[Federal Register Volume 59, Number 168 (Wednesday, August 31, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-21325]


[[Page Unknown]]

[Federal Register: August 31, 1994]


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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 August 8, 1994, through August 19, 1994. The 
last biweekly notice was published on August 17, 1994 (59 FR 42332).

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 Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, 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 
20555. The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By September 30, 1994, 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 20555 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, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 20555, and at the local public document 
room for the particular facility involved.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: June 9, 1994
    Description of amendment request: The proposed amendment would 
increase the allowed out-of-service time to increase from 7 days to 14 
days for the automatic depressurization system (ADS), the high pressure 
coolant injection (HPCI) system and the reactor core isolation cooling 
(RCIC)system. The proposed change includes a change to Section 4.5.H, 
``Maintenance of Filled Discharge Pipe'' to reflect Amendment 149, 
issued September 28, 1993.
    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 Pilgrim Station in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Safety criteria used to determine the acceptability of extending 
continued operation with one ADS valve, the HPCI or RCIC system out-
of-service (OOS) is consistent with Pilgrim's licensing basis. For 
example, events with the expected frequency of occurrence greater 
than once-per-reactor lifetime are required to meet the transient 
MCPR [minimum critical power ratio] thermal limit: more than 99.9% 
of the fuel rods are expected to avoid boiling transition. Very low 
probability events, such as a LOCA [loss-of-coolant accident], are 
required to satisfy the criteria of 10CFR50.46: the primary 
criterion being that the Peak Cladding Temperatures (PCT) be 
maintained less than 2200 deg.F.
    For intermediate frequency events, e.g. safe shutdown in the 
event of a fire, 10CFR50 Appendix R involves a ``no fuel damage'' 
criterion. To evaluate these types of events, the GE [General 
Electric] SAFER/GESTR-LOCA licensing methodology was used to 
calculate the system responses and PCTs.
    Analyses performed by Pilgrim's NSSS [nuclear steam supply 
system] vendor, General Electric, [***] for various limiting-case 
scenarios involving ADS, HPCI, or RCIC out-of-service situations 
demonstrated 10CFR50.46 limits (i.e. a PCT less than 2200 deg.F) 
were met. (The most severe PCT was 1500 deg.F). The core damage 
frequency analysis for Pilgrim is unchanged by operating Pilgrim in 
accordance with this proposed amendment. The 14 day OOS for HPCI, 
RCIC and ADS also conforms to the 00S time for these systems found 
in BWR [boiling-water reactor] Standard Technical Specifications. 
Hence, increasing the allowed 00S time from 7 to 14 days does not 
result in a challenge to fuel cladding integrity or BWR Standard 
Technical Specifications, and operating Pilgrim in accordance with 
the proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The removal of the association between LPCI [low-pressure 
coolant injection] and Core Spray system testing and surveilling 
their filled discharge pipes is an administrative change because the 
specified surveillance frequency is unchanged. This proposed change 
reflects Amendment 149, issued by the NRC September 28, 
1993, and is proposed to ensure consistency between Pilgrim's 
Technical Specification sections. This administrative change will 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    As discussed above, a variety of limiting-case scenarios were 
analyzed to demonstrate the effects of increasing the 00S time for 
one ADS valve, the HPCI system, or the RCIC system. The conclusion 
of the analyses is that this proposed change does not violate 
Pilgrim's licensing basis or 10CFR50.46 requirements.
    Some scenarios result in elevated PCTs, but they are still 
significantly below the 10CFR50.46 limit of 2200 deg.F. Therefore, 
since the licensing-basis and code required PCT continues to be met 
and because the proposed change comports the requirements of
    BWR Standard Technical Specifications, operating Pilgrim in
    accordance with the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    As discussed in above question 1, the proposed change to section 
4.5.H.1 is administrative and does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.
    Certain scenarios analyzed for system unavailability result in 
evaluated PCTs. However, these elevated PCTs are significantly below 
the 10CFR50.46 limit of 2200 deg.F. Therefore, there is no reduction 
in the safety margin for PCT resulting from the change from 7 to 14 
days. The proposed change also corresponds to the requirements of 
BWR Standard Technical Specifications concerning 00S for HPCI, RCIC 
and ADS. Therefore, operating Pilgrim Station in accordance with 
this proposed amendment does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis, and based on 
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: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Walter R. Butler

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County,North Carolina

    Date of amendment request: July 22, 1994
    Description of amendments request: The proposed amendment would 
implement a performance based assessment program, including 
corresponding organizational and functional changes. Specifically, the 
changes affect the Independent Review (IR) function, the independent 
assessment of plant activity and the Independent Safety Engineering 
Group. These functions will be performed by the proposed Nuclear 
Assessment Section (NAS). The NAS would perform internal evaluations 
and assessment activities and serve as plant management's staff for the 
objective oversight of plant performance relating to nuclear safety, 
reliability, and quality. The NAS's fundamental role will be to: (1) 
assist plant management in the early identification of issues which may 
prevent the plant from achieving quality performance on a sustained 
basis; and (2) ensure effective correction of deficiencies.
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because it is a programmatic and administrative 
change which does not physically alter any safety-related systems, 
nor does it affect the way in which any safety-related systems 
perform their functions. Since the design of the facility and system 
operating parameters are not changing, the proposed amendment does 
not involve an increase in the probability or consequences of any 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated in Item 1, the proposed amendment is a 
programmatic and administrative change which does not physically 
alter any safety-related systems; nor does it affect the way in 
which any safety-related systems perform their functions. Since the 
design of the facility and system operating parameters are not 
changing, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety because it is a programmatic and 
administrative change which provides assurance that plant operations 
continue to be conducted in a safe manner through the performance 
based assessment programs. As stated in Item 1, the proposed 
amendment does not physically alter any safety-related systems; nor 
does it affect the way in which any safety-related systems perform 
their functions. Since the design of the facility and system 
operating parameters are not changing, the proposed amendment 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

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: July 22, 1994
    Description of amendment request: The proposed amendment would 
implement a performance based assessment program, including 
corresponding organizational and functional changes. Specifically, the 
changes affect the Independent Review (IR) function, the independent 
assessment of plant activity and the Independent Safety Engineering 
Group. These functions will be performed by the proposed Nuclear 
Assessment Section (NAS). The NAS would perform internal evaluations 
and assessment activities and serve as plant management's staff for the 
objective oversight of plant performance relating to nuclear safety, 
reliability, and quality. The NAS's fundamental role will be to: (1) 
assist plant management in the early identification of issues which may 
prevent the plant from achieving quality performance on a sustained 
basis; and (2) ensure effective correction of deficiencies.
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because it is a programmatic and administrative 
change which does not physically alter any safety-related systems, 
nor does it affect the way in which any safety-related systems 
perform their functions. Since the design of the facility and system 
operating parameters are not changing, the proposed amendment does 
not involve an increase in the probability or consequences of any 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated in Item 1, the proposed amendment is a 
programmatic and administrative change which does not physically 
alter any safety-related systems; nor does it affect the way in 
which any safety-related systems perform their functions. Since the 
design of the facility and system operating parameters are not 
changing, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety because it is a programmatic and 
administrative change which provides assurance that plant operations 
continue to be conducted in a safe manner through the performance 
based assessment programs. As stated in Item 1, the proposed 
amendment does not physically alter any safety-related systems; nor 
does it affect the way in which any safety-related systems perform 
their functions. Since the design of the facility and system 
operating parameters are not changing, the proposed amendment 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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

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: July 28, 1994
    Description of amendment request: The proposed amendment would 
change TS Sections 5.3.1.3, 5.4.2.1, 5.4.22, and the Section 5 
references to allow the use of fuel enriched to 4.95 plus 0.05 weight 
percent (w/o) U235.
    The proposed license change is required to support delivery of 
reload batch enrichments anticipated for Cycle 17 and beyond. These 
reloads will require the use of fuel enrichments exceeding the current 
TS limit of 4.20 plus 0.05 weight percent (w/o) U235 (nominal 
4.20).
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. Increasing the allowable U235 enrichment 
will have no influence on the probability of an accident previously 
evaluated. No changes will be made to any safety related equipment, 
systems, or setpoints used in determining the probability of an 
evaluated accident. Neither will the proposed amendment allow 
operation of the facility or safety equipment outside applicable 
limitations or restrictions. Plant design bases will not be altered. 
With respect to the Fuel Handling Accident, the manner in which the 
fuel is handled will not be altered. The heat load on the Spent Fuel 
Pool will not be increased and the cooling and circulation systems 
and equipment will be unaltered. Therefore, there will be no 
significant increase in the probability of an accident previously 
evaluated.
    The proposed change does not increase maximum allowable burnup 
or fission product inventory. Since fission product inventory is an 
inconsequential function of enrichment, radiological consequences 
evaluated in the Updated Final Safety Analysis Report (UFSAR) will 
not increase. The proposed change will not alter the function of 
safety related equipment designed to mitigate the consequences of an 
accident previously evaluated or allow operation of the facility 
outside applicable limitations or restrictions. Accordingly the 
proposed change will not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed increase in allowable enrichment will not 
result in any design, operation, or function changes to any safety 
related equipment designed to prevent and/or mitigate accidents, to 
any setpoints or systems, or to any portion of the plant design 
basis. Operation of the facility will remain within all required 
limitations and restrictions. With respect to the Fuel Handling 
Accident, the manner in which the fuel is handled will not be 
altered. The heat load on the Spent Fuel Pool will not be increased 
and the cooling and circulation systems and equipment will be 
unaltered. 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 amendment does not involve a significant 
reduction in the margin of safety. NRC acceptance criteria and thus 
the acceptable margin of safety to criticality for the Spent Fuel 
Pool and New Fuel Storage Vault criticality are defined in Section 
5.0 of the Technical Specifications. For the Spent Fuel Pool the 
criteria specify that Keff must be maintained less than 0.95 when 
the pit is flooded with unborated water. For the New Fuel Storage 
Vault, the Keff must remain less than 0.95 if the vault is flooded 
with unborated water, and must remain below 0.98 in an optimum 
moderation event. Analyses performed in support of the proposed 
change demonstrate that these acceptance criteria will continue to 
be met. With respect to radiological consequences, the margin of 
safety is defined by 10 CFR [Part] 100 limits which will not be 
challenged. The analyses conclude that fission product inventory and 
thus radiological consequences reported in Chapter 15 of the UFSAR 
will not change. Accordingly the proposed license amendment 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 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: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

