[Federal Register Volume 59, Number 91 (Thursday, May 12, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-11226]


[[Page Unknown]]

[Federal Register: May 12, 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 April 18, 1994, through April 29, 1994. The 
last biweekly notice was published on April 28, 1994 (59 FR 22000).

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, 11555 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 June 10, 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.

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

    Date of amendment request: March 30, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.4.9, Pressure/Temperature 
Limits, and its associated Bases, by changing the Unit 1 heatup and 
cooldown curves to incorporate a newly determined reactor vessel 
reference nil-ductility temperature, RTNDT, and by updating the 
removal schedule of vessel surveillance capsules for both units in 
accordance with ASTM E185-82. Changes would also be made to the Unit 1 
Low Temperature Overpressure Protection System (LTOPS) setpoint curve 
in TS 3.4.9.3 to reflect the new pressure/temperature limits.
    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. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The use of new pressure-temperature limit curves and low 
temperature overpressure protection curves will not change any 
postulated accident scenarios. The revised curves were developed using 
industry standards and regulations which are recognized as being 
inherently conservative. The pressure-temperature low temperature 
overpressure curves provide reactor coolant system (RCS) limits to 
protect the reactor pressure vessel from brittle fracture by clearly 
separating the region of normal operations from the region where the 
vessel is subject to brittle fracture. The heatup and cooldown limits 
are designed to ensure that the 10 CFR 50 Appendix G Pressure 
Temperature limits for the RCS are not exceeded during any condition of 
normal operation including anticipated operational occurrences.
    General Design Criterion 32 of 10 CFR 50 Appendix A requires that 
the reactor coolant boundary shall be designed with sufficient margin 
to assure that when stressed under operating, maintenance, testing, and 
postulated accident condition[s], (1) the boundary behaves in a 
nonbrittle manner and (2) the probability of rapidly propagating 
fracture is minimized.
    10 CFR 50 Appendix G, ``Fracture Toughness Requirements,'' requires 
that the effects of changes in the fracture toughness of reactor vessel 
materials caused by neutron radiation throughout the service life of 
[a] nuclear reactor be considered in the pressure-temperature limits. 
The change is used in conjunction with the material initial reference 
temperature (RTNDT) to establish the limiting pressure-temperature 
curves. Regulatory Guide 1.99, Rev. 2, contains procedures for 
calculating the effects of neutron radiation embrittlement of the low-
alloy steels currently used for light-water-cooled reactor vessels.
    Using the Regulatory Guide 1.99, Revision 2, Braidwood Unit 1 
Surveillance Capsule U results, and Appendix G to 10 CFR 50, new 
Pressure-Temperature curves [were] prepared for the projected reactor 
vessel exposure at 32 EFPY of operation. These new curves, in 
conjunction with the heatup and cooldown ranges and the revised Low-
Temperature Overpressure Protection System setpoints, provide the 
required assurance that the reactor pressure vessel is protected from 
brittle fracture up 32 EFPY of operation. No changes to the design of 
the facility have been made and no new equipment has been added or 
removed. The revised analysis and resultant adjustment of the operating 
limitations provide assurance that the Reactor Coolant System is 
protected from brittle fracture.
    Revising the Reactor Vessel Material Surveillance Program 
Withdrawal Schedule does not result in the addition or removal of any 
equipment, or any design changes to the facility. Capsule lead times 
are revised and, for Braidwood Unit 2, Capsule X will be removed next 
vice Capsule W. The proposed removal schedules remain consistent with 
ASTM 185-82.
    Therefore, the proposed amendment to the pressure temperature 
limitations does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    B. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The use of the new pressure-temperature operating limits and the 
new low temperature overpressure protection curve does not change any 
postulated accident scenarios. The new curves do not represent any 
appreciable change in the current methodologies; they merely provide 
assurance that the Reactor Coolant System is protected from brittle 
fracture. No new accident or malfunction mechanism is introduced by the 
amendment and no physical plant changes will result from this 
amendment.
    Revision of the Reactor Vessel Material Surveillance Program 
Withdrawal Schedule does not introduce a new accident or malfunction 
mechanism. Capsule lead times are revised, and, other than changing the 
order of specimen removal, consistent with ASME 185-82, no physical 
plant changes will result from this revised schedule.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    C. The proposed change does not involve a significant reduction in 
a margin of safety.
    The new pressure-temperature operating limits low temperature 
overpressure protection curves were generated with the currently 
accepted conservative methodology using capsule surveillance data. The 
new pressure-temperature curves were developed using industry standards 
and regulations (ASME Code Section III, and NRC Regulatory Guide 1.99, 
Revision 2) which are recognized as being inherently conservative. The 
use of the new pressure- temperature operating limits and low 
temperature overpressure protection limits would not change postulated 
accident scenarios.
    The proposed revision to the Reactor Vessel Material Surveillance 
Program Withdrawal Schedule would not change postulated accident 
scenarios. Capsule lead times are revised, and, other than changing the 
order of specimen removal, consistent with ASTM 185- 82, no physical 
plant changes will result from this revised schedule. 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: Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690.
    NRC Project Director: James E. Dyer.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of amendment request: March 31, 1994.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) to provide allowable outage 
times for automatic actuation channel surveillance testing and 
restoration time for an inoperable engineered safety feature actuation 
system automatic actuation channel.
    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 evaluation is provided for the three categories of 
the significant hazards consideration standards:
    a. Proposed changes to allow 8 hours for master relay and logic 
testing, 12 hours for slave relay testing and 6 hours to restore an 
inoperable ESFAS Automatic Actuation Channel prior to entering the 
shutdown action clock.
    (1) The determination that these changes are within all acceptable 
criteria was established in the NRC's SER prepared for WCAP-10271, 
Supplement 2, Revision 1. The Technical Specification changes proposed 
by this license amendment request conform to NRC guidance contained in 
the SER. The NRC found that implementation of the proposed changes is 
