as 
supplemented by letter dated March 18, 1998, April 21, 1998, and May 
15, 1998.
    Brief description of amendments: The amendments change the 
Technical Specifications to allow an extended allowed outage time for 
one emergency diesel generator of 14 days.
    Date of issuance: May 20, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-100; Unit 2-78.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications and Operating Licenses.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6998) The March 18, 1998, April 21, 1998, and May 15, 1998, 
supplements provided clarifying information that did not change the 
scope of the January 22, 1998, application and the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 20, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

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

    Date of amendment request: December 11, 1996, as supplemented by 
letter dated November 3, 1997 (TS-386).
    Description of amendment request: The amendment modifies the 
Appendix A Technical Specifications (TSs) Limiting Safety System 
Setting 2.2.A, which relates to the main steam safety/relief valve set 
points and the set point tolerance. Specifically, the revision 
increases the set point tolerance to <plus-minus>3% vice the current 
<plus-minus> 11 pound per square inch (approximately 1% of set point 
value) tolerance. Bases 1.2 and 3.6D/4.6D also are revised.
    Date of issuance: May 18, 1998.
    Effective date: May 18, 1998.
    Amendment No.: 251 and 210.
    Facility Operating License Nos. DPR-33, DPR-52, and DPR-68. 
Amendment revised the TSs.
    Date of initial notice in Federal Register: January 15, 1997 (62 FR 
2194). The licensee's letter dated November 3, 1997, provided 
additional supporting information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 18, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Athens Public Library, 405 
South Street, Athens, Alabama 35611.

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

    Date of application for amendment: March 20, 1998
    The licensee proposed to modify the licensing basis by limiting the 
time the large (18'') purge and vent valves may be open to 90 hours per 
year. This is a change to the Final Safety Analysis Report (FSAR) and 
technical specification bases.
    Date of Issuance: May 14, 1998.
    Effective date: May 14, 1998.
    Amendment No.: 161.
    Facility Operating License No. DPR-28: Amendment authorizes 
revision to the FSAR.
    Date of initial notice in Federal Register: March 27, 1998 (63 FR 
14976).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated May 14, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

    Dated at Rockville, Maryland, this 27th day of May 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-14519 Filed 6-2-98; 8:45 am]
BILLING CODE 7590-01-P