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: July 29, 1994
    Description of amendment request: The amendment would allow 
operation of the plant with one Emergency Diesel Generator (EDG) 
inoperable without entering a condition prohibited by Section 3.0 of 
the Technical Specifications (TS). This TS request includes provisions 
to avoid testing the operable EDG altogether under certain conditions 
to ensure that one EDG is available to provide emergency power, if 
needed, and to preserve the EDG overall life and reliability.
    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 amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The proposed change involves changes in the 
testing frequency of the EDGs when one EDG is inoperable, as well as 
provision of additional measures to ensure that a source of off-site 
power is available. The proposed change will also avoid testing of 
an EDG when one EDG is inoperable if the EDG became inoperable for 
reasons other than a common cause. Since the changes involve the 
EDGs which perform an accident mitigation function and are not 
involved in any accident initiation sequence, there is no 
significant increase in the probability of a previously analyzed 
accident. Since the changes involve the EDGs which perform an 
accident mitigation function, and the changes provide additional 
assurance that emergency power will be available for accident 
mitigation, [there] is no significant increase in the consequences 
of a previously analyzed accident. Therefore, there would be no 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The proposed change involves changes in the testing 
frequency of the EDGs when one EDG is inoperable, as well as 
provision of additional measures to ensure that a source of off-site 
power is available. The proposed change will also avoid testing of 
an EDG when one EDG is inoperable if the EDG became inoperable for 
reasons other than a common cause. Since these changes do not 
involve changes in the operation of the plant, or physical or 
equipment changes and involve controls for accident mitigation 
equipment, the proposed amendment will not created the possibility 
of new or different kind of accident from any accident previously 
evaluated. 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 amendment does not involve a significant 
reduction in the margin of safety. The proposed change involves 
changes in the testing frequency of the EDGs when one EDG is 
inoperable, as well as provision of additional measures to ensure 
that a source of off-site power is available. The proposed change 
will also avoid testing of an EDG when one EDG is inoperable if the 
EDG became inoperable for reasons other than a common cause. The 
change reduces the required testing frequency of an operable EDG, 
hence reducing time that no EDG will be available for automatic 
starting and loading. These changes will provide assurance that 
emergency power will be available to mitigate the effects of any 
accident and will prevent excessive wear on the EDGs. 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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of amendment request: July 22, 1993
    Description of amendment request: The proposed amendment would 
allow implementation of a performance based assessment program and the 
corresponding functional and organizational changes in the Nuclear 
Assessment Department. The changes affect the independent review 
function, the independent assessment of plant activity, and the 
independent Safety Engineering Group. These functions will be performed 
by the proposed Nuclear Assessment Section.
    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 change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because it is a programmatic and administrative 
change which does not physically alter any safety-related systems, 
nor does it affect the way in which any safety-related systems 
perform their functions. Since the design of the facility and system 
operating parameters are not changing, the proposed amendment does 
not involve an increase in the probability or consequences of any 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated in Item 1, the proposed amendment is a 
programmatic and administrative change which does not physically 
alter any safety-related systems; nor does it affect the way in 
which any safety-related systems perform their functions. Since the 
design of the facility and system operating parameters are not 
changing, the proposed amendment does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.3. The proposed amendment does not involve a significant 
reduction in the
     margin of safety because it is a programmatic and 
administrative change which provides assurance that plant operations 
continue to be conducted in a safe manner through the performance 
based assessment programs. As stated in Item 1, the proposed 
amendment does not physically alter any safety-related systems; nor 
does it affect the way in which any safety-related systems perform 
their functions. Since the design of the facility and system 
operating parameters are not changing, the proposed amendment 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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of amendment request: August 1, 1994
    Description of amendment request: The proposed amendment would 
revise the technical specifications to incorporate a 1.0 volt steam 
generator tube interim plugging criteria (IPC) for Unit 1 beginning 
with Cycle 7, which will begin in the fall of 1994.
    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.
    Consistent with Regulatory Guide (RG) 1.121, ``Basis for 
Plugging Degraded PWR Steam Generator Tubes, '' Revision 0, August 
1976, the traditional depth-based criteria for SG tube repair 
implicitly ensures that tubes accepted for continued service will 
retain adequate structural and leakage integrity during normal 
operating, transient, and postulated accident conditions. It is 
recognized that defects in tubes permitted to remain in service, 
especially cracks, occasionally grow entirely through-wall and 
develop small leaks. Limits on allowable primary-to-secondary 
leakage established in Technical Specifications ensure timely plant 
shutdown before the structural and leakage integrity of the affected 
tube is challenged.
    The proposed license amendment request to implement voltage 
amplitude SG tube support plate Interim Plugging Criteria for Byron 
Unit 1 meets the requirements of RG 1.121. The IPC methodology 
demonstrates that tube leakage is acceptably low and tube burst is a 
highly improbable event during either normal operation or the most 
limiting accident condition, a postulated main steam line break 
(MSLB) event.
    Adequate SG tube leakage integrity during normal operating 
conditions is assured by limiting allowable primary-to-secondary 
leakage to 150 gpd per SG or 600 gpd total. Currently, this limit is 
administratively controlled. However, a license amendment request 
was submitted on 06/03/94 to incorporate this limit into the Byron 
Technical Specifications. During normal operating conditions, the 
tube support plate constrains the [outer diameter stress corrosion 
cracking] ODSCC affected area of the tube to provide additional 
strength that precludes burst. Any leakage of a tube exhibiting 
ODSCC at the [tube support plate] TSP is fully bounded by the 
existing SG tube rupture analysis included in the Byron UFSAR. 
Therefore, probability of failure of a tube left in service or 
consequences of tube failure during normal operating conditions is 
not significantly increased by the application of IPC.
    During transients, the TSP is conservatively assumed to displace 
due to the thermal-hydraulic loads associated with the transient. 
This may partially expose a crack which is within the boundary of 
the TSP during normal operations to free span conditions. Burst is 
therefore conservatively evaluated assuming the crack is fully 
exposed to free span conditions. The structural eddy current bobbin 
coil voltage limit for free-span burst is 4.54 volts. This limit 
takes into consideration a 1.43 safety factor applied to the steam 
line break differential pressure that is consistent with RG 1.121 
requirements. With additional considerations for growth rate 
assumptions and an upper 95% confidence estimate on voltage 
variability, the maximum voltage indication that could remain in 
service is reduced to 2.7 volts. For added conservatism, the 
allowable indication voltage is further reduced in the proposed 
amendment to a 1.0 volt confirmed ODSCC indication limit. All 
indications between 1.0 and 2.7 volts will be subject to an RPC 
examination. Tubes with RPC confirmed ODSCC indications will be 
plugged or sleeved. Any ODSCC indications between 1.0 volt and 2.7 
volts which are not confirmed as ODSCC will be allowed to remain in 
service since these indications are not as likely to affect tube 
structural integrity or leakage integrity over the next operating 
cycle as the indications that are detectable by both bobbin and 
[rotating pancake coil] RPC inspections.
    The eddy current inspection process has been enhanced to address 
RG 1.83, ``Inservice Inspection of PWR Steam Generator Tubes,'' 
Revision 1, July 1975, considerations as well as the EPRI SG 
Inspection Guidelines. Enhancements in accordance with NUREG-1477 
and Appendix A of the Catawba IPC report (WCAP-13698) are in place 
to increase detection of ODSCC indications and to ensure reliable, 
consistent acquisition and analysis of data. Based on the 
conservative selection of the voltage criteria and the increased 
ability to identify ODSCC, the probability of tube failure during an 
accident is also not significantly increased due to application of 
requested IPC.
    For consistency with current offsite dose limits, the site 
allowable leakage limit during a MSLB has been conservatively 
calculated to be 12.8 gpm. This leakage limit includes maximum 
allowable operational leakage from the unaffected SGs and the 
accident leakage from the affected SG. As a requirement for 
operation following application of IPC, the projected distribution 
of crack indications over the operating period must be verified to 
result in primary to secondary accident leakage less than the site 
allowable leakage limit. Thus, the consequences of a MSLB remain 
unchanged.
    Therefore, as implementation of the 1.0 volt IPC for Byron Unit 
1 does not adversely affect steam generator tube integrity and 
results in acceptable dose consequences, the proposed license 
amendment request does not result in any significant increase in the 
probability or consequences of an accident previously evaluated 
within the Byron Updated Final Safety Analysis Report.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed SG tube IPC does not introduce 
any significant changes to the plant design basis. Use of the 
criteria does not provide a mechanism which could result in an 
accident outside the tube support plate elevations since industry 
experience indicates that ODSCC originating within the tube support 
plate does not extend significantly beyond the thickness of the 
support plate. This criteria only applies to ODSCC contained within 
the region of the tube bounded by the tube support plate.
    In addressing the combined effects of Loss of Coolant Accident 
(LOCA) coincident with a Safe Shutdown Earthquake (SSE) on the SG 
(as required by General Design Criteria 2), it has been determined 
that tube collapse of select tubes may occur in the SGs at some 
plants, including Byron Unit 1. There are two issues associated with 
SG tube collapse. First, the collapse of SG tubing reduces the RCS 
flow area through the tubes. The reduction in flow area increases 
the resistance to flow of steam from the core during a LOCA which, 
in turn, may potentially increase Peak Clad Temperature (PCT). 
Second, there is a potential that partial through-wall cracks in 
tubes could progress to through-wall cracks during tube deformation 
or collapse.
    A number of tubes have been identified, in the ``wedge'' 
locations of the SG TSPs, that demonstrate the potential for tube 
collapse during a LOCA + SSE event. Because of this potential, these 
tubes have been excluded from application of the voltage-based SG 
TSP IPC.
    Therefore, neither a single or multiple tube rupture event would 
be expected in a steam generator in which IPC has been applied.
    ComEd has implemented a maximum primary to secondary leakage 
limit of 150 gpd through any one SG at Byron to help preclude the 
potential for excessive leakage during all plant conditions. The 150 
gpd limit provides for leakage detection and plant shutdown in the 
event of an unexpected single crack leak associated with the longest 
permissible free span crack length. The 150 gpd limit provides 
adequate leakage detection and plant shutdown criteria in the event 
an unexpected single crack results in leakage that is associated 
with the longest permissible free span crack length. Since tube 
burst is precluded during normal operation due to the proximity of 
the TSP to the tube and the potential exists for the crevice to 
become uncovered during MSLB conditions, the leakage from the 
maximum permissible crack must preclude tube burst at MSLB 
conditions. Thus, the 150 gpd limit provides a conservative limit to 
prompt plant shutdown prior to reaching critical crack lengths under 
MSLB conditions.
    Upon implementation of the 1.0 volt IPC, steam generator tube 
integrity continues to be maintained through inservice inspection 
and primary-to-secondary leakage monitoring. Therefore, the 
possibility of a new or different kind of accident from any 
previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The use of the voltage based bobbin coil probe SG TSP IPC for 
Byron Unit 1 will maintain steam generator tube integrity 
commensurate with the criteria of RG 1.121 as discussed above. Upon 
implementation of the criteria, even under the worst case 
conditions, the occurrence of ODSCC at the TSP elevations is not 
expected to lead to a steam generator tube rupture event during 
normal or faulted plant conditions. The distribution of crack 
indications at the TSP elevations result in acceptable primary-to-
secondary leakage during all plant conditions and radiological 
consequences are not adversely impacted by the application of IPC.
    The installation of SG tube plugs and sleeves reduces the RCS 
flow margin. As noted previously, implementation of the SG TSP IPC 
will decrease the number of tubes which must be repaired by plugging 
or sleeving. Thus, implementation of IPC will retain additional flow 
margin that would otherwise be reduced due to increased tube 
plugging. Therefore, no significant reduction in the margin of 
safety will occur as a result of the implementation of this proposed 
license amendment request.
    Although not relied upon to prove adequacy of the proposed 
amendment request, the following analyses demonstrate that 
significant conservatisms exist in the methods and justifications 
described above:
    LIMITED TUBE SUPPORT PLATE DISPLACEMENT
    An analysis was performed to verify [the effect] of limited TSP 
displacement during accident conditions (MSLB). Application of 
minimum TSP displacement assumptions reduce the likelihood of a tube 
burst to negligible levels. Consideration of limited TSP 
displacement would also reduce potential MSLB leakage when compared 
to the leakage calculated assuming free span indications.
    PROBABILITY OF DETECTION
    The Electric Power Research Institute (EPRI) Performance 
Demonstration Program analyzed the performance of approximately 20 
eddy current data analysts evaluating data from a unit with 3/4'' 
inside diameter and 0.049'' wall thickness tubes. The results of 
this analysis clearly show that the detectability of larger voltage 
indications is increased which lends creditability for application 
of a POD of  0.62 for ODSCC indications larger than 1.0 
volt.
    RISK EVALUATION OF CORE DAMAGE
    As part of ComEd's evaluation of the operability of Byron Unit 1 
Cycle 7, a risk evaluation was completed. The objective of this 
evaluation was to compare core damage frequency under containment 
bypass conditions, with and without the interim plugging criteria 
applied at Byron Unit 1.
    The total Byron core damage frequency is estimated to be 3.09E-5 
per reactor year with a total contribution from containment bypass 
sequences of 3.72E-8 per reactor year according to the results of 
the current individual plant evaluation (IPE). Operation with the 
requested IPC resulted in an insignificant increase in core damage 
frequency resulting from MSLB with containment bypass conditions.
    Therefore, based on the evaluation above, ComEd has concluded 
that this proposed license amendment request does not involve a 
significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Byron Public Library, 109 N. 
Franklin, P.O. Box 434, Byron, Illinois 61010
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, 
Illinois;Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power 
Station, Units 1 and 2, Rock Island County, Illinois;Docket Nos. 
50-295 and 50-304, Zion Nuclear Power Station, Units 1 and 2, Lake 
County, Illinois

    Date of amendment request: July 8, 1994
    Description of amendment request: The proposed amendment would add 
a License Condition to specify that commitments made in response to the 
March 14, 1983, NUREG-0737 Order shall be maintained pursuant to the 
requirements of 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:
    Involve a significant increase in the probability or 
consequences of any accident previously analyzed:
    Commonwealth Edison has addressed all issues made in response to 
NUREG-0737. As such, the purpose of the post-TMI Order is no longer 
served. The inclusion of the modified Order as a license condition 
is administrative in nature and does not allow unregulated decreases 
in the level of safety; therefore, this license amendment is 
appropriate and safe. The proposed license amendment requires 
control of NUREG-0737 commitments through 10 CFR 50.59. If an 
unreviewed safety question occurs during the review of a NUREG 0737 
item then Commonwealth Edison is obligated to submit a change to the 
NRC staff as a license amendment. As a result of the proposed 
amendment, there are no physical changes to the facility and all 
operating procedures, limiting conditions for operation (LCO), 
limiting safety system settings, and safety limits specified in the 
Technical Specifications will remain unchanged. Therefore, the 
proposed license amendment to modify the post-TMI Order will not 
increase the probability or the consequences of any accident 
previously analyzed.
    Create the possibility of a new or different kind of accident 
from any previously evaluated:
    Since there are no changes in the way the plant is operated, the 
potential for a new or different kind of accident is not created. 
The proposed changes are administrative in nature and do not affect 
any accident initiators for Dresden, Quad Cities, and Zion Stations. 
No new failure modes are introduced.
    Involve a significant reduction in a margin of safety:
    Plant safety margins are established through LCOs, limiting 
safety system settings, and safety limits specified in the Technical 
Specifications. As a result of the proposed amendment, there will be 
no changes to either the physical design of the plant or to any of 
these settings and limits. The proposed changes are administrative 
and do not affect the safe operation of the sites. Therefore, there 
will be no changes to any of the margins of safety.
    Guidance has been provided in 51 FR 7744 for the application of 
standards to license change requests for determination of the 
existence of significant hazards considerations. This document 
provides examples of amendments which are not likely considered to 
involve significant hazards considerations.
    This proposed amendment does not involve a significant 
relaxation of the criteria used to establish safety limits, a 
significant relaxation of the bases for the limiting safety system 
setting or a significant relaxation of the bases for the limiting 
conditions for operations. The proposed changes are administrative 
in nature without consequence to the safety of the plant. Therefore, 
based on the guidance provided in the Federal Register and the 
criteria established in 10 CFR 50.92(c), the proposed change does 
not constitute a significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: for Dresden, Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; for Quad Cities, 
Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021; for 
Zion, Waukegan Public Library, 128 N. County Street, Waukegan, Illinois 
60085
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: Robert A. Capra

Connecticut Yankee Atomic Power Company, and Northeast Nuclear 
Energy Company, Docket Nos. 50-213, 50-245, 50-336, and 50-423 
Haddam Neck Plant, and Millstone Nuclear Power Station, Units 1, 2, 
and 3, Middlesex County, and New London County, Connecticut

    Date of amendment request: June 30, 1994
    Description of amendment request: The proposed amendments would 
modify the Administrative Controls Section of the Technical 
Specifications by replacing the present Nuclear Review Board (NRB) for 
the Haddam Neck Plant, and the NRB and Site Nuclear Review Board (SNRB) 
with a Nuclear Safety Assessment Board (NSAB) which will serve 
Millstone Units 1, 2, and 3, and Haddam Neck.
    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 (SHC), which is presented below:
    ... These 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 NSAB is an oversight group which provides independent 
assessments of activities at the Haddam Neck Plant and Millstone 
Unit Nos. 1, 2, and 3. The members of the NSAB are appointed by the 
Executive Vice President - Nuclear to provide oversight and feedback 
on the operation of the units. The NSAB adds to the defense-in-depth 
provided by the design, operation, maintenance, and quality 
oversight of the nuclear units by promoting excellence through the 
conduct of its affairs and advising the Executive Vice President - 
Nuclear in matters concerning nuclear safety.
    The proposed modification to the Technical Specifications are 
administrative in nature and will establish a new group which will 
accomplish the guidance provided in ANSI N18.7-1976. The charter of 
the NSAB will be controlled by procedure.
    These administrative changes will not increase the probability 
of occurrence or the consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed addition of the NSAB and its subcommittees and the 
ensuing elimination of the NRB and the SNRB is an administrative 
reorganization. There are no changes in the way in which the plants 
are physically operated. The administrative changes being 
accomplished by the establishment of the NSAB fulfills the function 
previously provided by the NRB and the SNRB. The organization of the 
NSAB will follow the guidance found in ANSI N18.7-1976 and will be 
controlled by procedure.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes establish the requirements of the NSAB. The 
NSAB replaces those activities previously performed by the NRB and 
the SNRB. With these changes the new organization will provide more 
consistent and clearer feedback to the four units and the Executive 
Vice President - Nuclear.
    The changes do not directly affect any protective boundaries nor 
do they impact the safety limits for the protective boundaries. 
These proposed changes are administrative in nature. Therefore, 
there can be no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, Connecticut 06457, for the Haddam Neck Plant, and 
the Learning Resource Center, Three Rivers Community-Technical College, 
Thames Valley Campus, 574 New London Turnpike, Norwich, Connecticut 
06360, for Millstone 1, 2 and 3.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, 
Connecticut, 06141-0270.
    NRC Project Director: John F. Stolz

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: August 11, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 6.5.1, Station Nuclear Safety 
Committee (SNSC), to change the designation of the Chairman and to 
clarify the maximum allowable alternate members for quorum purposes.
    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 does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    This is an administrative change. Since this change maintains a 
consistent level of chairmanship while continuing to ensure 
independence and technical expertise of the SNSC chairman, this 
change does not increase the probability or consequences of an 
accident.
    2. The possibility of a new or different kind of accident from 
any previously evaluated has not been created.
    This is an administrative change of the designation of the 
Chairman of SNSC which does not significantly decrease the level of 
senior management which is responsible for chairing SNSC. No new or 
different kind of accident has been created.
    3. There has been no reduction in the margin of safety.
    The independence and technical expertise of the SNSC Chairman 
will be preserved. SNSC will continue to be composed of those 
individuals most related to matters of nuclear safety. The margin of 
safety will not be reduced by this 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: Pao Tsin Kuo