expected to result in a small and acceptable increase in ESFAS 
unavailability. This increase in probability results in a small 
increase in calculated core damage frequency and public risk. The 
calculated increase in core damage frequency was judged to be 
acceptable since the increase was small and well within the range of 
uncertainty associated with the analysis. The values presented in WCAP-
10271 Supplement 2 Revision 1 for increase in core damage frequency 
were verified by Brookhaven National Laboratory as part of an audit and 
sensitivity analyses performed for the NRC Staff.
    Based on the small increase in core damage frequency as compared 
with the range of uncertainty in the analysis, the NRC agreed that the 
calculated increase is acceptable. This conclusion was documented in 
the NRC's SER dated February 22, 1989. The applicability of these 
conclusions has been verified through a plant specific review of the 
generic analysis in WCAP-10271, Supplement 2, Revision 1. The ESFAS 
Automatic Actuation Channel allowed outage and restoration times 
included in this license amendment request are consistent with the 
generic analysis. In addition, the NRC stated that the majority of the 
increase in unavailability was due to the decrease in frequency of 
surveillance testing vice the changes in allowed outage and restoration 
times. Therefore, considering the above information, the proposed 
allowed outage and restoration time changes do not involve a 
significant increase in the probability of occurrence or consequences 
of an accident previously evaluated.
    (2) The proposed changes do not involve the physical alteration of 
any plant system and do not result in a change in the manner in which 
the ESFAS system performs its function. The increases in allowed outage 
and restoration times only affects the probability of the ESFAS 
Automatic Actuation Channel functioning properly as described above. 
Therefore, the allowed outage and restoration time changes proposed in 
this license amendment request do not create a new or different type of 
accident from any previously evaluated.
    (3) The proposed allowed outage time and restoration time changes 
do not alter the manner in which safety limits, limiting safety system 
setpoints or limiting conditions for operation are determined. The 
impact of the revised ESFAS Automatic Actuation Channel allowed outage 
and restoration times is addressed above. Implementation of the 
proposed changes is expected to result in an overall improvement in 
safety by allowing adequate time for required ESFAS testing and quality 
repairs leading to improved equipment reliability due to a more 
appropriate restoration time. Therefore, it may be concluded that the 
proposed allowed outage and restoration time changes do not involve a 
significant reduction in margin of safety.
    b. Proposed change to the minimum required degree of redundancy for 
the High-High Containment Pressure channels in Table 3.4-1.
    (1) Changing the minimum required degree of redundancy in Table 
3.4-1 for the High-High Containment Pressure Channels (Table 3.4-1 
items II.3, III.B.3, and IV.3) provides consistency with Technical 
Specification 3.4.2.c which allows an inoperable High-High Containment 
Pressure channel to be placed in bypass. Placement of an inoperable 
High-High Containment Pressure Channel in bypass is preferred to reduce 
the probability of an inadvertent containment spray event. Also, these 
channels are designed with a two out of four logic so that the failed 
channel may be bypassed rather than tripped. With the failed channel 
bypassed, single failure criterion is still met because the logic is 
now a two out of three. Furthermore, with the one channel bypassed, a 
single channel failure will not inadvertently initiate a containment 
spray. Therefore, this change can be considered an administrative 
change to correct Table 3.4-1 to agree with the Action requirements of 
Technical Specification 3.4.2.c. As such this proposed change does not 
involve an increase in the probability of occurrence or consequences of 
an accident previously evaluated.
    (2) Correcting the minimum required degree of redundancy in Table 
3.4-1 for the High-High Containment Pressure channels is an 
administrative change which does not involve the physical alteration of 
any plant system and does not result in a change in the manner in which 
the ESFAS system performs its function. Therefore, the proposed 
correction to Table 3.4-1 does not create the possibility of a new or 
different kind of accident from any previously analyzed.
    (3) Correcting the minimum required degree of redundancy in Table 
3.4-1 to be consistent with the Actions of Technical Specification 
3.4.2.c is an administrative change and as such does not involve any 
reduction in a margin of safety.
    c. Proposed change to the delete footnote +++ from Table 3.4-1.
    (1) Deleting footnote +++ from Table 3.4-1 removes the 
inconsistency between it and Technical Specification 3.4.2.c which 
states that channels other than the High-High Containment Pressure 
channels shall be placed in trip during testing. The change does not 
affect the manner in which ESFAS provides plant protection. In addition 
the change does not affect the functioning of ESFAS or the way Zion 
Station conducts channel testing. Instrument channel testing will 
continue to be conducted in the tripped mode with the exception of the 
High-High Containment Pressure channels, which can be tested in bypass 
because of the risk of a spurious Containment Spray event. Automatic 
Actuation Channel testing will be performed in accordance with the 
allowed outage times of new Specification 3.4.2.d. As such this 
proposed change does not involve any significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated.
    (2) Deleting footnote +++ from Table 3.4-1 does not involve the 
physical alteration of any plant system and does not result in a change 
in the manner in which ESFAS performs its function. Therefore this 
change does not involve the physical alteration of any plant system and 
does not result in a change in the manner in which the ESFAS system 
performs its function. Therefore, the proposed correction to Table 3.4-
1 does not create the possibility of a new or different kind of 
accident from any previously analyzed.
    (3) Deleting footnote +++ from Table 3.4-1 does not alter the 
manner in which safety limits, limiting safety system setpoints or 
limiting conditions for operation are determined. Implementation of 
this change will not alter ESFAS testing. Therefore implementation of 
this change does not involve any reduction in a margin of safety.
    d. Proposed editorial change to Technical Specification 3.4.2.c.
    The editorial change to Technical Specification 3.4.2.c to change 
``Containment Hi-Hi pressure channels'' to ``High-High Containment 
Pressure channels'' is purely an administrative change which has no 
affect on plant safety.
    e. Summary.
    The foregoing analyses demonstrate that the proposed License 
Amendment to the Zion Station Technical Specifications does not involve 
a significant increase in the probability of occurrence or consequences 
of a previously evaluated accident, does not create the possibility of 
a new or different kind of accident and does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: 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: James E. Dyer.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan