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

    Date of amendment request: May 24, 1994
    Description of amendment request: The proposed amendments would 
transfer the boron concentration in Technical Specification (TS) 3.9.1 
for the reactor coolant system and the refueling canal during MODE 6, 
and the boron concentration in TS 4.7.13.3 for the spent fuel pool from 
the TS to the Core Operating Limits Report (COLR). The application is 
submitted in response to the guidance in Generic Letter 88-16 which 
addresses the transfer of fuel cycle-specific parameter limits from the 
TS to the COLR.
    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 following analysis, performed pursuant to 10 CFR 50.91, 
shows that the proposed amendment will not create a significant 
hazards consideration as defined by the criteria of 10 CFR 50.92.
    1. This amendment will not significantly increase the 
probability or consequence of any accident previously evaluated.
    No component modification, system realignment, or change in 
operating procedure will occur which could affect the probability of 
any accident or transient. The relocation of boron concentration 
values to the COLR is an administrative change which will have no 
effect on the probability or consequences of any previously-analyzed 
accident. The required values of boron concentration will continue 
to be determined through use of approved methodologies.
    2. This amendment will not create the possibility of any new or 
different accidents not previously evaluated.
    No component modification or system realignment will occur which 
could create the possibility of a new event not previously 
considered. The administrative change of relocating parameters to 
the COLR, in this case boron concentration, cannot create the 
probability of an accident.
    3. This amendment will not involve a significant reduction in a 
margin of safety.
    Required boron concentrations will remain appropriate for each 
cycle, and will continue to be calculated using approved 
methodologies. 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: February 5, 1993 as supplemented by 
letter dated August 1, 1994.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to incorporate a technical review 
and control process to supplement the onsite technical review and 
approval of new procedures and changes thereto affecting nuclear 
safety. This process is discussed in Section 5.5 of the Revised 
Standard Technical Specifications, NUREG-1432. This notice supersedes 
the notice issued on April 14, 1993 (58 FR 19478), and acknowledges the 
clarification in the licensee's August 1, 1994, letter.
    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 is administrative in nature and provides for 
1) procedural reviews through the use of qualified technical review 
personnel designated by the PORC [Plant Operating Review Committee] 
and 2) procedural approval through the use of group heads designated 
by the General Manager Plant Operations as authorized by 
administrative controls upon their development. As part of this 
process, qualified technical reviewers will be individuals other 
than the preparer who will document and implement necessary cross-
discipline reviews prior to approval. The process will be controlled 
by administrative controls which will be reviewed by the PORC and 
approved by the General Manager Plant Operations.
    The procedures governing plant operation will continue to ensure 
that plant parameters are maintained within acceptable limits. 
Procedures and changes thereto will be reviewed and approved at a 
level commensurate with their importance to safety. Therefore, the 
proposed changes will not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    The proposed changes are administrative in nature. The proposed 
changes do not involve physical changes to the plant, changes to 
setpoints, or operating parameters. The applicable procedures 
governing the operation of the plant will receive reviews and 
approvals commensurate with their importance to nuclear safety, and 
where appropriate cross-discipline review will be performed. 
Therefore, the proposed changes will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The proposed changes are administrative in nature. The Waterford 
3 safety margins are defined and maintained by the Technical 
Specifications in Sections 2-5 which are unaffected. Therefore, the 
proposed change will not involve a significant reduction in a margin 
of safety.
    The licensee's letter dated August 1, 1994, provided a 
clarification of the proposed wording of the technical specifications 
to assure the personnel performing the technical reviews would have the 
necessary technical knowledge base.
    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 New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: August 19, 1994
    Description of amendment request: The proposed amendment would move 
the requirements of Technical Specification 3/4.3.4 Turbine Overspeed 
Protection from the technical specifications (TS) and relocate them in 
the Updated Final Safety Analysis Report (UFSAR) consistent with the 
NRC Final Policy Statement on Technical Specifications Improvements for 
Nuclear Power Reactors.
    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 relocates the Turbine Valve Overspeed 
Protection requirements from the TS to the Waterford 3 UFSAR 
consistent with the NRC Policy Statement on Technical Specification 
Improvements. Testing and inspections of the turbine Overspeed 
Protection System will remain governed by an approved turbine 
maintenance program, described in the UFSAR. This proposed change 
has no affect on the current Turbine Overspeed Protection 
requirements other then to relocate them to the UFSAR. Thus, the 
probability of a turbine missile causing damage to a safety-related 
component or structure at Waterford 3 as described in the FSAR 
analysis (Reference 5) is not affected. The purpose of the Turbine 
Overspeed Protection System is to prevent an overspeed event, the 
precursor to a potential turbine fragment missile. Since the purpose 
of this system is preventive, it serves no function to mitigate any 
accident previously evaluated.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed change does not involve any change to the 
configuration or method of operation of any plant equipment. No new 
failure modes or limiting failures have been identified as result of 
the proposed change. The proposed change will not alte the operation 
of the plant or the manner in which it is operated. Any subsequent 
change to the Turbine Oversspeed Protection System requirements will 
undergo a review in accordance with the criteria of 10 CFR 50.59 to 
ensure that the change does not involve an unreviewed safety 
question.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident prveiously 
evaluated.
    The proposed change will relocate Turbine Overspeed Protection 
System requirements from the TS to the Waterford 3 UFSAR on the 
basis that the Turbine Overspeed Protection System does not meet the 
criteria of the NRC Final Policy Statement on Technical 
Specifications Improvements for Nuclear Reactors. The requirements 
that will reside in the UFSAR for the Turbine Overspeed Protection 
system will ensure that the system remains capable of protecting the 
turbine from excessive overspeed. The proposed change will have no 
adverse impact on any protective boundary or safety limit.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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 New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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: July 25, 1994
    Description of amendment request: The proposed amendments evise 
various Technical Specification sections to implement enhancements 
recommended by NRC Generic Letter (GL) 93-05, ``Line-Item Technical 
Specification Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation,'' for St. Lucie Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The standards used to arrive at a determination that a request 
for amendment involves a no significant hazards consideration are 
included in the Commission's regulation, 10 CFR 50.92. 10 CFR 50.92 
states that no significant hazards considerations are involved if 
the 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. Each standard is discussed as 
follows:
    (1) Operation of the facility in accordance with the proposed 
amendments 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 amendments conform to the guidance given in 
Enclosure 1 of the NRC Generic Letter 93-05. The overall functional 
capabilities of the incore detector system, reactor coolant system 
pressure isolation valves, safety injection tank, or containment 
sump will not be modified by the proposed change. Therefore, the 
probability or consequences of an accident are not significantly 
increased by the changes.
    (2) Use of the modified specification would not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The use of the modified specifications can not create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
the surveillance interval changes and clarifications, 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.
    (3) Use of the modified specification would not involve a 
significant reduction in a margin of safety.
    The operating limits and functional capabilities of the affected 
systems are unchanged by the proposed amendments. Therefore, the 
modified specifications which establish new or clarify old 
surveillance intervals consistent with the NRC Generic Letter 93-05 
line-item improvement guidance do not significantly reduce any of 
the margins of safety.
    Based on the above, we have determined that the proposed 
amendments do not (1) involve a significant increase in the 
probability or consequences of an accident previously evaluated, (2) 
create the probability of a new or different kind of accident from 
any previously evaluated, or (3) involve a significant reduction in 
a margin of safety; and therefore do not involve a significant 
hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Victor McCree, Acting

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: July 25, 1994
    Description of amendment request: The amendment will upgrade 
Technical Specification (TS) 3/4.7.1.6 for the Main Feedwater Line 
Isolation Valves to be consistent with NUREG-1432, ``Standard Technical 
Specifications for Combustion Engineering Plants.'' The changes include 
all related requirements of NUREG-1432, Revision O, specification 
3.7.3. Accordingly, the proposal is consistent with the Commission's 
Final Policy Statement on Technical Specifications Improvements (58 FR 
39132).
    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:
    Pursuant to 10 CFR 50.92, a determination may be made that a 
proposed license amendment involves no significant hazards 
consideration if 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. Each standard is discussed as 
follows:
    (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 will upgrade the existing Limiting 
Condition for Operation (LCO) associated with the Main Feedwater 
Line Isolation Valves (MFIVs) to be consistent with NUREG-1432, 
Standard Technical Specifications for Combustion Engineering Plants. 
The MFIVs are not initiators of accidents previously evaluated, but 
are included as part of the success paths associated with mitigating 
various accidents and transients. The redundancy afforded by two 
MFIVs per feedwater line in conjunction with the requirements of the 
proposed LCO assure that the feedwater isolation safety function of 
these valves can be accomplished considering single failure 
criteria. Neither the feedwater system design nor the safety 
function of the MFIVs have been altered from those previously 
evaluated, and the proposed amendment does not change the applicable 
plant safety analyses.
    Therefore, operation of the facility in accordance with the 
proposed amendment will 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 proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. The changes are 
administrative in nature in that they do not involve the addition of 
new equipment or the modification of existing equipment, nor do they 
otherwise alter the design of St. Lucie Unit 2 systems. 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 safety function of the MFIVs is to terminate main feedwater 
flow and isolate the safety related portion from the non-safety 
related portion of the feedwater system. The proposed amendment, in 
conjunction with the redundancy afforded by the feedwater system 
design, assures that this safety function can be accomplished 
considering single-failure criteria. The bases for required actions 
and the action completion times specified for inoperable MFIVs is 
consistent with the corresponding specifications in NUREG-1432, 
which are equally applicable to St. Lucie Unit 2. The safety 
analyses for applicable accidents and transients remain unchanged 
from those previously evaluated and reported in the Updated Final 
Safety Analysis Report. Therefore, operation of the facility in 
accordance with the proposed amendment would not involve a 
significant reduction in a margin of safety.
    Based on the discussion presented above and on the supporting 
Evaluation of Proposed TS Changes, FPL has concluded that this 
proposed license amendment involves no significant hazards 
consideration.
    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: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Victor McCree, Acting

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
Matagorda County, Texas

    Date of amendment request: July 18, 1994
    Description of amendment request: The licensee proposes to revise 
Technical Specification Table 4.3-1, Reactor Trip System 
Instrumentation Surveillance Requirements, Technical Specification 
3.3.4, Turbine Governor Valves and Technical Specification 3.7.1.2, 
Turbine Driven Auxiliary Feedwater Pump. The purpose of this amendment 
is to remove one-time amendments that are no longer necessary. In 
addition, six minor editorial changes are proposed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of a previously evaluated 
accident.
    The changes proposed to remove the one-time amendments return 
the Technical Specifications to the exact wording prior to the one-
time amendments. Returning the Technical Specifications to their 
original wording is administrative because the one-time amendments 
are no longer applicable. Hence, removing the one-time amendments 
would not increase the probability or consequences of an accident. 
The other changes are purely editorial in nature, hence, would not 
increase the probability or consequences of an accident. Based on 
the above, removal of the one-time amendments from the Technical 
Specifications will not significantly increase the probability or 
consequences of an accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The changes proposed to remove the one-time amendments return 
the Technical Specifications to the exact wording prior to the one-
time amendments. Returning the Technical Specifications to their 
original wording is administrative because the one-time amendments 
are no longer applicable. Therefore, removing the one-time 
amendments would not create the possibility of a new or different 
kind of accident. The other changes are purely editorial in nature, 
hence, would not create the possibility of a new or different kind 
of accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The changes proposed to remove the one-time amendments return 
the Technical Specifications to the exact wording prior to the one-
time amendments. Returning the Technical Specifications to their 
original wording is administrative because the one-time amendments 
are no longer applicable. Therefore, removing the one-time 
amendments would not involve a significant reduction in a margin to 
safety. The other changes are purely editorial in nature, hence, 
would not involve a significant reduction in a margin to safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location:  Wharton County Junior 
College, J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, 
Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: William D. Beckner

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

    Date of amendment request: July 12, 1994
    Description of amendment request: The proposed amendment changes 
the requirement to perform the surveillance test for the channel 
functional test Rod Block Monitor, Flow-biased Average Power Range 
Monitor and Recirculation Flow instruments from within 24 hours prior 
to startup to after the reactor is in the RUN mode, but prior to when 
each system is assumed to function in the plant safety analysis.
    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 to the Channel Functional Test frequency 
for the RBM will not significantly increase the probability or 
consequences for any previously-evaluated event as we are only 
matching the mode requirements for performing the SR to the 
OPERABILITY requirement for the RBM system, i.e., prior to 30% RTP. 
The system will be verified to be OPERABLE prior to when it is 
assumed to be OPERABLE in the Updated Final Safety Analysis Report 
(UFSAR) for the DAEC.
    Allowing the Channel Functional Test for the APRM Flow-Biased 
Rod Block Upscale and Downscale trips to be performed ``within 24 
hours of entering RUN mode and prior to exceeding 25% RTP'' will not 
increase either the probability or consequences of any previously-
analyzed event. The applicable event for the rod block function 
during reactor startup is a control Rod Withdrawal Error (RWE), 
which is initiated by either an operator error or malfunction within 
the Reactor Manual Control System, not by a malfunction within the 
APRM system. However, a RWE event that could challenge the fuel 
thermal limits is precluded because, as documented in the DAEC UFSAR 
(see Section 15.4) and the analysis submitted to support DAEC TS 
Amendment No. 120,( NEDC-30813-P,
    Average Power Range Monitor, Rod Block Monitor and Technical 
Specification Improvement (ARTS) Program for the Duane Arnold Energy 
Center, December 1984.), significant margin exists below 25% RTP to 
assure the Safety Limit Minimum Critical Power Ratio (SLMCPR) is not 
violated by a RWE event. In addition, rod pattern controls are in 
place during this period to limit the rod withdrawal sequence, i.e., 
rod worth, such that the fuel thermal limits would not be exceeded. 
The Control Rod Drop Accident is unaffected by the requested SR 
change as the ``accident'' control rod is assumed to be de-coupled 
from its drive mechanism and free-falls from fully inserted to 
fully-withdrawal. As the drive for that rod is assumed to be fully-
withdrawn as an initial condition in the event, the APRM rod block 
has no role in either preventing or mitigating the rod drop 
accident. Thus, revising the SR for the APRM Flow-Biased Rod Block 
has no impact upon the Control Rod Drop Accident.
    The SRs for the Recirculation Flow Rod Block trips are being 
modified for consistency with the APRM Rod Block changes above, as 
the sole purpose of this Recirculation Flow signal is to provide the 
flow input signal into the APRM Flow-Biased trips. The Recirculation 
Flow units are a support system to the APRM Flow-Biased Rod Blocks. 
There is no event that is either caused by or mitigated by the 
Recirculation Flow Rod Block trips. They are provided solely to 
ensure that if the flow signal being input into the APRM circuits is 
not valid, a precautionary rod block will be generated as the APRM 
Flow-Biased Rod Block setpoint could be in error. Consequently, 
allowing the Channel Functional Test for the Recirculation Flow Rod 
Block Upscale, Downscale and Comparator trips to be performed 
``within 24 hours of entering RUN mode and prior to exceeding 25% 
RTP'' will not increase either the probability or consequences of 
any previously-analyzed event as these rod blocks are not involved 
in either preventing or mitigating any analyzed event.
    2) The proposed change to the Channel Functional Test frequency 
for the RBM will not introduce any new or different event, as no 
changes in system design or operation are being made. We are only 
matching the requirement for performing the SR to the OPERABILITY 
requirement for the RBM system.
    The proposed change to the Channel Functional Test frequency for 
the APRM and Recirculation Flow Rod Blocks will not introduce any 
new or different event, as no changes in either system design or 
operation are being made. In fact, by allowing the Channel 
Functional Test to be performed in an operating state which does not 
require extensive use of jumpers and/or relay blocks, we reduce the 
possibility of an error being made that could cause an inadvertent 
actuation of an ESF or disabling of an ESF.
    3) The proposed change matches the mode requirement for 
performing the SR to the OPERABILITY requirement for the RBM system,
    i.e., prior to 30% RTP. The system will be verified to be 
OPERABLE prior to when it is assumed to be OPERABLE in the UFSAR 
accident analysis. Thus, the margin of safety for the RBM is not 
reduced.
    As stated in the BASES for TS Chapter 3/4.2, the margin of 
safety for the APRM rod block is to prevent violation of the SLMCPR 
in RUN by a RWE event. The analysis of the RWE event during Startup 
(See DAEC UFSAR Section 15.4.2) and during Power Operation (Ibid), 
demonstrates that violations of the SLMCPR are not possible in RUN 
below 25% RTP when normal control rod patterns are followed (which 
are reinforced by procedural and/or automatic rod pattern controls). 
Because the proposed change to the SR for the APRM Flow-Biased Rod 
Block will still ensure that the trip will be OPERABLE prior to 
exceeding 25% RTP, this change will not reduce the existing margin 
of safety.
    Again, the Recirculation Flow units are a support system to the 
APRM flow-biased circuits. The Recirculation Flow Rod Blocks are 
merely precautionary, they do not prevent or mitigate any accident. 
Therefore, the proposed revision to the Recirculation Flow Rod block 
SR frequency will not reduce the margin of safety for the same 
reasons given above for the APRM Rod Blocks.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036.
    NRC Project Director: John N. HannonIES Utilities Inc., Docket No. 
50-331, Duane Arnold Energy Center, Linn County, Iowa
    Date of amendment request: July 29, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications by allowing the processing and 
implementation of an ISI or IST request for relief from the ASME Code 
under 10 CFR 50.59 without prior NRC approval, provided that the relief 
request has been reviewed and approved by the plant staff and plant 
safety committee.
    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 any increase in the probability of 
occurrence or consequences of an accident previously evaluated. The 
Inservice Inspection and Testing Programs, pursuant to 10 CFR 
50.55a, are described in the Technical Specifications. The proposed 
amendment, in accordance with NUREG-1433 and draft NUREG-1482, 
permits relief from an ASME Code requirement in the interim between 
the time of submittal of a relief request and NRC approval of the 
relief. The changes being proposed do not affect assumptions 
contained in plant safety analyses or change the physical design 
and/or operation of the plant, nor do they affect Technical 
Specifications that preserve safety analysis assumptions. Any relief 
from the approved ASME Section XI Code requirements that is 
implemented prior to NRC review and approval will require evaluation 
under the 10 CFR 50.59 process to determine that no TS changes or 
unreviewed safety questions exist. This evaluation process will 
ensure that the impact of any Code relief is thoroughly evaluated 
and that the structures, systems and components remain in 
conformance with assumptions made in the safety analysis. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not affect 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 
Inservice Inspection and Testing Programs, pursuant to 10 CFR 
50.55a, are described in the Technical Specifications. The proposed 
amendment, in accordance with NUREG-1433 and draft NUREG-1482, 
permits relief from an ASME Code requirement in the interim between 
the time of submittal of a relief request and NRC approval of the 
relief. The changes being proposed will not change the physical 
plant or the modes of operation defined in the Facility License. The 
changes do not involve the addition or modification of equipment nor 
do they alter the design or operation of plant systems. Any relief 
from the approved ASME Section XI Code requirements that is 
implemented prior to NRC review and approval will require evaluation 
under the 10 CFR 50.59 process to determine that no TS changes or 
unreviewed safety questions exist. This evaluation process will 
ensure that the impact on any Code relief is thoroughly evaluated 
and that the structures, systems and components remain in 
conformance with assumptions made in the safety analysis. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not create the possibility of a new or different kind of 
accident previously evaluated.
    3) Operation of the facility in accordance with the proposed 
amendment would not involve any reduction in a margin of safety. The 
Inservice Inspection and Testing Programs, pursuant to 10 CFR 
50.55a, are described in the Technical Specifications. The proposed 
amendment, in accordance with NUREG-1433 and draft NUREG-1482, 
permits relief from an ASME Code requirement in the interim between 
the time of submittal of a relief request and NRC approval of the 
relief. The changes being proposed do not alter the bases for 
assurance that safety-related activities are performed correctly or 
the basis for any TS that is related to the establishment of or 
maintenance of a safety margin. Any relief from the approved ASME 
Section XI Code equirements that is implemented prior to NRC review 
and approval will require evaluation under the 10 CFR 50.59 process 
to determine that no TS changes or unreviewed safety questions 
exist. This evaluation process will ensure that the impact on any 
Code relief does not affect the ability of structures, systems or 
components to perform their design function, affect compliance with 
any TS requirements or reduce the margin of safety. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not involve a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036
    NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: July 26, 1994
    Description of amendment request: The proposed amendment would 
revise the existing limiting condition for operation (LCO) 3.12.A.2.c 
to allow for increased flow capacity of the control room emergency 
filter system. By increasing the maximum allowed makeup capacity of 
this system, additional margin is provided for the positive 
pressurization of the control room envelope.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Evaluation
    This license amendment request involves the upgrading of the 
Control Room Emergency Filter System from 341 cubic feet per minute 
(CFM) plus or minus 10% to a maximum of < 1000 CFM. By establishing 
a new maximum flowrate for this system, additional filtered makeup 
air can be supplied to the Control Room, thus increasing the 
positive pressure in the Control Room envelope.
    The purpose ofthe Control Room Emergency Filter System is to 
remove radioactive iodine and other radioactive materials from the 
makeup air during design basis accidents. Therefore, any change to 
this system will not increase the probability of an accident 
previously evaluated. Radiological calculations show that the 
increased flowrate of this system will not result in a significant 
increase in Control Room operator dose during a design basis 
accidents, and these doses remain well below the established limits. 
Therefore, the consequences of an accident previously evaluated are 
not significantly increased. The addition of the new Surveillance 
Requirement provides a Technical Specification required periodic 
demonstration of the positive pressurization function of the system. 
This requirement has previously been implemented per existing 
surveillance procedures as part of the overall Control Room 
Emergency Filter System operability demonstration, and does not 
represent a new requirement. This proposed change does not introduce 
any new modes of plant operation nor affect any operational 
setpoints. The change does not degrade the performance of any safety 
system assumed to function in the accident analysis. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility for a new or 
different kind of accident from any accident previously evaluated?
    Evaluation
    This license amendment request involves the upgrading of the 
Control Room Emergency Filter System from 341 cubic feet per minute 
(CFM) plus or minus 10% to a maximum of < 1000 CFM. This proposed 
change involves a physical modification to the Control Room 
Emergency Filter System where the filter fan is replaced with a new 
fan with greater capacity. To accommodate the additional flow 
capacity of the system, an additional charcoal tray is installed in 
the charcoal adsorber unit. The District has evaluated the potential 
effects of this modification and has determined that the increased 
air flowrate is within the system capacity and that radiological 
doses, through the filter system, during the design basis accident 
are largely unaffected. Because this is a modification of an 
existing system with no direct interface with other systems 
responsible for prohibiting or mitigating design basis events, the 
District has concluded that this proposed change cannot create the 
possibility for a new or different kind of accident. This proposed 
change does not involve the creation, deletion, or modification of 
the function of any structure or system, except as described above, 
nor does this change introduce or change any mode of plant 
operation. This proposed change does not create the possibility for 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change create a significant reduction in 
the margin of safety?
    Evaluation
    This license amendment request involves the upgrading of the 
Control Room Emergency Filter System from 341 cubic feet per minute 
(CFM) plus or minus 10% to a maximum of < 1000 CFM. By establishing 
this new maximum flowrate for the Emergency Filter System, 
additional filtered makeup air can be supplied to the Control Room 
envelope, thus improving the margin for positive pressurization with 
respect to adjacent areas. Recent tests, utilizing the Main Control 
Room Air Conditioning System, have provided information that 
supports an increase in the Emergency Filter System flowrate, from 
341 CFM to the proposed maximum system capability of approximately 
1000 CFM. These tests indicate that an increase of positive pressure 
can be achieved by this increased flowrate. This positive pressure 
increase provides additional margin of Control Room envelope 
positive pressure.
    The District has performed radiological calculations to 
determine the increase in Control Room operator dose during the 30-
day design basis LOCA event, as a result of increased system air 
flow. These calculations show increasing the Control Room Emergency 
Filter System to a maximum of 1000 CFM results in a dose of 1.799 
Rem whole body and 12.81 Rem thyroid. These doses are not 
significantly different than the doses received at a system flowrate 
of 341 CFM, which is 1.745 Rem whole body and 11.39 Rem thyroid. 
These doses are well within the limits of 10 CFR 20, 10 CFR 50, 
Appendix A, General Design Criteria 19, and the guidance provided in 
NUREG 0800, which require that doses be limited to less than 5 Rem 
whole body, or its equivalent to any part of the body including 30 
Rem thyroid, for the duration of any design basis accident. The 
above calculated values have also been evaluated and determined to 
be within the Updated Safety Analysis Report (USAR) Section XII 
requirement of 0.5 Rem in any eight-hour period, whole body from the 
reactor building. Increasing the Control Room Emergency Filter 
System maximum flowrate has a minimal effect on quantifiable dose 
rates, while increasing positive pressurization in the Control Room 
envelope. By increasing the positive pressurization in the Control 
Room envelope, the possibility of non quantifiable radiation dose to 
the Control Room operators, through inleakage, is reduced. This 
proposed change does not involve any change to instrument setpoints 
or operation. Therefore, the District has concluded that this 
proposed change does not create 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: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305
    Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power 
District, Post Office Box 499, Columbus, Nebraska 68602-0499
    NRC Project Director: William D. Beckner

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

    Date of amendment request: July 21, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 2.2.2 (Reactor Coolant System), 
3.2.8/4.2.8 (Pressure Relief Systems - Safety Valves), and the 
associated Bases to reduce the number of reactor head safety valves 
required operable from 16 valves to 9 valves. The Nine Mile Point 
Nuclear Station Unit No. 1 (NMP1) reactor vessel was designed to 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME Code), Section I-1962 and Code Case 1271N. In order to show 
compliance with the ASME Code, it was assumed that a main steam 
isolation valve (MSIV) closure occurred without scram. This assumption 
demonstrated that it was necessary to have 16 reactor head safety 
valves. However, the licensee now states that Section 5.2.2.II.A of 
NUREG-0800 (Standard Review Plan) requires that safety valves shall be 
designed with sufficient capacity to limit the pressure to less than 
110 percent of the reactor coolant system design pressure during the 
most severe abnormal operational transient with credit for reactor 
scram. The licensee now proposes to use the high neutron flux scram in 
analyzing this event. The licensee states that this results in a 
reduction in the number of safety valves required operable from 16 
valves to 9 valves. The setpoints of the valve groups would remain 
unchanged. Testing of the safety valves for setpoint and partial lift 
would be changed to be in accordance with the NMP1 Inservice Test 
Program which is based upon ASME Code, Section XI, 1983, including 
Summer Addenda.
    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 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 reduce the number of safety valves from 
sixteen (16) to nine (9). The function of the safety valves is to 
provide code overpressure protection. This design basis event has 
been reanalyzed using the methodology documented in NEDE-24011-P-A, 
General Electric Standard Application for Reactor Fuel (GESTAR). The 
reanalysis takes credit for the high neutron flux scram in order to 
reduce the number of valves. Since peak pressure remains below the 
safety limit, the consequences of the event remain the same. The 
resultant peak pressure is below the pressure safety limit of 1375 
psig. The only event initiator that involves safety valves is the 
spurious actuation of one valve. Since the number of valves has been 
reduced, the probability of a spurious actuation has been reduced. 
Testing in accordance with ASME Section XI will ensure that the 
safety valves lift at the required setpoints although the frequency 
of testing has been reduced. Therefore, 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 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 change represents physical changes to the plant as 
described in the NMP1 Final Safety Analysis Report (Updated). The 
proposed changes however, do not alter the method of providing 
overpressure protection, i.e., safety valves. The valves continue to 
function to limit peak pressure below the safety limit. Therefore, 
maintaining reactor vessel integrity. The reduction in number of 
safety valves results from taking credit for the high neutron flux 
scram in the safety valve actuation transient as allowed by NUREG-
0800. The initiating event, MSIV closure, remains unchanged. 
Although the frequency of testing has been reduced, testing in 
accordance with the requirements of Section XI of the ASME Code will 
ensure valves lift at the required setpoints. Thus, no potential 
initiating events are created which would cause any new or different 
kinds of accidents. Therefore, operation of Nine Mile Point Unit 1 
in accordance with the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The overpressure safety limit of 1375 psig remains unchanged. In 
addition to the initial conditions associated with the safety valve 
actuation transient, NUREG-0800 allows the use of the high neutron 
flux scram. This results in the reduction from sixteen (16) to nine 
(9) safety valves with peak pressure in the vessel still below the 
overpressure safety limit. The margin of safety is defined as the 
range between the safety limit (1375 psig) and the failure point of 
the vessel. Thus, since peak pressure is below the safety limit, the 
margin of safety has not been reduced. Additionally, testing in 
accordance with ASME Section XI ensures operation at the required 
setpoints and does not result in reduction with margin of safety. 
Therefore, the operation of Nine Mile Point Unit 1 in accordance 
with the proposed change will not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: Pao Tsin Kuo