    Date of amendment request: April 22, 1994.
    Description of amendment request: The proposed change revises the 
reactor vessel pressure-temperature limits in the Technical 
Specifications. The change insures that the vessel fracture toughness 
requirements of Section V of 10 CFR Part 50, Appendix G, are satisfied 
through the end of life.
    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 the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The margin above Nil Ductility Transition Temperature (NDTT) is 
governed by 10 CFR 50 Appendix G and remains unchanged. The proposed 
change will not involve a significant increase in the probability or 
consequences of a previously evaluated accident.
    2. Will the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The predicted shifts in NDTT are based on a revised reference 
temperature consistent with Regulatory Guide 1.99, Revision 2, dated 
May 1988. This method of revising temperature-pressure limits is the 
same as in the past (ASME Code Section III, Appendix G).
    3. Will the proposed change involve a significant reduction in the 
margin of safety?
    The proposed curves were generated for an End of Licensed Life (May 
31, 2000) Effective Full Power Year exposure and are conservative in 
nature until that time. The margin of safety [is] unchanged.
    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: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: L. B. Marsh.

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: April 18, 1994.
    Description of amendment request: License Amendment No. 81, issued 
on July 15, 1993, changed the numbering of surveillance requirements 
for Technical Specifications 3/4.3.1, ``Control Rod Operability,'' 3/
4.3.2, ``Control Rod Maximum Scram,'' and 3/4.10.2, ``Rod Pattern 
Control System.'' However, Action Statements referencing these 
surveillance requirements were overlooked and were not appropriately 
renumbered. The purpose of the proposed technical specification change 
would be to renumber the overlooked references.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    (1) These changes do not affect the intent or implementation of the 
applicable Technical Specifications. The changes simply make the 
affected Technical Specifications consistent. Since these are only 
editorial changes which do not impact the plant design or operations, 
they cannot increase the probability or the consequences of any 
accident previously evaluated.
    (2) The proposed changes are editorial only and do not affect the 
plant design or operation. No new failure modes are introduced by such 
changes and, therefore, the request will not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    (3) The proposed changes merely correct an editorial oversight. 
These changes do not alter or delete any technical requirements and, 
therefore, 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, 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.

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: April 18, 1994.
    Description of amendment request: Test methods for carbon adsorber 
filters specified in Technical Specification Sections 3/4.6.6.3, 
``Standby Gas Treatment System,'' and 3/4.7.2, ``Control Room 
Ventilation System,'' specify the 1979 version of ASTM D3803. The 
proposed change would delete the year of the standard so that more 
recent versions of the standard could be used.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    (1) The proposed changes to the Technical Specification 
surveillance requirements for determining the methyl iodide penetration 
of carbon samples would not involve a significant increase in the 
probability or the consequences of any accident previously evaluated 
because the proposed change merely allows Illinois Power (IP) to 
utilize a more up-to-date version of the same test method currently 
specified. More recent versions of the test method are more effective 
at detecting unsatisfactory charcoal performance because they include 
equilibration periods to ensure that all samples have a common starting 
point before being challenged with radioactive gas. The proposed change 
would not affect the quality of the charcoal or the reliability of the 
filter subsystems as it only relates to testing and involves no changes 
to the design or operation of the ventilation subsystems themselves. 
The updated standards provide more accurate and repeatable test results 
and do not change the properties or acceptance criteria for these 
properties. As a result, the performance capabilities of the associated 
filter subsystems would not be adversely impacted by the proposed 
change.
    (2) The proposed change would not involve a change in the design or 
operation of any plant system or component. In addition, the proposed 
change would not reduce the level of filter train subsystem reliability 
nor would it create an initiating event for any accident. Because the 
performance, function, and redundancy of the original design remain 
unchanged, the proposed change would not create the potential for a new 
event. Furthermore, since no new types of equipment would be introduced 
into the plant design and the proposed change would not adversely 
impact existing equipment, no potential for a different type of 
malfunction is created by the proposed change. Therefore, this proposed 
change cannot create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) The margin of safety for the charcoal filter subsystems as 
defined in the Bases to the Technical Specifications associated with 
the proposed change refers to the ability of the filters to remove 
radioiodines. The proposed change would allow IP to upgrade the 
currently specified test for determining charcoal adsorber performance 
with one which utilizes the same type of methodology, but provides 
greater accuracy and repeatability. The newer versions of the test 
method are more effective at detecting unsatisfactory charcoal 
performance because they include equilibration periods to ensure that 
all samples have a common starting point before being challenged with 
radioactive gas. Thus, the proposed change would not involve a 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    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.

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

    Date of amendment request: March 28, 1994.
    Description of amendment request: The proposed amendment would 
revise technical specifications Tables 3.2.4 and 4.2.1, to change one 
of the initiating parameters of the reactor building ventilation 
isolation system and standby gas treatment system (SGTS) from Low 
Reactor Water Level to Low Low Reactor Water Level. This revision is 
being made in order to improve plant performance by reducing the 
potential for unnecessary secondary containment isolation and SGTS 
initiations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The function of the Standby Gas Treatment System and secondary 
containment is to mitigate the consequences of a loss of coolant 
accident and fuel handling accidents. The proposed changes maintain 
this capability. The revised Standby Gas Treatment System initiation 
and secondary containment isolation parameter of low low reactor water 
level provides the required detection of loss of coolant accidents and 
is consistent [with] ECCS actuation to mitigate the consequences of 
this accident. The low low reactor water level instrumentation is set 
to trip when reactor water level is 6'6'' above the top of the active 
fuel. This trip currently initiates closure of the Group 1 Primary 
containment isolation valves, activates the Emergency Core Cooling 
systems and starts the emergency diesel generator. This trip setting 
level was chosen to be low enough to prevent spurious operation but 
high enough to initiate Emergency Core Cooling system operation and 
primary system isolation so that no melting of the fuel cladding will 
occur, post accident cooling can be accomplished, and the guidelines of 
the 10 CFR 100 will not be violated. Therefore, this amendment will not 
cause a significant increase in the probability or consequences of an 
accident previously evaluated for the Monticello plant.
    The proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously analyzed. The 
proposed changes to Technical Specifications for the standby gas 
treatment system and secondary containment do not alter the function 8 
of the systems or its interrelationships with other systems. An adverse 
interaction which could be postulated to occur is the initiation of the 
Standby Gas Treatment System without a coordinated trip of the 
Mechanical Vacuum Pump. The Mechanical Vacuum Pump is operated during 
plant startups to draw a vacuum on the main condenser prior to 
admission of steam. The Mechanical vacuum discharges to the offgas 
stack and thus can create a back pressure on the Standby Gas Treatment 
System, reducing initiation of Standby Gas Treatment System flow below 
required values. The proposed initiation of Standby Gas Treatment 
System on low low reactor water level maintains the necessary 
coordination by having the Standby Gas Treatment System initiate 
subsequent to isolation or tripping of the Mechanical Vacuum Pump on a 
low reactor water level signal from the primary containment isolation 
logic. Therefore, this amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed amendment will not involve a significant reduction in 
the margin of safety.
    The proposed amendment changes the initiation of the Standby Gas 
Treatment System and secondary containment isolation from being 
concurrent with the low reactor water signal (which is indicative that 
the reactor core is in danger of being inadequately cooled) to being 
concurrent with reactor low low water level (which is also an indicator 
that the capability to cool the core is threatened and assures that no 
melting of the fuel cladding will occur, post accident cooling can be 
accomplished, and the guidelines of 10 CFR 100 will not be violated). A 
review of the accident analyses provided in Section 14 of the USAR has 
determined that these analyses did not specifically credit initiation 
of the Standby Gas Treatment Systems and secondary containment 
isolation at the accident precursor reactor water level of low level. 
Furthermore, this review determined that the low low reactor water 
level setpoint has no adverse impact on the ability of the Standby Gas 
treatment System and secondary containment to perform its design basis 
function as credited in the accident analyses.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Ledyard B. Marsh.