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

    Date of amendment request: July 21, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 2.1.2 (Fuel Cladding Integrity), 
3.1.7 (Fuel Rods), 3.4.6/4.6.2 (Protective Instrumentation), and the 
associated Bases to allow the use of Range 10 on the Intermediate Range 
Neutron Flux Monitors (IRMs) with the Reactor Protection System (RPS) 
low pressure trip for main steam line isolation valve closure not in 
bypass. Changes are also being proposed to TS Tables 3.6.2.a/4.6.2.a 
(Instrumentation that Initiates Scram) and TS Tables 3.6.2.g/4.6.2.g 
(Instrumentation that Initiates Control Rod Withdrawal Block) to extend 
the calibration frequency of the Source Range Neutron Flux Monitors 
(SRMs) and the IRMs from prior to startup and shutdown to once per 
operating cycle. In addition, the proposed amendment would change the 
Instrument Channel Test interval for the SRMs and IRMs from prior to 
startup and shutdown to once per week. The licensee stated that these 
changes are in accordance with NUREG-1433 (Improved Standard Technical 
Specifications for BWR/4). Associated changes to TS Setpoints, Bases, 
References, and Notes for TSs 2.1.2, 3.1.7, and 3.6.2/4.6.2 are also 
being 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 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 expand the IRM operating range, deletes the 
coincident APRM [Average Power Range Monitor] downscale scram trip 
and extend the calibration interval for the SRM/IRM System 
setpoints. The expansion of the startup operating range is required 
to achieve the 1/2 decade overlap between the IRMs and the APRMs. 
Proper overlap improves plant safety by ensuring a smooth transition 
between IRMs and APRMs. The evaluation of operation in IRM range 10 
demonstrates that the addition of range 10 along with the RPS low 
pressure isolation activated ensures that the fuel cladding 
integrity safety limits would not be exceeded.
    The increased IRM/APRM overlap reduces the probability of 
multiple APRM channels downscale in the transition between the IRM 
and APRM Systems and thus eliminates the need for re-activation of 
the IRM scram when in the run mode. The scram is replaced by an 
overlap surveillance which requires that the IRMs overlap by at 
least 1/2 decade with the APRMs during normal shutdown. This 
surveillance ensures that the IRM/APRM overlap is maintained which 
is the basis for deletion of the APRM downscale scram. With the 
improved overlap, the probability of multiple APRM channels being 
downscale is reduced such that it is no longer a credible event and 
therefore, the APRM rod block in combination with proper operating 
procedures, provides the same level of protection. Thus, normal 
plant operation is not affected by these changes and the probability 
of previously analyzed accidents is not increased.
    The new surveillance intervals and setpoints were calculated 
using the General Electric approved methodology documented in NEDC-
31336. The methodology in NEDC-31336 provides assurance that safety 
system actuation (i.e., reactor scram or control rod withdrawal 
block) will occur prior to the associated system parameter, neutron 
flux, from exceeding its analytical limit. Thus, plant response to 
previously analyzed accidents remains within previously determined 
limits.
    Therefore, 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 operation of Nine Mile Point Unit 1, in accordance with the 
proposedamendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The addition of IRM range 10, deletion of the APRM downscale 
scram tripand extension of the surveillance interval for the SRM/IRM 
instrumentation, does not involve an initiation or failure not 
considered in the Final Safety Analysis Report (Updated). The 
proposed changes do not alter the plant configuration and the 
initial conditions used for the design basis accident are still 
valid. Thus, no potential initiating events are created which would 
cause any new or different kinds of accidents. Therefore, 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 previously analyzed.
    The operation of Nine Mile Point Unit 1 in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The addition of IRM range 10 ensures sufficient overlap with the 
APRM System such that switching between startup and run can be 
easily accomplished. The requirement for having the low reactor 
pressure isolation in effect when operating in IRM range 10 is to 
prevent a potential depressurization event. Analysis has shown that 
the margin between the existing safety limits and those events 
previously analyzed has not been reduced. The deletion of the 
coincident APRM scram trip has also been shown not to result in a 
decrease in the margin of safety as the APRM downscale control rod 
withdrawal block provides adequate protection. The analytical limits 
associated with the SRM/IRM instrumentation have been reconstituted 
in conjunction with extending the surveillance interval to once per 
operating cycle. The results using the methodology defined in NEDC-
31336 required that various setpoints in the Technical 
Specifications be changed, however, these changes do not reduce the 
margins between any existing safety limits and previously analyzed 
events. Therefore, operation of Nine Mile Point Unit 1, 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: Pao Tsin Kuo

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone 
Nuclear Power Station, Unit 1, New London County, Connecticut

    Date of amendment request: June 23, 1994
    Description of amendment request: The proposed change would reword 
Technical Specification 3.7, ``Containment Systems,'' to permit 
operation with one of the two circuits of the reactor building 
ventilation logic temporarily inoperable. In addition, Section 
3.7.C.1.b will be reworded to not permit movement of irradiated fuel, 
or movement of any loads over irradiated fuel, without secondary 
containment integrity.
    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:
    NNECO has reviewed the proposed changes in accordance with 
10CFR50.92 and concluded that the changes do not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed changes do not involve a SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The function of a reactor building ventilation isolation is to 
limit fission product release in the event of a design basis 
accident (DBA). Two dampers in series are provided in both the 
supply and exhaust lines of the reactor building ventilation so that 
no single failure would result in the failure to isolate these 
secondary containment penetrations. The proposed change permits one 
of the two circuits of the reactor building ventilation isolation 
logic to be temporarily inoperable.
    The Millstone Unit No. 1 DBAs that could be affected by the 
proposed LCO [Limiting Condition for Operation] are those in which 
the secondary containment is credited to isolate and contain fission 
products. The probability of occurrence of any of these accidents is 
not altered by changes to the operation of secondary containment. 
There is a small potential increase in the consequence of an 
accident previously evaluated. This is due to the small increase in 
probability that a release could occur while Millstone Unit No. 1 is 
operating in the proposed LCO.
    The increase in public risk due to the proposed LCO is 
negligible. The associated increase in risk is similar to the 
increase in public risk permitted by other LCOs. A PRA 
[probabilistic risk assessment] has determined that the probability 
of an event with a radioactive release in the reactor building, 
concurrent with a failure of both operable ventilation dampers to 
function, while in the 7-day LCO, is very small.
    The wording change to Technical Specification 3.7.C.1.b would 
result in more conservative restrictions, in that secondary 
containment integrity would be required when any load (e.g., new 
fuel) is being moved over irradiated fuel. As currently written, the 
Technical Specification only requires secondary containment 
integrity when the fuel cask or irradiated fuel is being moved. 
Therefore, qualitatively, this change would have a positive impact 
on the probability and consequences of accidents involving spent 
fuel.
    Based upon the above, the proposed changes do not constitute a 
significant increase in the probability or consequence of an 
accident previously evaluated.2.
    Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed addition of a new LCO which allows operation with 
one circuit of the reactor building ventilation isolation logic 
inoperable for up to seven days, would only affect the reliability 
of the secondary containment isolation. No new equipment is being 
added, and no new type of operation is being introduced. This change 
allows a short (seven day) period of operation when the reactor 
building ventilation is not single failure proof. The reactor 
building ventilation functions to mitigate the consequences of 
accidents. Failure to function, therefore, does not create the 
possibility of a different kind of accident.
    The proposed change to Section 3.7.C.1.b increases the 
restrictions on load handling, thereby decreasing the possibility of 
any kind of accident involving irradiated fuel.
    Therefore, the proposed changes to the Technical Specifications 
do 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 proposed changes would permit temporary plant operation with 
a small decrease in the reliability of secondary containment 
isolation. However, the reliability of the reactor building 
ventilation isolation would remain high enough with the proposed 
LCO, that the impact on the protective boundaries and the margin of 
safety would be insignificant.
    The proposed change to Section 3.7.C.1.b increases the 
restrictions on load handling, decreasing the possibility of any 
kind of accident involving irradiated fuel. Therefore this change 
would not constitute a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 22, 1994
    Description of amendment request: The proposed amendment would 
revise the technical specifications to (1) change the title of Figure 
3.1-5 to be consistent with the applicable Limiting Condition For 
Operation (LCO), (2) relocate the Chemical and Volume Control System 
(CVCS) valve position requirements to the Reactivity Control Systems - 
Shutdown Margin specifications, and (3) consolidate action statements 
to be expressed in the LCOs rather than in Surveillance Requirements, 
and also clarify the requirements for calculating the heat flux hot 
channel factor FQ(z) when using the base load option.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ... The proposed changes would not involve an SHC [significant 
hazards consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are clarifications or relocation of 
existing technical specification requirements and do not 
substantively affect plant operation. Since they do not affect plant 
operations, they cannot be initiators of any events.
    The safety analysis of the plant is unaffected by the proposed 
changes. Since the safety analysis is unaffected, the calculated 
radiological releases associated with the accident analyses are not 
affected. Therefore, the proposed changes will not increase the 
probability or consequences of previously evaluated accidents.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    As previously stated, the proposed changes are clarifications or 
relocation of existing technical specifications and do not 
substantively affect plant operation. No new failure modes are 
introduced. Since the proposed modifications do not affect plant 
operations, they cannot be initiators of new events.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes are clarifications or relocation of 
existing technical specifications and are not substantive changes. 
The correction of the title in Figure 3.1-5 will ensure consistency 
throughout the technical specifications. The relocation of the CVCS 
valves requirements from the RCS [Reactor Coolant System] - Cold 
Shutdown Specification to the Reactivity Control Systems - Shutdown 
Margin specification will ensure the CVCS valves requirements are 
located in the most appropriate location and will help the operators 
from the commission of errors or omission of actions due to 
inappropriately located material. The final change will revise the 
action statement sections of the specification pertaining to heat 
flux hot channel factor to ensure all actions in these 
specifications are clearly displayed and not contained in the 
corresponding surveillance requirements. Therefore, since these 
changes are editorial in nature, the proposed modification will have 
no impact on the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, 
Connecticut 06141-0270.
    NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket No. 50-171, Peach Bottom 
Atomic Power Station, Unit 1, York County, Pennsylvania
    Date of Application for Amendment: May 9, 1994
    Brief description of amendment: This Licensee Amendment Request 
(LAR) proposes to revise the Peach Bottom Atomic Power Station, Unit 1, 
Possession-Only License and Technical Specifications (TS) to reflect 
the name change of Philadelphia Electric Company to PECO Energy 
Company, to provide proper reference to 10 CFR Part 20 requirements (56 
FR 23360), and to reduce the required frequency for performing periodic 
inspections in the containment vessel below ground level for water 
accumulation. 
    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. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed changes do not alter the operation of equipment 
assumed to be an initiator or any analyzed event or assumed to be 
available for the mitigation of accidents or transients. Proposed 
changes 1 and 2 are administrative in nature. Proposed change 3 to 
reduce the required frequency for performing the periodic inspection 
for water accumulation in the containment vessel below ground level 
does not impact the probability of ground water intrusion into the 
containment building. Proposed change 3 maintains adequate assurance 
that integrity of the containment building with respect to ground 
water entry will be maintained. The design of Unit 1 makes it very 
difficult for ground water to reach the exterior of the containment 
liner to start the metal corrosion process. The concrete layer 
between the rock and the containment liner serves as a barrier to 
prevent water migration to the liner shell. A cathodic protection 
system provides protective current to the containment liner as well 
as nearby underground piping. The steel containment liner of Unit 1 
should not corrode under the present environmental conditions or any 
anticipated future conditions even without an operating cathodic 
protection system. Monthly inspections from May 1990 (following 
issuance of Amendment No. 7 to the Possession-Only License No. DR-12 
on April 25, 1990) through April 1994 have not detected any water in 
the containment building. Prior to Amendment No. 7, the inspection 
of Unit 1 was performed semi-annually. A review of these semi-annual 
inspections dating back to October 1981 determined that water has 
never been detected in the accessible areas below ground level in 
the containment building. The TS limit water accumulation in the 
containment sump to 500 gallons. Twelve and one-half years of 
inspections have confirmed the reliability of the design of Unit 1 
to maintain integrity against any ground water intrusion. There is 
no reason, based on the review of inspection data, why the 
inspection could not be performed semi-annually rather than monthly. 
Therefore, these proposed changes do not increase the probability or 
consequences of an accident previously evaluated.
    b. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because implementation of the proposed changes do not involve any 
physical changes to plant systems, structures, or components. The 
proposed changes do not affect the plant SAFSTOR status. Therefore, 
the possibility of a new or different kind of accident from any 
accident previously evaluated is not created.
    c. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The proposed changes do not involve a significant reduction in a 
margin of safety because the proposed changes do not affect the 
plant SAFSTOR status. Because proposed changes 1 and 2 are 
administrative in nature, they do not involve a question of safety. 
The semi-annual inspection of the accessible areas below ground 
level in the containment building for water accumulation, as 
proposed by change 3, is adequate to ensure containment building 
integrity is maintained with respect to ground water. 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 requests involve no significant hazards consideration.
    Local Public Document Room location:  Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101
    NRC Branch Chief: John H. Austin