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

    Date of amendment request: March 24, 1994.
    Description of amendment request: The proposed modification to 
Technical Specification (TS) Sections 3.11.1.4, 6.9.1.8, and 6.14.1 
would change the frequency for submitting the Semiannual Radioactive 
Effluent Release Report to the NRC from semiannually to annually.
    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 Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed TS changes are administrative in nature. The proposed 
changes simply involve revising the frequency for submitting the 
Semiannual Radioactive Effluent Release Report to the NRC from 
semiannually to annually in order to implement the amended reporting 
requirements of 10 CFR 50.36a as promulgated in Final Rule 57 FR 39353. 
Since the information contained in this report is reviewed and 
evaluated after the effluents are released, no accidents previously 
evaluated are impacted by the proposed TS changes. Radiological 
effluent releases from the station will continue to be controlled as 
required by the TS, including those requirements specified in the 
Offsite Dose Calculation Manual (ODCM) and Process Control Program 
(PCP). The proposed TS changes do not involve any changes to the 
operation or physical configuration of any plant systems or equipment. 
The proposed changes do not impact any initial or final accident 
conditions or assumptions previously evaluated. The radiological 
consequences of these previously evaluated accidents are not affected 
by the proposed changes.
    Therefore, the proposed TS changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed TS changes are administrative in nature. The proposed 
changes simply involve revising the frequency for submitting the 
Semiannual Radioactive Effluent Release Report to the NRC from 
semiannually to annually in order to implement the amended reporting 
requirements of 10 CFR 50.36a as promulgated in Final Rule 57 FR 39353. 
Radiological effluent releases from the station will continue to be 
controlled as required by the TS, including those requirements 
specified in the ODCM and PCP. The proposed TS changes do not involve 
any modifications to plant systems or equipment.
    Therefore, the proposed TS changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. The proposed TS changes do not involve a significant reduction 
in a margin of safety.
    The proposed TS changes are administrative in nature, and will only 
involve revising the frequency for submitting the Semiannual 
Radioactive Effluent Release Report to the NRC from semiannually to 
annually as currently stipulated in 10 CFR 50.36a. The specific 
radiological effluent release information contained in this report will 
continue to be provided as required. The station radiological effluent 
releases will continue to be controlled as required by TS, including 
those requirements specified in the ODCM and PCP.
    Therefore, the proposed TS 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 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101.
    NRC Project Director: Charles L. Miller.

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

    Date of amendment request: February 25, 1994.
    Description of amendment request: The proposed Technical 
Specification (TS) changes would permit the submittal of the 
Radioactive Effluents Release Report on an annual rather than a 
semiannual basis; allow changes to the Offsite Dose Calculation Manual 
(ODCM) to be submitted in the Radioactive Effluent Release Report 
rather than in the monthly operating report; remove the title of 
Executive Vice President--Operations from the TS; remove the list of 
audit frequencies from the TS and place them under Quality Systems 
management; change the title of Associate Manager, Health Physics to 
Radiation Protection Manager; remove references to specific letters; 
remove TS 6.4 on training; and correct various typographical errors.
    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 the administrative change does not affect plant operations in 
any manner.
    2. The proposed amendment does not create the possibility of a new 
or different kind of accident than previously evaluated because the 
proposed change is administrative in nature and no physical alterations 
of plant configuration or changes to setpoints or operating parameters 
are proposed.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety. The change is only administrative.
    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 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: March 11, 1994.
    Description of amendment request: The proposed Technical 
Specification (TS) changes would delete surveillance requirement 
4.8.1.4.a.3, which requires periodic testing of penetration protection 
fuses, and its associated Basis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The probability or consequences of an accident previously 
evaluated is not significantly increased.
    Fuses are simple protection devices and can only degrade by being 
more resistive which is in the conservative direction. The proper type 
and size fuse is assured as part of design, procurement, and initial 
installation. The testing provides no additional assurance of 
operability. Therefore, the deletion of periodic retesting of these 
fuses will not increase the probability or consequences of an accident 
previously evaluated.
    2. [The proposed amendment will not] [c]reate the possibility of a 
new or different kind of accident from any previously analyzed.
    The design of the penetration protection and the installation of 
the fuses has not changed in any way. Any undetected failure of a fuse 
would fall under single failure criteria. A current limiting fuse must 
have high electrical current in order to perform its intended function. 
Any fuse which has opened the circuit through the penetration would be 
detected. (This is not a concern of the Technical Specifications.) 
Therefore, this change does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety. Deletion 
of this surveillance requirement will not minimize the intent of this 
Technical Specification. This TS is to assure continued operability of 
the containment penetration conductor overcurrent protection which 
helps to ensure containment integrity. Testing, however, may introduce 
the potential for damage to the fuses and fuse clips. Therefore, the 
deletion of this TS requirement 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 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: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180.
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
    NRC Project Director: William H. Bateman.