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: August 4, 1994
    Description of amendment request: The proposed changes would revise 
Sections 3.4 and 3.5 of the Technical Specifications. The Section 3.4 
revision would reduce the maximum allowable percent of rated power 
associated with inoperable Main Steam Safety Valves (MSSVs). This 
change would modify Table 3.4-1 and the associated Basis such that the 
maximum power level allowed for operation with inoperable MSSVs is 
below the heat removing capability of the operable MSSVs. The Section 
3.5 revision would correct administrative errors in the action 
statements associated with Items 2.a and 2.c of Table 3.5-4. 
Additionally, the proposed changes to Item 2.b of Table 3.5-3 and Item 
2.b of Table 3.5-4 would clarify the action statements associated with 
inoperable high containment pressure (Hi-Hi Level) instrumentation.
    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 evaluated?
    Response:
    The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. This proposed technical specification change 
would modify Table 3.4-1 and the associated basis such that the 
maximum power level allowed for operation with inoperable MSSVs is 
below the heat removing capability of the operable MSSVs. This 
proposed technical specification change will be more conservative 
than the current technical specifications. Proposed changes to Items 
2.a and 2.c of Table 3.5-4 would restore the original intent of the 
specifications and remove undue restrictions on the plant. Proposed 
changes to Item 2.b of Table 3.5-3 and Item 2.b of Table 3.5-4 
clarify the action statements associated with inoperable high 
containment pressure (Hi-Hi Level) instrumentation.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed change incorporates more conservative limits 
on the maximum power level allowed for operation with inoperable 
MSSVs, restores the original intent of items 2.a and 2.c of Table 
3.5-4, and clarifies action statements associated with item 2.b of 
Table 3.5-3 and item 2.b of Table 3.5-4.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed amendment would not involve a significant reduction 
in a margin of safety. This proposed technical specification change 
would modify Table 3.4-1 and the associated basis such that the 
maximum power level allowed for operation with inoperable MSSVs is 
below the heat removing capability of the operable MSSVs. This 
proposed technical specification change will be more conservative 
than the current technical specifications. Proposed changes to Items 
2.a and 2.c of Table 3.5-4 would restore the original intent of the 
specifications and remove undue restrictions on the plant. Proposed 
changes to Item 2.b of Table 3.5-3 and Item 2.b of Table 3.5-3 and 
Item 2.b of Table 3.5-4 clarify the action statements associated 
with inoperable high containment pressure (Hi-Hi Level) 
instrumentation.
    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. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Pao Tsin Kuo
    Power Authority of The State of New York, Docket No. 50-286, 
Indian PointNuclear Generating Unit No. 3, Westchester County, New 
York
    Date of amendment request: August 4, 1994
    Description of amendment request: The proposed changes would revise 
the fuel oil availability requirements for the Emergency Diesel 
Generators (EDGs) from Section 3.7 of the Technical Specifications 
(TSs). This TS change would require that 30,026 gallons of fuel oil be 
available onsite in addition to the oil in the EDG storage tanks. TS 
3.7.F.4 is also being changed to require a total of 7056 gallons of 
fuel in the EDG fuel oil storage tanks. In addition, several 
administrative changes are being proposed to remove the word 
``available'' from the phrase ''... gallons of fuel available...'' in 
Section 3.7.A.5 (for the individual storage tanks) to avoid confusion 
regarding the amount of usable fuel in the tanks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below: 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 do not involve a significant increase in 
the probability or consequences of an accident previously analyzed. 
The change in the minimum required volume for the EDG fuel oil 
storage tanks ensures that two EDGs can power minimum safeguards 
equipment for 48 hours. The new required levels allow for 
temperature effects on fuel density and calibration uncertainties. 
The change to the minimum amount of fuel that must be stored onsite 
is based on a new fuel consumption profile and ensures that 
sufficient oil is present, even in the unlikely event that one EDG 
storage tank (and its associated day tank) is unavailable. The 
change to specification 3.7.F.4 is consistent with the newly 
calculated amount of usable fuel and instrument uncertainties.
    The deletion of the word ``available'' from Section 3.7.A.5 
(concerning the individual storage tanks) and the change to 
Reference 2 of Section 3.7 are administrative in nature and do not 
involve a significant increase in the probability or consequences of 
a previously analyzed accident.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the changes do not affect current plant configuration or how 
the plant operates. The proposed change in the minimum required 
volume for the EDG fuel oil storage tanks ensures an adequate amount 
of usable fuel and allows for temperature effects on fuel density 
and calibration uncertainties. The change to the minimum amount of 
fuel that must be stored onsite is based on a new fuel consumption 
profile and ensures that sufficient oil is present, even in the 
unlikely event that one EDG storage tank (and its associated day 
tank) is unavailable. These changes do not alter how the fuel 
storage tanks operate and therefore do not create the possibility of 
a new or different kind of accident. Specification 3.7.F.4 is being 
changed consistent with the revised calculation.
    The deletion of the word ``available'' from Section 3.7.A.5 
(concerning the individual storage tanks) and the change to 
Reference 2 of Section 3.7 are administrative in nature and do not 
create the possibility of a new or different kind of accident.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed change in the minimum required volume 
for the EDG fuel oil storage tanks ensures the required amount of 
usable fuel is available for two EDGs to operate minimum safeguards 
for 48 hours, and it allows for temperature effects on fuel density 
and calibration uncertainties. The change to the minimum amount of 
fuel that must be stored onsite is based on a new fuel consumption 
profile and ensures that sufficient oil is present, even in the 
unlikely event that one EDG storage tank (and its associated day 
tank) is unavailable. Specification 3.7.F.4 is being changed 
consistent with the revised calculation.
    The deletion of the word ``available'' from Section 3.7.A.5 
(concerning the individual storage tanks) and the change to 
Reference 2 of Section 3.7 are administrative in nature and 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Pao Tsin Kuo
    Power Authority of the State of New York, Docket No. 50-333, 
James A. FitzPatrick Nuclear Power Plant, Oswego County, New York
    Date of amendment request: August 4, 1994
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TSs) to revise the primary 
containment atmosphere monitoring and drywell to torus differential 
pressure requirements. Specifically, TS 3.7.A.6 would be revised to 
adopt primary containment inerting/deinerting requirements that are 
consistent with NUREG-1433, ``Standard Technical Specifications - 
General Electric Plants, BWR/4.'' TSs 4.7.A.6.a and 4.7.A.7.a would be 
revised to provide frequencies for the verifications of primary 
containment oxygen concentration and pressure differential between the 
drywell and torus. TSs 3.7.A.7.a.(1) and 3.7.A.7.a.(3) would be revised 
to provide requirements for establishing and maintaining differential 
pressure between the drywell and torus that are consistent with NUREG-
1433. Several administrative changes to Tables 3.2-8 and 4.3-8 were 
also proposed to improve the overall quality of the TSs.
    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:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes revise primary containment atmosphere 
monitoring requirements. The proposed changes adopt reference plant 
operating conditions (i.e., 15% rated thermal power) for inerting/
de-inerting requirement as well as for the drywell to torus 
differential pressure monitoring consistent with the NRC guidance 
provided in the Standard Technical Specifications. The FitzPatrick 
Technical Specifications currently allow a 24 [hour] grace period 
following startup or before shutdown in which the primary 
containment does not have to be inerted. During this 24 hour time 
period required leak inspections as well as inerting or shutdown 
evolutions are completed. Making the 24 hour ``window'' contingent 
upon core thermal power will allow [operators] to place the mode 
switch in run sooner, removing startup neutron monitoring 
instrumentation scrams (i.e., APRM 15% and IRM upscale/inop). This 
reduces the probability of spurious trips due to spiking of this 
instrumentation. The proposed changes do not involve physical 
modification to the plant nor involve any accident initiators. 
Therefore, the probability of an accident occurring remains 
unchanged. Accident analyses contained in FSAR [Final Safety 
Analysis Report] Chapter 14 assume that a LOCA [Loss-of-coolant 
accident] occurs from full power. The consequences of a LOCA below 
15% rated thermal power would be less severe and would produce less 
hydrogen.
    The proposed changes to Tables 3.2-8 and 4.2-8 will eliminate 
the reference to Specifications 3.7.A.9 by moving the primary 
containment atmosphere monitoring requirements from Specification 
3.7.A.9 to Table 3.2-8, Note F. Note F is also revised such that if 
recorder 279CR-101A or B is inoperable, a daily monitoring and 
logging of the appropriate parameter on the associated indicator on 
panel 279CX-101A, B is acceptable in lieu of taking grab samples. 
The monitoring will be performed using indicators on 279CX-101A and 
B which are Regulatory Guide 1.97 qualified analyzers. The proposed 
new Note K is added for completeness. These changes are 
administrative in nature and will improve the overall quality of the 
technical specifications. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes revise primary containment atmosphere 
monitoring requirements by adopting STS [Standard Technical 
Specifications] guidance regarding inerting/de-inerting 
requirements. Consistent with this change the drywell to torus 
differential pressure monitoring requirement is being revised. 
Adopting the STS reference plant operating condition of 15% rated 
thermal power adds operational flexibility. The proposed changes do 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated because the plant safety 
analyses assume that a LOCA occurs at full power. In addition, 
several changes are proposed to Tables 3.2-8 and 4.2-8 which 
simplify hydrogen/oxygen monitoring requirements by moving the 
primary containment monitoring requirements from Specification 
3.7.A.9 to Table 3.2-8. These changes are administrative in nature 
and will result in the overall improvement to the Technical 
Specifications. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. involve a significant reduction in a margin of safety.
    The proposed changes revise the primary containment atmosphere 
inerting/de-inerting requirements and the drywell to torus 
differential pressure monitoring requirement. The proposed change 
will allow inerting within 24 hours of exceeding 15% rated thermal 
power during startup and de-inerting 24 hours prior to reducing 
thermal power to less than 15% of rated before a plant shutdown. 
These requirements are consistent with the guidance provided in the 
STS. This proposed change does not affect the assumptions or 
conclusions contained in the plant safety analyses which assume that 
a LOCA occurs from full power. The consequences of a LOCA below 15% 
rated thermal power would be less severe and would produce less 
hydrogen. The proposed changes to Tables 3.2-8 and 4.3-8 are 
administrative in nature. 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: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Pao Tsin Kuo

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

    Date of amendment request: July 27, 1994
    Description of amendment request: The amendment request proposes to 
revise the Allowed Out-of-service Times (AOTs) for: inoperable Station 
Service Water System (SSWS) pumps, inoperable Safety Auxiliaries 
Cooling System (SACS) pumps, and inoperable Emergency Diesel Generators 
(EDGs). In addition, this request is also proposing to allow online 
maintenance of the EDGs.
    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:
    PSE&G has, pursuant to 10 CFR 50.92, reviewed the proposed 
amendment to determine whether our request involves a significant 
hazards consideration. We have determined that operation of the Hope 
Creek Generating Station in accordance with the proposed changes:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    LCR 94-08
    Station Service Water System (SSWS) Changes
    Engineering evaluations of the SSWS/Safety Auxiliaries Cooling 
System (SACS) demonstrate that adequate heat removal capability is 
maintained in the post LOCA/LOP period with either two SSWS/SACS 
pumps in one loop or with one SSWS/SACS pump in each independent 
loop. The risk evaluations contained in the Probabilistic Safety 
Assessment analyses of the SSWS determined that the probability of 
an accident previously evaluated does not significantly change by 
increasing the SSWS pump AOT from 7 days to 30 days. The evaluations 
demonstrated that the relative risk remained low with an increased 
(and more appropriate) AOT due to capabilities of the Hope Creek 
SSWS to accommodate active failures.
    Increasing the SSWS pump AOT does not involve physical 
alteration of any plant equipment and does not affect analysis 
assumptions regarding functioning of required equipment designed to 
mitigate the consequences of accidents. Further, the severity of 
postulated accidents and resulting radiological effluent releases 
will not be affected by the increased AOT.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Safety Auxiliaries Cooling System Changes
    Engineering evaluations of the SSWS/SACS demonstrate that 
adequate heat removal capability is maintained in the post LOCA/LOP 
period with either two SSWS/SACS pumps in one loop or with one SSWS/
SACS pump in each independent loop. The risk evaluations contained 
in the Probabilistic Safety Assessment analysis of the SACS 
determined that the probability of an accident previously evaluated 
does not significantly change by increasing the SACS pump AOT from 
72 hours to 30 days. Similarly, the provision of a 72 hour AOT for 
one SACS pump inoperable in each SACS loop does not significantly 
change the probability of an accident previously evaluated. The 
evaluations demonstrated that the relative risk remained low with an 
increased (and more appropriate) AOTs due to capabilities of the 
Hope Creek SACS to accommodate active failures.
    Increasing the SACS pump AOTs does not involve physical 
alteration of the plant equipment and does not affect analysis 
assumptions regarding functioning of required equipment designed to 
mitigate the consequences of accidents. Further, the severity of 
postulated accidents and resulting radiological effluent releases 
will not be affected by the increased AOTs.
    The proposed changes to ACTION Statement a.2 of Technical 
Specification 3.7.1.1 precludes overly conservative and improper 
operator action (initiation of plant shutdown procedures) to comply 
with the requirements in the situation in which one of the affected 
EDGs (an EDG cooled by the inoperable SACS loop) is not realigned to 
OPERABLE SACS loop. Currently, Hope Creek can simultaneously be in 
the ACTION Statement for Technical Specifications 3.7.1.1 and 
3.8.1.1. Simultaneous entry into these ACTION Statements bounds the 
conditions of the plant when the proposed requirements of the 
Technical Specification 3.7.1.1, ACTION Statement a.2 are met. For 
this reason, the proposed changes will not increase the 
probabilities or consequences of an accident previously evaluated.
    Technical Specification 3.7.1.1, ACTION Statements b., c. and d. 
are being revised to require that the RHR loop or safety related 
equipment must be declared inoperable when two SACS pumps in the 
associated SACS loop are inoperable. This change permits one SACS 
pump to be inoperable without affecting the operability of the 
associated RHR loop or safety related equipment. Engineering 
evaluations demonstrate that two SACS loops with one pump and two 
heat exchangers per loop can provide the required heat removal 
capability in the post DBA LOCA/LOP scenario and maintain safe 
shutdown conditions. Therefore, a SACS loop with one OPERABLE SACS 
pump should still be considered as a 100% functional SACS loop, 
capable of supplying sufficient cooling for RHR and safety related 
equipment required by Specifications 3.4.9.1, 3.4.9.2, 3.5.2, 
3.9.11.1 and 3.9.11.2. For this reason, the proposed changes will 
not increase the probabilities or consequences of an accident 
previously evaluated.
    In conclusion, the above SACS changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    LCR 94-11
    Emergency Diesel Generator AOT Extensions.
    The Hope Creek offsite and onsite power systems are highly 
reliable. The risk evaluations contained in the Probabilistic Safety 
Assessment analyses of the onsite power system determined that the 
probability of an accident previously evaluated does not 
significantly change by increasing the diesel generator AOT from 72 
hours to 30 days for one inoperable diesel generator or from 2 hours 
to 72 hours for two inoperable diesel generators. The evaluations 
demonstrated that the relative risk remained low with an increased 
(and more appropriate) AOT due to capabilities of the four channel 
onsite Class-1E electrical system design at Hope Creek.
    Increasing the diesel generator AOT does not involve physical 
alteration of any plant equipment and does not affect analysis 
assumptions regarding functioning of required equipment designed to 
mitigate the consequences of accidents. Further, the severity of 
postulated accidents and resulting radiological effluent release 
will not be affected by the increased AOT.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    LCR 94-12
    Emergency Diesel Generator Online Maintenance
    The proposed changes would require that the requisite number of 
diesel generators be in an operable condition, but would eliminate 
the restriction that the 18 month maintenance inspection and other 
surveillance tests be performed only while the unit is shutdown. 
Because all operational conditions (governed by the operability of 
the equipment prescribed as necessary in Technical Specification 
3.8.1.1) and the associated actions are defined elsewhere in the 
Technical Specifications, the removal of this restriction would not 
involve as significant increase in the probability or consequences 
of an accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    LCR 94-08
    Station Service Water System (SSWS) Changes
    Extending the SSWS pump AOTS does not necessitate physical 
alteration of the plant or changes in parameters governing normal 
plant operation. Thus, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated for Hope Creek.
    Safety Auxiliaries Cooling System Changes
    The changes to the SACS do not necessitate physical alteration 
of the plant or changes in parameters governing normal plant 
operation. Thus, these changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated for Hope Creek.
    LCR 94-11
    Emergency Diesel Generator AOT Extensions
    Extending the diesel generator AOTs does not necessitate 
physical alteration of the plant or changes in parameters governing 
normal plant operation. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated for Hope Creek.
    LCR 94-12
    Emergency Diesel Generator Online Maintenance
    The proposed revisions will not change the method in which any 
of the 4.8.1.1.2.h surveillance activities are to be performed, only 
the prescriptive operational condition is being removed. Since the 
operational conditions and the associated actions are defined 
elsewhere in the Technical Specifications, the removal of this 
restriction will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    LCR 94-08
    Station Service Water System (SSWS) Changes
    As discussed above, the Probabilistic Safety Assessment analyses 
determined that the change in core damage frequency for extended 
SSWS pump AOT is insignificant. Therefore, this change does not 
result in a significant reduction in a margin of safety.
    Safety Auxiliaries Cooling System Changes
    As discussed above, the Probabilistic Safety Assessment analyses 
determined that the change in core damage frequency for the SACS 
changes are insignificant. Therefore, these changes do not result in 
a significant reduction in a margin of safety.
    LCR 94-11
    Emergency Diesel Generator AOT Extensions
    As discussed above, the Probabilistic Safety Assessment analyses 
determined that the change in core damage frequency for extended 
diesel generator AOTs is insignificant. Therefore, this change does 
not result in a significant reduction in a margin of safety.
    LCR 94-12
    Emergency Diesel Generator Online Maintenance
    The margin of safety for the emergency power system depends on 
the proven, historical reliability of the diesel generators and the 
surveillances verifying the power circuits between the offsite and 
the onsite power systems. The elimination of the restrictions for 
performance of the maintenance tear down inspection would remain 
within the action parameters of Technical Specification 3.8.1.1. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Mohan C. Thadani, Acting

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

    Date of amendment request: July 28, 1994
    Description of amendment request: The proposed Technical 
Specification changes contained herein represent changes to Section 3/
4.8.1 ``AC Sources.'' The revised specification removes the 
surveillance requirements, methodology and frequency for Emergency 
Diesel Generator (EDG) fuel oil from the Technical Specifications and 
relocates them in a controlled plant procedure, VSH.SS-CA.ZZ-0013(Q) 
``Procedure for Testing Diesel Fuel and 2 Fuel Oil at 
Artificial Island for PSE&G Nuclear Operations.'' The changes also 
delete an unnecessary lab test for the fuel oil and extend the 
surveillance frequency from once per 92 days to once per 184 days.
    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. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to delete a test not required by Regulatory 
Guide 1.137 or ASTM-D975-77 will not result in degradation of fuel 
oil quality below acceptable limits. Based on established fuel oil 
quality history, the proposed increase in surveillance frequency 
from once per 92 days to once per 184 days will not significantly 
decrease confidence in fuel oil quality and EDG operability, nor 
will the relocation of fuel oil quality surveillance from the 
Technical Specifications to the Diesel Fuel Oil Testing Program have 
any effect on established plant practices in regards to the testing 
of EDG fuel oil. The proposed changes involve no hardware changes, 
no changes to the operation of any systems or components, and no 
changes to existing structures. Therefore, these changes will not 
alter or impact previously evaluated accidents.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes are procedural in nature concerning fuel 
oil testing and, therefore, will not directly impact the operation 
of any plant safety related component or equipment. Any reduction in 
fuel oil quality will not be significant or result in a decrease in 
EDG operability. Therefore, these changes will not create a new or 
unevaluated operating condition.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes concern how EDG fuel oil quality is to be 
determined, how frequently this determination is to be performed, 
and how to control the process for determining fuel oil 
acceptability, and therefore EDG operability. There are no 
associated safety margins and the only margin of concern is that of 
fuel oil combustibility due to the presence of either contaminants 
or particulate buildup from long term storage. Based on historical 
data, PSE&G believes that EDG fuel oil quality will not be affected 
or impacted by the proposed changes. Therefore, the proposed 
amendment does not involve any reduction in a safety margin.
    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, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Mohan C. Thadani, Acting

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: May 20, 1994.
    Description of amendment requests: This is a proposal to revise the 
Units 2 and 3 Technical Specification (TS) 3/4.7.3, ``Component Cooling 
Water System,'' and the corresponding Bases to support the addition of 
the component cooling water surge tank backup nitrogen supply (BNS) 
system. The amendment is necessary to establish new operability and 
surveillance requirements for the 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:
    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The Component Cooling Water (CCW) system removes heat to 
mitigate the consequences of those design basis accidents included 
in chapter 15 of the Updated Final Safety Analysis Report (UFSAR). A 
CCW system failure is not an accident initiating event as listed in 
the UFSAR, Table 15.0-2. The addition of the Backup Nitrogen Supply 
(BNS) system does not change the CCW system function and does not 
interface with any system which relates to the initiating events 
listed in Table 15.0-2 of the UFSAR. The BNS system is designed to 
Quality Class II, Seismic Category I requirements and will increase 
CCW reliability by minimizing CCW system voiding during and after a 
Design Basis Event (DBE). Failure of the BNS system will not by 
itself result in an accident or have any effect on normal plant 
operation.
    The proposed revision of Technical Specification (TS) 3/4.7.3 
will not change the CCW system operation. This amendment request 
retains the original CCW TS requirements and adds provisions 
specifically limited to the BNS system. The proposed revisions 
provide an 8-hour Allowed Outage Time (AOT) for one or both trains 
of the BNS system inoperable to avoid unnecessary plant power 
reductions. If the 8-hour AOT for BNS system inoperability is not 
met, the associated CCW train(s) must be declared inoperable. The 8-
hour AOT followed by either the 72-hour AOT for one train of CCW 
inoperable or the 1-hour AOT provided by TS 3.0.3 for both trains of 
CCW inoperable results in overall AOTs of 80 and 9 hours, 
respectively. The results of a conservative Probabilistic Risk 
Assessment demonstrate that for the overall 80-hour and 9-hour AOTs 
the increases in core damage risk per year are 6.5E-7 and 8.6E-7, 
respectively. This results in less than a 3% increase in the annual 
core damage risk for Units 2 and 3.
    The proposed revisions to TS 3/4.7.3 include surveillance 
requirements to provide assurance that the BNS system remains 
OPERABLE when required to support CCW operation. Therefore, 
operation of the facility in accordance with this proposed TS change 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed amendment create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    Response: No
    The BNS system does not change the CCW system function and does 
not interface with any system related to the initiating events 
listed in Table 15.0-2 of the UFSAR. The BNS system is designed to 
Seismic Category I requirements and will minimize CCW system voiding 
and the potential for a subsequent water hammer by maintaining the 
CCW surge tank pressure during and after a DBE. No new High Energy 
Line Break considerations apply because the nitrogen bottle pressure 
is reduced at the bottle header and all connections are less than 
one inch in diameter. The BNS system is independent from all systems 
possibly related to the initiating DBEs listed in the UFSAR Table 
15.0-2.
    The proposed TS 3/4.7.3 revision does not change the existing 
CCW system requirements. This proposed change adds operability and 
surveillance requirements for the BNS system to support CCW system 
operability and provide additional assurance that plant operation is 
consistent with the design basis. Failure of the BNS system will not 
by itself result in an accident or have any effect on normal plant 
operation. Therefore, the proposed changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed amendment involve a significant reduction in a margin of 
safety?
    Response: No
    The addition of the BNS system enhanced the CCW system by 
minimizing the possibility for water hammer following certain 
postulated events. Surveillance and nitrogen bottle change-out 
procedures assure that the BNS system is available to perform its 
safety-related function. The redundant cooling capacity of the CCW 
system is maintained by providing an independent dedicated BNS 
system for each CCW critical loop, assuming a single failure.
    The safety function of the BNS system is limited to the 
minimization of void formation in the CCW system under a specific 
set of coincident circumstances following a DBE. The proposed 
revision to TS 3/4.7.3 allows the BNS system to have one or both 
trains inoperable for 8 hours before the associated CCW train(s) 
must be declared inoperable. The BNS system AOTs do not affect plant 
operation because the BNS system is not normally in operation. The 
BNS system action statements are not normally entered for normal 
bottle change out since the BNS system is designed with one more 
bottle than is required for seven days of BNS system operation. 
Therefore, the proposed changes do 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 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: James A. Beoletto, Esquire, Southern 
California Edison Company, P. O. Box 800, Rosemead, California 91770
    NRC Project Director: Theodore R. Quay

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

    Date of amendment request: May 20, 1994
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications to incorporate improvements 
endorsed by the NRC Final Policy Statement on Technical Specification 
Improvements for Nuclear Power Reactors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed Technical Specification changes involve relocating 
requirements that are not conditions or limitations on reactor 
operation necessary to obviate the possibility of an abnormal 
situation or event giving rise to an immediate threat to the public 
health and safety. The proposed changes were identified through the 
application of criteria designed to cull those requirements that are 
not important to operational safety from the Technical 
Specifications. In this process, selected provisions of the 
Technical Specifications identified for relocation were retained if 
necessary to support a Technical Specification that was to be 
retained. Thus, only specification requirements that have little or 
no operational safety significance are proposed for relocation. In 
addition, those requirements that would be relocated will be 
included in the Final Safety Analysis Report (FSAR) and, therefore, 
will be controlled and implemented as FSAR commitments. In this 
manner, those requirements that have no operational safety 
significance but involve maintaining the plant in its as-designed 
state, (for example, through surveillance programs) would be 
controlled.
    In addition, the criteria for identifying requirements to be 
retained in Technical Specifications specifically call out, for 
retention, those structures, systems, or components that are 
required to mitigate accidents previously evaluated.
    Based on the above, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed changes involve relocating Technical Specification 
requirements to another licensee-controlled document, i.e. FSAR 
Chapter 16. No changes or physical alterations of the plant are 
involved. Also, no changes to the operation of the plant or 
equipment are involved. Therefore, the proposed changes do 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 proposed changes involve relocating Technical Specification 
requirements to the FSAR. The requirements to be relocated were 
identified by applying the criteria endorsed in the Commission's 
Policy Statement. Thus, those specifications that would be relocated 
do not impose constraints on design and operation of the plant that 
are derived from the plant safety analysis report or from 
probabilistic safety assessment (PSA) information and do not belong 
in the Technical Specifications in accordance with 10 CFR 50.36 and 
the purpose of the Technical Specifications stated in the Policy 
Statement. Therefore, relocation of these requirements 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: June 21, 1994
    Description of amendment request: The proposed amendment would 
delete the sections describing the On-Site Review Committee (ORC) and 
the Nuclear Safety Review Board (NSRB) from the Technical 
Specifications. This change also removes reference to the Manager, 
Nuclear Safety and Emergency Preparedness. Additionally, the change 
reflects an organizational restructuring which addresses the 
Independent Safety Engineering Group (ISEG) reporting to the Manager, 
Quality Assurance.
    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 changes are administrative and equivalent descriptions and 
requirements for these oversight committees are contained in FSAR 
Section 13.4.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    These changes do not involve any physical alterations to the 
plant. There is no new type of accident or malfunction created and 
the method and manner of plant operation will not change. The 
changes are administrative and equivalent descriptions and 
requirements for these oversight committees are contained in FSAR 
Section 13.4.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety remains unaffected since no design change 
is made and plant operation remains the same. The changes are 
administrative and equivalent descriptions and requirements for 
these oversight committees are contained in FSAR Section 13.4.
    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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: July 12, 1994
    Description of amendment request: The proposed amendment would 
modify the technical specifications (TS) to remove instrument response 
time limit tables for the reactor protection system (RPS) and isolation 
actuation and emergency core cooling system (ECCS) from the TS. The 
affected instrument response time limit tables would be located in the 
Final Safety Analysis Report (FSAR).
    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 RPS, Isolation Actuation and ECCS Instruments provide 
signals to the actuation logic for safety equipment needed to 
mitigate accidents and transients. The proposed change relocates the 
instrument response times from the Technical Specifications to the 
FSAR but will not affect the operability or surveillance 
requirements of the affected instruments. The instruments will 
continue to be proven operable on the schedule provided in the 
Technical Specifications.
    The FSAR change process and Plant Operations Committee review 
responsibilities ensure that changes to the response time limits 
cannot be made without adequate review and approval. Since 
operability confirmation as required by the Technical Specifications 
(surveillance testing requirements) will not be affected by the 
change and the limits themselves cannot be altered without adequate 
review and approval, there is no possibility of a significant 
increase in the probability of an accident previously approved as a 
result of this change.
    The instruments provide signals to the actuation logic of 
equipment used to mitigate the consequences of an accident. However, 
since no changes are being made in the methods or frequencies of 
proving operability the systems will not be degraded or be made 
susceptible to degradation that could go unidentified. As discussed 
above, changes to the limits will not be made without adequate 
review and approval. Hence, this change will not affect the 
capability of the plant to mitigate a previously evaluated accident. 
Because the mitigative capability is not affected there is no 
significant increase in the consequences of a previously evaluated 
accident as a result of this change.
    For the above reasons, the change does not represent 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 relocates only the tables containing the 
instrument response times for the RPS, Isolation Actuation and ECCS 
response time limits from the Technical Specifications to the FSAR. 
The change does not affect how these instruments will function. 
Relocation of this information does not represent a change in the 
configuration or operation of the plant. No new hardware is being 
added to the plant as part of the proposed change. Plant procedures 
are not affected by the change. The Technical Specification sections 
for the surveillance testing of these instruments will not be 
affected. Therefore, the Technical Specifications will continue to 
require that the same operability and surveillance requirements be 
met for the affected instruments.
    Consequently, the possibility of a new or different kind of 
accident from any accident previously analyzed is not introduced as 
a result of this change.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety established by the response time limits is 
in ensuring that the RPS, Isolation Actuation and ECCS systems will 
respond in time to support the assumptions of the accident analysis. 
Relocating the response time limits to the FSAR does not alter the 
operability or the surveillance requirements applicable to the 
affected instruments. These instruments will continue to be tested 
for operability and therefore remain capable of responding to 
accident events within the time limits required by the accident 
analysis. The administrative change control provisions for the FSAR, 
the plant procedures implementing the requirements of 10 CFR 50.59 
and the administrative sections of the Technical Specifications are 
adequate to control changes to the response time limits such that 
they cannot be altered in a manner that would adversely affect plant 
safety.
    Therefore, for these reasons, the 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 request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: Theodore R. Quay

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.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of amendment request: March 23, 1994, as supplemented on July 
26, 1994
    Description of amendment request: The proposed amendments consist 
of two parts: Part one, would revise ``Moderator Temperature 
Coefficient (MTC)'' Technical Specifications (TSs) to allow the use of 
a slightly positive MTC for the core design. The licensee has stated 
that a positive MTC will reduce the burnable rod requirements and 
improve operational flexibility. Because of using a positive MTC, the 
TSs would be revised to permit a higher boron concentration in the 
refueling water storage tank, the reactor coolant system (RCS) 
accumulators, and the refueling cavity, in order to ensure adequate 
shutdown margin is maintained at all times. Part two, would revise the 
TSs to reduce the required RCS flow to offset any reduction in flow due 
to increased steam generator tube plugging. Additionally, the 
associated Bases for the above TSs would be revised to describe the 
basis for the TS requirements.
    Because Byron, Unit 1, and Braidwood, Unit 2, will be in refueling 
outage in the fall of 1994, the proposed TS changes will apply to them. 
Byron, Unit 2 and Braidwood, Unit 1 will continue to operate in 
accordance with the current TSs. The licensee's submittal identified 
the appropriate unit applicability of the TSs pertaining to the 
positive MTC and the required RCS flows.Date of publication of 
individual notice in Federal Register: August 15, 1994 (59 FR 41802)
    Expiration date of individual notice: September 14, 1994
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, Byron, Illinois 61010; and for Braidwood, the 
Wilmington Township Public Library, 201 S. Kankakee Street, Wilmington, 
Illinois 60481.

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

    Date of amendment request: August 5, 1994
    Brief description of amendment request: The proposed amendment 
would modify Technical Specification Table 4.8.1.1.2-1, ``Diesel 
Generator Test Schedule,'' to exclude selected valid failures of the 
Division 1 diesel generator from contributing to an accelerated testing 
frequency.Date of publication of individual notice in Federal Register: 
August 16, 1994 (59 FR 42080).
    Expiration date of individual notice: September 15, 1994
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.
    Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and 
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606
    NRC Project Director: John N. Hannon

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

    Date of amendment request: January 14, 1994
    Brief description of amendment request: The proposed amendment 
would increase the storage capacity in each spent fuel pool from their 
current 2040 fuel assemblies to 4117 fuel assemblies. In addition, the 
proposed amendment would extend the ``fuel core reserve'' capability 
from year 1998 to 2013.
    Date of publication of individual notice in Federal Register: 
August 8, 1994 (59 FR 40376)
    Expiration date of individual notice: September 7, 1994
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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 20555, and at the local public document 
rooms for the particular facilities involved.