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 amendment request: November 22, 1993.
    Description of amendment request: The proposed amendment would 
result in the replacement of most of the analog Riley temperature 
instrumentation associated with leak detection with digital equipment 
from the General Electric Company NUMAC product line. Technical 
Specification changes would be made to instrumentation surveillance 
requirements for temperature instruments associated with main steam 
line isolation, reactor water cleanup system isolation, reactor core 
isolation cooling system isolation, and residual heat removal system 
isolation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The Technical specifications are proposed to be revised to perform 
a Channel Functional Test on a semiannual frequency versus the current 
monthly frequency for both ambient and differential temperature and for 
the MSL tunnel temperature timer functions, for the above listed piping 
lines. Additionally, this evaluation addressees the potential for, and 
implications of, common mode failures due to software, hardware and/or 
electromagnetic and radio-frequency interference (EMI/RFI).
    The NUMAC instrumentation has certain design features which 
contribute to its reliability. The replacement NUMAC LDMs are digital 
instruments that use a microcomputer to monitor the ambient and 
differential temperatures (also the MSL tunnel temperature timer) and 
provide outputs and automatic self-testing and calibration. A 
description of the major design features include: a) isolation of the 
essential microcomputer by a serial data link from the front panel 
display (and display microcomputer), b) a self-test system feature that 
provides automatic testing of internal circuits and reports failures, 
c) thermocouple failure detection, d) provisions to test the output 
relays without the use of jumpers (reducing the threat of spurious 
isolation), and e) two independent built-in instrument power supplies 
(that automatically switchover to the other supply in the event of 
failure). Also, several features have been included, among these, a) a 
hardware ``watchdog'' timer to monitor against software cycling in 
continuous loops, and b) software structured with the safety-related 
essential tasks running at the highest priority in the system. These 
capabilities increase the reliability of the collected data, reduce the 
possibility of inadvertent isolation and plant shutdowns, reduce the 
need for frequent calibrations, and reduce the likelihood of common 
mode failures.
    The NUMAC Leak Detection Monitors will maintain the same 
environmental and electrical physical independence criteria 
(qualifications) as the existing Leak Detection System components. The 
LDMs, the associated thermocouple input units (TCIUs), and relay output 
units (ROUs) will be mounted seismically such that qualification of 
these components and the Control Room panels will be maintained. The 
LDMs are qualified for the PNPP Control Room environment. The LDMs (one 
per division) will be physically and electrically independent of each 
other and do not share power supplies, thermocouple inputs, output 
relays, microcomputer logic units, display units or enclosures and 
mounting locations. A postulated gross failure of any one NUMAC LDM, 
such as gross malfunction of the input unit, microcomputer logic unit 
or the relay output unit, will not propagate to the other NUMAC LDM, 
such as gross malfunction of the input unit, microcomputer logic unit 
or the relay output unit, will not propagate to the other NUMAC LDM. 
Thus, a failure within one NUMAC LDM will not prevent or disable the 
function of the other NUMAC LDM. A failure within one NUMAC LDM may 
cause the loss of one division of the isolation trip logic. However, 
since the other redundant division (the MSLs have three other 
divisions) will not be affected by this failure, the Leak Detection 
System will still be able to perform its designed safety-related 
function and provide the necessary system isolation. This is the same 
as the current Leak Detection System design basis.
    The possibility of a common mode failure of both NUMAC LDM 
divisions is minimized by the design of the NUMAC hardware and 
software, the verification and validation (V&V) of the software to 
reduce the likelihood of errors, the testing of the software (to 
discover and eliminate errors), and the design of and testing of the 
hardware to demonstrate its resistance to EMI/RFI. The NUMAC instrument 
design features, by effectively eliminating the potential for common 
mode failures, maintain the Leak Detection System within its current 
licensing basis. (A discussion of common mode failure protection is 
presented in more detail in the answer to question two.) Therefore, the 
design, isolation and separation criteria remain the same.
    Additionally, as described within Chapter 7 of the Updated Safety 
Analysis Report (USAR), diversity is provided to the ambient and 
differential temperature monitoring trip functions for the various 
systems by alternative leak detection methods (such as measuring steam 
line flow or pressure) that provide backup in the event of the loss of 
both divisions of the NUMAC Leak Detection Monitors. These alternative 
leak detection methods are physically separate from those being 
performed by the NUMAC LDM and constitute a diverse, redundant, safety 
related backup capable of responding to a design basis line break for 
the various systems. Therefore, a common mode failure of both LDM 
divisions would not prevent any of the necessary system isolation from 
occurring.
    No changes are being made to the isolation logic of the Leak 
Detection System. No accident initiators or precursors are affected by 
the proposed changes to the Channel Functional Test surveillance 
intervals for the various trip functions. One purpose of a Channel 
Functional Test is to check the instrument setpoints. The NUMAC 
instrument setpoints are set digitally and do not drift. An engineering 
evaluation has established that the Channel Functional Test 
surveillance interval can be extended from one to six months. The 
potential for common mode failures has been accounted for in the design 
and measures have been taken to lower the probability of this to an 
acceptable level (see the answer to question two). Also, alternative 
leak detection methods exist for this eventuality. Since the NUMAC Leak 
Detection Monitoring equipment meets or exceeds the design and 
licensing criteria specified for the Leak Detection System, the 
proposed upgrade cannot increase the probability of occurrence of any 
accident previously evaluated.
    A portion of the Leak Detection System logic causes a closure of 
the Main Steam Isolation Valves on a steam leak signal. This transient, 
described in Chapter 15 of the USAR, may also occur due to a LDS 
equipment malfunction. Since this modification replaces some of the 
existing Riley temperature monitoring instrumentation with more 
reliable instrumentation the probability of this transient is reduced 
(no radiological consequences are associated with this event). The LDS 
is also used to mitigate the consequences of a pipe break outside 
primary containment by isolating the affected system connected to the 
Reactor Coolant Pressure Boundary (RCPB). The replacement of the Riley 
instrumentation with NUMAC LDMs will not change, degrade, or prevent 
the Leak Detection System response to mitigate the radiological 
consequences of an accident. Therefore, replacement of the Riley 
temperature modules with NUMAC Leak Detection Monitors will not 
significantly increase the consequences of any accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The single failure criterion requires that any single failure 
within a safety-related system not prevent proper protective action of 
the overall system when the system is required to function. The Leak 
Detection System design is such that a failure of one division will not 
prevent the system from performing its safety function. Common mode 
failure protection provisions have been addressed in the NUMAC LDM 
design.
    The comprehensive General Electric software V&V and configuration 
management control programs minimize, although they cannot entirely 
eliminate, the likelihood of a common mode NUMAC instrument failure due 
to software problems. The hardware (firmware) and software for the PNPP 
NUMAC Leak Detection Monitors will undergo a formal software 
verification and validation (V&V) process by General Electric, that is 
to be completed by the end of the year, equivalent to the one reviewed 
and approved by the NRC for the safety-related Wide Range Neutron 
Monitor.
    The NUMAC instruments are designed to minimize both their 
susceptibility to, and generation of, electromagnetic and radio-
frequency interference (EMI/RFI) to prevent spurious operations and 
allow their use in safety-related systems. As part of a broader plan by 
GE to improve the testing has been performed by GE on the Leak 
Detection Monitor configuration in order to both expand the overall 
qualification region, and to obtain test data specific to this 
application. This testing ensures the qualification of the Thermocouple 
Input Unit (TCIU), a NUMAC circuit board which is unique to the LDM 
application, and also extends the NUMAC EMI/RFI qualification region to 
include both higher and lower frequencies than previously tested.
    The NUMAC instrument design concept has undergone review by the 
NRC, and the initial instruments of the NUMAC product line (the 
Logarithmic Radiation Monitor and Wide Range Neutron Monitoring System) 
have received NRC approval via Safety Evaluation of the associated GE 
Licensing Topical Reports. The various types of NUMAC equipment in 
operation at other nuclear power plants have components and software 
modules which are similar to and in some instance identical to the 
NUMAC LDMs. Therefore, based on the NRC reviewed and approved NUMAC 
software and hardware control programs instituted by GE, the design 
features to minimize software/hardware (or their interface) problems, 
design features to minimize susceptibility to EMI/RFI, and testing to 
demonstrate resistance to EMI/RFI, installation of NUMAC Leak Detection 
Monitors at the PNPP does not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed changes do not involve a significant reduction in a 
margin of safety.
    The replacement of the analog Riley temperature modules with the 
microcomputer based NUMAC Leak Detection Monitors will not affect any 
design conditions or impact the margins of safety for the various Leak 
Detection System monitored parameters in the Technical Specification 
Table 3.3.2-2 will not be changed or affected by this modification. 
Only the CHANNEL FUNCTIONAL test interval is being extended.
    The NUMAC Leak Detection Monitor design, with the attention paid 
towards minimizing the potential for, and the effect of, software/
hardware and/or EMI/RFI related problem or common mode failures and 
resulting operational experience has demonstrated that replacement of 
the existing Riley temperature modules with NUMAC Leak Detection 
Monitors would not result in a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street NW., 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: February 8, 1994, as supplemented March 
25, 1994.
    Description of amendment request: The amendment would revise the 
WNP-2 Technical Specifications. Specifically, the amendment would 
increase the stroke time, as specified in Table 3.6.3-1, for reactor 
core isolation cooling (RCIC) valve RCIC-V-8, from 13 seconds to 26 
seconds and the note (j) reference would be deleted from RCIC-V-8 and 
RCIC-V-63. The note (j) indicates that the stroke time specified in the 
Table reflects the requirement for containment isolation only.
    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. This proposed action does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
RCIC-V-8 and V-63 are containment isolation valves and are normally 
open. Failure of the valves to open or close cannot cause an accident. 
The mitigating capability of RCIC-V-8 and V-63 is not changed in that 
the valves will continue to be closed within the established time 
limits. This ensures protection of the safety related equipment 
necessary for continued compliance with the requirements of General 
Design Criterion 4. In those accidents which involve a source term and 
potential adverse dose release consequences, no credit is taken for the 
closing of the valves; therefore the increase in the allowable time for 
closing does not increase the consequences of those accidents.
    2. This proposed action does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. The 
requested Technical Specifications change does not represent a change 
in modes of operation. It does not, in itself, require physical 
modification to the plant, although it will be used to allow a gear 
change in RCIC-MO-8. The new gears represent a standard configuration 
for Limitorque motor operators and will require a routine design 
change. The required Technical Specification change maintains the 
licensing basis for the plant as discussed in response to question 1. 
Hence, no new or different kind of accident is possible as a result of 
implementing this change.
    3. This proposed action does not involve a significant reduction in 
a margin of safety. The increase in stroke time will increase the peak 
temperature in the HELB profiles and thereby decrease the margin 
available from the equipment qualification limits. However, sufficient 
margin remains to assure the equipment operability is maintained and 
there is no reduction in the margin of safety. Additionally, there is 
no reduction in the margin of safety because increasing the stroke time 
will not change the postulated radiological releases.
    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.

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 amendments request: April 14, 1994.
    Description of amendments request: The proposed change would 
relocate the instrument response time tables 3.3.1-2, Reactor 
Protection System (RPS) Response Times; 3.3.2-3, Isolation System 
Instrumentation (ISI) Response Time; and 3.3.3-3, Emergency Core 
Cooling System (ECCS) Response Times, from the Technical Specifications 
to the Updated Final Safety Analysis Report. The RPS, ISI, and ECCS 
instrument limiting conditions for operation (LCO) will be revised to 
read that the instruments ``shall be operable'' without a reference to 
a specific response time table in these LCOs. The references to the 
response time tables will also be deleted from the Surveillance 
Requirements.
    Date of publication of individual notice in Federal Register: April 
26, 1994 (59 FR 21785).
    Expiration date of individual notice: May 26, 1994.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 
1, West Feliciana Parish, Louisiana

    Date of application for amendment: January 14, 1994.
    Brief description of amendment request: The proposed amendment 
would revise various instrumentation technical specifications by 
extending the allowable outage times (AOTs) of the instruments, and by 
increasing their channel functional surveillance test intervals (STIs) 
to quarterly. The amendment also revises certain technical 
specification actions to address loss-of-function concerns associated 
with the AOT and STI changes.
    Date of individual notice in Federal Register: April 26, 1994(59 FR 
21787).
    Expiration date of individual notice: May 26, 1994.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 
1, West Feliciana Parish, Louisiana

    Date of application for amendment: February 22, 1994.
    Brief description of amendment request: The proposed amendment 
would revise the technical specifications (TS) for the main steam-
positive leakage control system (MS-PLCS) and the penetration valve 
leakage control system (PVLCS) to be consistent with the requirements 
contained in NUREG-1434, ``Standard Technical Specifications, General 
Electric Plants (BWR/6).''
    Date of individual notice in Federal Register: March 10, 1994 (59 
FR 11331).
    Expiration date of individual notice: April 11, 1994.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Gulf States Utilities Company, Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 
1, West Feliciana Parish, Louisiana

    Date of application for amendment: March 3, 1994.
    Brief description of amendment request: The amendment would revise 
the technical specifications in accordance with the guidance provided 
by Generic Letter 93-08, ``Relocation of Technical Specification Tables 
of Instrument Response Time Limits.'' Generic Letter 93-08 recommends 
the removal and subsequent relocation of various technical 
specification tables which denote instrument and system response time 
limits.
    Date of individual notice in Federal Register: March 16, 1994 (59 
FR 12380).
    Expiration date of individual notice: April 15, 1994.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