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

    Date of application for amendments: July 1, 1994, as supplemented 
by letter dated August 11, 1994
    Brief description of amendment: The proposed amendment changes the 
minimum cold-leg temperature for core power levels between 90 percent 
and 100 percent to 552 degrees Fahrenheit for Unit 2 (which is a 
reduction of 10 degrees Fahrenheit from the previous technical 
specification (TS) requirement). This TS change permits reactor 
operation at full power with a lower reactor coolant temperature to 
minimize potential steam generator tube degradation. The cold-leg 
temperature reduction at power levels above 90 percent was previously 
granted for Units 1 and 3 by letter dated June 7, 1994.
    Date of issuance: August 12, 1994
    Effective date: August 12, 1994
    Amendment No.: 65
    Facility Operating License No. NPF-51: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 13, 1994 (59 FR 
35767) The additional information contained in the supplemental letter 
was clarifying in nature, was within the scope of the initial notice, 
and did not affect the NRC staff's proposed no significant hazards 
determination.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 12, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

Commonwealth Edison Company, Docket No. STN 50-456, Braidwood 
Station, Unit No. 1, Will County, Illinois

    Date of application for amendment: June 20, 1994, as supplemented 
on August 18, 1994.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 4.4.5.4.a(11) by moving a footnote into the body of 
the text. Additionally, Item 3 of TS Section 4.4.5.4.a(11) has been 
revised to remove the licensee's previous calculation of primary-to-
secondary leakage of 26 gallons per minute (gpm) at the end of 100 
calendar days of operation in the present fuel cycle. In place of this 
value, the licensee's revised calculated value of less than 9.1 gpm at 
the end of Cycle 5 is inserted, including a reference to the basis for 
this revised estimate (i.e; WCAP-14046). Finally, Section 3.4.8.a is 
revised to remove the limit on operating time in the present fuel 
cycle. The maximum permissible dose equivalent Iodine-131 concentration 
in the footnote to Section 3.4.8.a remains at 0.35 microcuries per gram 
of coolant as proposed by the licensee in its letter dated August 18, 
1994. The net result of these revisions is to remove the limitation on 
permissible operating time from the Braidwood 1 TSs.
    Date of issuance: August 18, 1994
    Effective date: August 18, 1994
    Amendment No.: 54
    Facility Operating License No. NPF-72. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 11, 1994 (59 FR 
35389) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 18, 1994. The staff has 
found that its prior determination of no significant hazards 
consideration is not affected by the licensee's submittal of August 18, 
1994.No significant hazards consideration comments received: No
    Local Public Document Room location:  Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

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

    Date of application for amendment: June 28, 1989, as supplemented 
May 1 and September 26, 1991, March 18, August 24, and August 28, 1992, 
May 19, 1993, May 5 and July 7, 1994.
    Brief description of amendment: This amendment adds new operational 
requirements, action statements, and surveillance requirements to 
assure the availability of shutdown cooling to the primary coolant 
system during certain operational conditions.Date of issuance: August 
12, 1994 Effective date: August 12, 1994, with full implementation 
within 90 days
    Amendment No.: 161
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 7, 1990 (55 FR 
8221) and August 18, 1993 (58 FR 43924). The May 5 and July 7, 1994 
letters provided clarifying information within the scope of the August 
18, 1993, notice and did not affect the staff's proposed no significant 
hazards consideration findings. The Commission's related evaluation of 
the amendment is contained in a Safety Evaluation dated August 12, 
1994.No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendments: October 22, 1993
    Brief description of amendments: These amendments revise the 
Appendix A TSs relating to surveillance test intervals and allowed 
outage time for the analog instrumentation channels of the reactor trip 
system and the engineered safety feature actuation system.
    Date of issuance: August 8, 1994
    Effective date: August 8, 1994
    Amendment Nos.: 181 and 61
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34660) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 8, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear 
One,Unit No. 1, Pope County, Arkansas
    Date of amendment request: March 3, 1994
    Brief description of amendment: The amendment removed restrictions 
from the Arkansas Nuclear One, Unit No. 1 technical specifications that 
prohibit use of the auxiliary building crane to move spent fuel 
shipping casks.
    Date of issuance: August 4, 1994
    Effective date: August 4, 1994
    Amendment No.: 173
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17598) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 4, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russelville, Arkansas 72801

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket No. 
50-366, Edwin I. Hatch Nuclear Plant, Unit 2, Appling County, 
Georgia

    Date of Application for Amendment: July 19, 1994, as Supplemented 
August 4, 1994.
    Brief description of amendment: The amendment revised Technical 
Specification 3.3.6.6, ``Traversing Incore Probe System,'' for Hatch 
Unit 2 to permit the traversing incore probe (TIP) system to be 
considered operable with less than four operable TIP units. Date of 
issuance: August 8, 1994
    Effective date: August 8, 1994
    Amendment No.: 134 (Unit 2)
    Facility Operating License No. NPF-5: The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes. (59 FR 37516 dated July 22, 
1994). The notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice also provided 
for an opportunity to request a hearing by August 22, 1994, but 
indicated that if the Commission makes a final no significant hazards 
consideration determination, any such hearing would take place after 
issuance of the amendment. The August 4, 1994, letter provided 
additional information that did not change the scope of the July 19, 
1994, application and initial proposed no significant hazards 
consideration determinations.The Commission's related evaluation of the 
amendment, finding of unusual circumstances, and a final no significant 
hazards consideration determination are contained in a Safety 
Evaluation dated August 8, 1994.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

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

    Date of application for amendment: May 6, 1994
    Brief description of amendment: The amendment changes the monthly 
operational test of the reactor trip bypass breakers from monthly to 
monthly staggered, such that each breaker is tested every 62 days. 
Also, it changes the word Breakers in the Functional Unit title to 
Breaker.
    Date of issuance: August 12, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 93
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32233) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 12, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Community-Technical College, Thames Valley Campus, 574 New London 
Turnpike, Norwich, Connecticut 06360.

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: June 24, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications to change the Administrative Controls section to require 
an individual who serves as the Operations Manager to either hold a 
Millstone Unit 2 Senior Reactor Operator (SRO) license or have an SRO 
license at another pressurized water reactor. If the Operations Manager 
does not hold a Millstone Unit 2 SRO license, then an individual 
serving as the Assistant Operations Manager would be required to 
possess an SRO license at Millstone Unit 2.
    Date of issuance: August 11, 1994
    Effective date: As of the date of issuance.
    Amendment No.: 178
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (59 FR 34872, July 7, 1994). 
That notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice also provided 
for an opportunity to request a hearing by August 8, 1994, but 
indicated that if the Commission makes a final no significant hazards 
consideration determination any such hearing would take place after 
issuance of the amendment. The Commission's related evaluation of the 
amendment, finding of exigent circumstances, and final determination of 
no significant hazards consideration are contained in a Safety 
Evaluation dated August 11, 1994.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Northern States Power Company, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County, Minnesota

    Date of application for amendments: September 21, 1992, as revised 
December 29, 1992, November 24, 1993, May 17, 1994, and June 21, 1994.
    Brief description of amendments: The amendments revise Technical 
Specifications and associated Bases for surveillance test intervals and 
allowed outage times for the engineered safety features and reactor 
protection system instrumentation consistent with the NRC staff 
position as documented in NRC letters to the Westinghouse Owners Group. 
The amendments also update operation modes to be consistent with 
Westinghouse Standard Technical Specification operational modes and 
also include several editorial changes to the Prairie Island Technical 
Specifications that are unrelated to the changes described above.
    Date of issuance: August 10, 1994
    Effective date: August 10, 1994
    Amendment Nos.: 111 & 104
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10012). The May 17, 1994, and June 21, 1994, letters provided 
clarifying information within the scope of the March 2, 1994, notice. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated August 10, 1994.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Minneapolis Public 
Library,Technology and Science Department, 300 Nicollet Mall, 
Minneapolis, Minnesota 55401.

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

    Date of amendment request: December 28, 1993
    Brief description of amendment: The amendment to the technical 
specifications revised the surveillance test frequency from monthly to 
quarterly for several channel functional tests for reactor protective 
system and engineered safety feature instrumentation and controls based 
on Generic Letter 93-05.
    Date of issuance: August 17, 1994
    Effective date: August 17, 1994
    Amendment No.: 163
    Facility Operating License No. DPR-40. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10013) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 17, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of application for amendments: March 24, 1994
    Brief description of amendments: The amendments revise Technical 
Specification Sections 3.11.1.4, 6.9.1.8, and 6.14.1, and TS Definition 
1.24 to change the frequency for submitting the Semiannual Radioactive 
Effluent Release Report to the NRC from semiannually to annually.
    Date of issuance: August 10, 1994
    Effective date: August 10, 1994
    Amendment Nos. 73 and 35
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24751) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 10, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Public Service Electric and Gas 
Company, Delmarva Power and Light Company, and Atlantic City 
Electric Company, Docket No. 50-277, Peach Bottom Atomic Power 
Station, Unit No. 2, York County, Pennsylvania

    Date of application for amendment: April 1, 1993, as supplemented 
by letters dated April 7, July 16, and August 20, 1993, and June 8, 
1994
    Brief description of amendment: These amendments implement an 
expanded power-to-flow operating domain supported by the Average Power 
Range Monitor, Rod Block Monitor, Technical Specifications Improvement/
Maximum Extended Load Line Limit Analysis (ARTS/MELLLA) (NEDC-32162P, 
Revision 1, February 1993) submitted with the licensee's April 1, 1993, 
application.
    Date of issuance: August 10, 1994
    Effective date: Following startup from Refueling Outage 2R10.
    Amendment No.: 192
    Facility Operating License No. DPR-44: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (59 FR 
39058) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 10, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

Philadelphia Electric Company, Public Service Electric and Gas 
Company Delmarva Power and Light Company, and Atlantic City 
Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: April 27, 1994
    Brief description of amendments: The amendments modify the existing 
Limiting Conditions for Operation, surveillance requirements and bases 
to reflect new containment monitoring system hydrogen/oxygen analyzers. 
The new analyzers are to be installed in Unit 2 during the scheduled 
September 1994 refueling outage and will support the Containment 
Atmospheric Dilution system and the Containment Atmospheric Control 
system.
    Date of issuance: August 10, 1994
    Effective date: Prior to the startup of Unit 2 following refueling 
outage 2R10.
    Amendments Nos.: 193 and 197
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29629) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 10, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of application for amendment: June 25, 1993
    Brief description of amendment: The amendment revised the Plant 
Operations Review Committee (PORC) composition and quorum description, 
presented the membership composition through a set of requirements 
defining the necessary management titles to functional titles, changed 
the term ``designated alternate'' to ``designee,'' and removed the 
requirements in Specification 6.5.2.5 to have the Nuclear Safety Review 
Committee meetings ``at least once per calendar quarter during the 
initial year of operation following fuel loadings and...thereafter.''
    Date of issuance: August 17, 1994
    Effective date: Date of issuance, to be implemented within 90 days
    Amendment No. 65
    Facility Operating License No. NPF-58. This amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: September 15, 1993 (58 
FR 48390) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 17, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa 
County, Virginia

    Date of application for amendments: November 19, 1993
    Brief description of amendments: The amendments revise the NA-1&2 
TS to allow the substitution of solid stainless steel or zirconium 
alloy filler rods for a limited number of failed fuel rods in fuel 
assemblies. This will allow the use of reconstituted fuel assemblies, 
which were scheduled for reload, without requiring reload core design 
and selection of a replacement assembly during a refueling outage.
    Date of issuance: August 9, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 186 and 167
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67863) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 9, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of application for amendments: February 26, 1993, as 
supplemented on March 9, 1993.
    Brief description of amendments: The amendments revised Technical 
Specification Section 15.3.1.A.3, ``Limiting Conditions for Operation, 
Components Required for Redundant Decay Heat Removal Capability.'' The 
amendments clarified the exception for when one decay heat removal 
method must be in operation. In addition, the amendments changed the 
applicable Basis (page 15.3.1-3c) to improve the clarity and 
consistency of this section. Date of issuance: August 16, 1994
    Effective date: Immediately, to be implement within 20 days
    Amendment Nos.: 149 and 153
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43939) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 16, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: May 24, 1994
    Brief description of amendment: The amendment relocates the TS 
requirements related to seismic monitoring instrumentation from the TS 
to the Updated Safety Analysis Report (USAR). The requirements of these 
TS will be maintained and controlled pursuant to Appendix A to 10 CFR 
100 and other applicable regulations, including 10 CFR 50.59, 
``Changes, tests, and experiments.''
    Date of issuance: August 11, 1994
    Effective date: August 11, 1994, to be implemented within 120 days 
of issuance.
    Amendment No.: 75
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 6, 1994 (59 FR 
34671) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 11, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: October 27, 1993
    Brief description of amendment: The amendment revises Technical 
Specification 4.6.1.2.a, Overall Integrated Containment Leakage Rate, 
to provide one-time relief from the requirements to perform the 
surveillance at intervals of 40 months plus or minus 10 months. The 
schedule for the third Type A test is extended to the eighth refueling 
outage, approximately 54 months after the second test, in order to have 
it coincide with the 10-year inservice inspections.
    Date of issuance: August 12, 1994
    Effective date: August 12, 1994, to be implemented within 30 days 
days of issuance.
    Amendment No.: 76
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64616) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 12, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: February 24, 1994
    Brief description of amendment: The amendment revises Technical 
Specification 4.7.1.2.1.a to require that the turbine-driven and motor-
driven auxiliary feedwater pumps be tested at least quarterly on a 
staggered basis instead of the previously required testing once per 31 
days on a staggered basis. The revised surveillance frequency is 
consistent with the guidance issued in Generic Letter 93-05, ``Line-
Item Technical Specification Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation.'' The Bases to TS 3/
4.7.7, Emergency Exhaust System - Auxiliary Building, and TS 3/4.9.13, 
Emergency Exhaust System - Fuel Building, are also revised to eliminate 
the reference to the use of automatic control for the emergency exhaust 
system heaters.
    Date of issuance: August 16, 1994
    Effective date: August 16, 1994, to be implemented within 30 days 
of issuance.
    Amendment No.: 77
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17610) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 16, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: April 19, 1994
    Brief description of amendment: The amendment revises Technical 
Specification Table 3.6-1, ``Containment Isolation Valves,'' by 
deleting reference to two (2) valves. The technical specification 
change reflects a planned modification which removes the essential 
service water (ESW) containment air cooler return line isolation valve 
bypass valves and associated piping.
    Date of issuance: August 16, 1994
    Effective date: August 16, 1994
    Amendment No.: 78
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32239) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 16, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Dated at Rockville, Maryland, this 24th day of August 1994.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects I/II, Office of Nuclear Reactor 
Regulation
[Doc. 94-21325 Filed 8-30-94; 8:45 am]
BILLING CODE 7590-01-F