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 Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of application for amendments: October 26, 1993, as 
supplemented March 28, 1994.
    Brief description of amendments: The licensee is requesting a 
revision to TS 5.3.1 for Palo Verde Nuclear Generating Station Units 1, 
2, and 3 that will increase the maximum allowable fuel enrichment from 
4.05 weight percent U-235 to 4.30 weight percent U-235. There was no 
change requested to the current 52,000 MWD/MTU burnup. The licensee 
provided a supplemental letter dated March 28, 1994, at the request of 
the NRC to bring TS 5.3.1 into conformance with Generic Letter 90-02, 
Supplement 1 and to clarify assumptions used in the Fuel Handling 
Accident Analysis.
    Date of issuance: April 19, 1994.
    Effective date: April 19, 1994, to be implemented within 45 days of 
issuance.
    Amendment Nos.: 74, 60 and 46.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2860) The additional information contained in the supplemental letter 
dated March 28, 1994, was clarifying in nature and thus within the 
scope of the initial notice and did not affect the NRC staff's proposed 
no significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated April 19, 1994. No significant hazards consideration 
comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

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

    Date of application for amendments: September 1, 1992.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.4.3, ``Relief Valves,'' to improve the 
reliability of the reactor coolant system's power-operated relief 
valves (PORVs) and their associated block valves for overpressure 
protection during normal operation and anticipated transients. The 
amendments also revise TS 3/4.4.9, ``Pressure/Temperature Limits,'' to 
improve the availability of the PORVs for low temperature overpressure 
protection. Accompanying changes are also made to the associated TS 
Bases. These revisions were made in response to Generic Letter 90-06, 
``Resolution of Generic Issue 70, `Power-Operated Relief Valve and 
Block Valve Reliability,' and Generic Issue 94, `Additional Low-
Temperature Overpressure Protection for Light-Water Reactors,' pursuant 
to 10 CFR 50.54 (f).''
    Date of issuance: April 20, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 188 and 165.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 30, 1992 (57 
FR 45076).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated April 20, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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 application for amendment: February 4, 1994.
    Brief description of amendment: The amendment revises TS 
Surveillance Requirement 4.6.4.1 to delete the 12-hour channel check, 
thereby eliminating the need for continuous operation of the hydrogen 
monitors.
    Date of issuance: April 26, 1994.
    Effective date: April 26, 1994.
    Amendment No. 47.
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10001) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 26, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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 application for amendments: March 21, 1994, as supplemented 
March 24, 1994.
    Brief description of amendments: The amendments add a one-time 
revision to Technical Specification (TS) 3/4.7.1.1 to permit continued 
activities at all four units with main steam Code safety valve lift 
setpoint tolerances of 3%. The duration of this amendment 
is until May 9, 1994, at which time the tolerances will be reset to 
1%. A statement has also been added to TS 4.7.1.1 for 
Braidwood stating that the provisions of TS 4.0.4 are not applicable to 
Braidwood, Unit 1, Cycle 5 until initial entry into Mode 2 from its 
refueling outage.
    Date of issuance: April 18, 1994.
    Effective date: April 18, 1994.
    Amendment Nos.: 61, 61, 49, and 49.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
Amendment revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (59 FR 14685 dated March 29, 1994). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination within 15 days. No 
comments have been received. The notice also provided an opportunity to 
request a hearing by April 29, 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 and 
final no significant hazards consideration determination is contained 
in a Safety Evaluation dated April 18, 1994.
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

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

    Date of application for amendment: September 29, 1993, as 
supplemented by letter dated April 8, 1994.
    Brief description of amendment: The amendment revises Technical 
Specification Section 3.9.A.5 and Tables 3.9-1 and 4.10-2 to delete 
controls for the 21, 22, and 23 Boron Monitor Tanks, which are no 
longer in service.
    Date of issuance: April 28, 1994.
    Effective date: As of the date of issuance to be implemented after 
the inlet and outlet lines of the 21, 22, and 23 Boron Monitor Tanks 
have been cut and capped.
    Amendment No.: 169.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62154)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 28, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of application for amendment: May 23, 1993.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3.4.3.2.d and related Table 3.4.3.2-1 by changing 
the allowable leakage for certain low pressure coolant injection (LPCI) 
line pressure isolation valves and revises Table 3.6.3-1 to remove the 
designation as containment isolation valves from the LPCI injection 
reverse flow check and bypass valves. The related Bases are also 
changed. Concurrently, the Commission granted an exemption from the 
requirements of 10 CFR Part 50, Appendix J, III.C. for performing Type 
C containment integrated leak rate tests of the containment isolation 
valves in the low pressure coolant injection lines of the residual heat 
removal system and to perform alternative testing.
    Date of issuance: April 22, 1994.
    Effective date: April 22, 1994, with full implementation within 45 
days.
    Amendment No.: 98.
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46227) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 22, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

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

    Date of amendment request: February 24, 1993.
    Brief description of amendment: The amendment corrected 
typographical errors in the plant technical specifications (TSs). These 
errors were introduced in the original ANO-1 TS, and in subsequent 
amendments. These changes are administrative in nature and are intended 
to improve the readability of the plant technical specifications 
without changing the meaning or intent of any specifications.
    Date of issuance: April 26, 1994.
    Effective date: April 26, 1994.
    Amendment No.: 171.
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67843)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 26, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

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

    Date of amendment request: November 16, 1993, as supplemented by 
letter dated April 5, 1994.
    Brief description of amendment: The amendment revised the Technical 
Specifications to provide acceptable conditions for operation when the 
core operating limit supervisory system (COLSS) is out of service and 
either or both control element assembly calculators (CEACs) are 
operable.
    Date of issuance: April 22, 1994.
    Effective date: April 22, 1994.
    Amendment No.: 93.
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
620) The additional information contained in the supplemental letter 
dated April 5, 1994, withdrew a portion of the original application and 
thus, was within the scope of the initial notice and did not affect the 
staff's proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated April 22, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

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

    Date of amendment request: February 14, 1994.
    Brief description of amendment: The amendment revised the Technical 
Specifications in response to Generic Letter 93-08 issued by the NRC 
and dated December 29, 1993, by relocating the reactor trip system and 
engineered safety features actuation system response time limits to the 
updated final safety analysis report.
    Date of issuance: April 22, 1994.
    Effective date: April 22, 1994.
    Amendment No.: 94.
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12360) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 22, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: February 25, 1992.
    Brief description of amendments: The amendments relate to your 
application dated February 25, 1992, which requested a 40-year 
operating license commencing from the date of issuance of the operating 
license and, accordingly, would extend the operating license expiration 
date for Turkey Point Units 3 and 4 to July 19, 2012 and April 10, 
2013, respectively.
    Date of issuance: April 20, 1994.
    Effective date: April 20, 1994.
    Amendment Nos. 162 and 156.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 15, 1992 (57 FR 
13130) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 20, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey 
Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: December 28, 1993.
    Brief description of amendments: These amendments include steam 
generator overfill protection in the Technical Specifications in 
response to Generic Letter 89-19, Request for Action Related to 
Resolution of Unresolved Safety Issue A-47 ``Safety Implications of 
Control Systems in LWR Nuclear Power Plants.''
    Date of issuance: April 28, 1994.
    Effective date: April 28, 1994.
    Amendment Nos. 163 and 157.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10007) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 28, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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: March 14, 1994.
    Brief description of amendment: The amendment changes the Millstone 
Unit 2 Technical Specifications (TS) to provide a one-time extension of 
the surveillance frequency from the required 18-month to the next 
refueling outage but no later than September 30, 1994, of the power 
operated valves in the service water system (TS 4.7.4.1.b) and in the 
boron injection flowpath (TS 4.1.2.2.c). This extends the surveillance 
for these valves approximately 5 months.
    Date of issuance: April 22, 1994.
    Effective date: April 22, 1994.
    Amendment No.: 173.
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 23, 1994 (59 FR 
13751). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 22, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Northeast Nuclear Energy Company, Docket Nos. 50-245, 50-336, and 50-
423, Millstone Nuclear Power Station, Units 1, 2 and 3, New London 
County, Connecticut

    Date of application for amendment: December 22, 1993.
    Brief description of amendment: The amendments revise the Technical 
Specifications (TS) as follows:
    1. Change the title of the Nuclear Station Director to Senior Vice 
President--Millstone Station.
    2. Remove the requirement to provide a copy of Plant Operations 
Review Committee (PORC) and Site Operations Review Committee (SORC) 
meeting minutes to the Executive Vice President--Nuclear. The Senior 
Vice President--Millstone Station replaces the Executive Vice 
President--Nuclear for receipt of PORC and SORC meeting minutes.
    3. Make editorial changes to the Millstone Unit No. 1 TS Index.
    4. Correct a typographical error in Section 6.2.1.d of the 
Millstone Unit No. 1 TS.
    5. Correct a typographical error in Section 6.5.3.1.a of the 
Millstone Unit No. 3 TS.
    Date of issuance: April 26, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 74, 174, and 90.
    Facility Operating License No. DPR-21, DPR-65, AND NPF-49. 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7693) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 26, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

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: April 19, 1993, as supplemented 
by letter dated April 18, 1994.
    Brief description of amendments: These amendments extend 
surveillance test interval and allowed outage times for the containment 
isolation actuation instrumentation.
    Date of issuance: April 26, 1994.
    Effective date: April 26, 1994.
    Amendment Nos. 69 and 32.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34086) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 26, 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, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania

    Date of application for amendments: May 6, 1993, as supplemented by 
letter dated April 18, 1994.
    Brief description of amendments: These amendments extend 
surveillance test interval and allowed outage times for selected 
actuation instrumentation and makes editorial changes.
    Date of issuance: April 26, 1994.
    Effective date: April 26, 1994.
    Amendment Nos. 70 and 33.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34087) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 26, 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 Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit 
Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: November 19, 1993.
    Brief description of amendments: These amendments eliminate the 
listing of specific position titles for the Plant Operations Review 
Committee (PORC) composition in favor of allowing the Plant Manager to 
appoint PORC members. This revision eliminates the need to change the 
TS in the future whenever a position title is changed.
    Date of issuance: April 26, 1994.
    Effective date: April 26, 1994.
    Amendments Nos.: 190 and 195.
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
628) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 26, 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.

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

    Date of application for amendment: January 31, 1994.
    Brief description of amendment: This amendment to the Appendix B 
Technical Specifications (TSs), the Radiological Effluent TSs, revised 
Section 3.5, and the associated Bases, to establish a threshold level 
below which there will be no requirement to perform grab samples and 
isotopic analyses of steam jet-air ejector (SJAE) effluent and revised 
TS Table 3.10-1 to change the actions required when entering an SJAE 
limiting condition for operation. Additionally, the amendment revised 
the TSs to clarify instructions and make editorial corrections which 
are administrative in nature.
    Date of issuance: April 25, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 211.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10014).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 25, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: March 12, 1993.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TS) to incorporate the changes listed below:
    (1) The frequency of high pressure water fire protection system 
testing (specified in TS Section 4.12.A.1) was changed to accommodate 
operation on a 24-month cycle.
    (2) The frequency of fire pump diesel engine testing (specified in 
TS Section 4.12.A.2) was changed to accommodate operation on a 24-month 
cycle.
    (3) The frequency of electrical tunnel, diesel generator building, 
and containment fan cooler fire protection spray and/or sprinkler 
system testing (specified in TS Section 4.12.B.1) was changed to 
accommodate operation on a 24-month cycle.
    (4) The frequency of fire barrier penetration seal inspection 
(specified in TS Section 4.12.C.1) was changed to accommodate operation 
on a 24- month cycle.
    (5) The frequency of fire detection system testing (specified in TS 
Section 4.12.D.1) was changed to accommodate operation on a 24-month 
cycle.
    (6) The frequency of fire hose station testing (specified in TS 
Section 4.12.E.1) was changed to accommodate operation on a 24-month 
cycle.
    (7) The frequency of CO2 fire protection system testing 
(specified in TS Section 4.12.G.1) was changed to accommodate operation 
on a 24-month cycle. A new requirement was also added to exercise the 
fire dampers on an annual basis.
    These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle,'' as applicable.
    In addition, TS Section 4.12 was reformatted, in its entirety, and 
several administrative changes were made to improve clarity.
    Date of issuance: April 20, 1994.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 146.
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1993 (58 FR 
25862) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 20, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: March 4, 1994.
    Brief description of amendment: This amendment adds a new TS 3/
4.10.8, ``Inservice Leak and Hydrostatic Testing,'' to the Hope Creek 
Generating Station TSs. The amendment also includes corresponding 
changes to the TS Index, Table 1.2, ``OPERATIONAL CONDITIONS,'' and 
provides Bases for TS 3/4.10.8. The added TS 3/4.10.8 permits the unit 
to remain in OPERATIONAL CONDITION 4 with the average reactor coolant 
temperature being increased above 200 deg.F, but not to exceed 
212 deg.F, and certain OPERATIONAL CONDITION 3 Limiting Conditions for 
Operation for secondary containment isolation, secondary containment 
integrity and filtration, recirculation and ventilation system (FRVS) 
operability being met.
    Date of issuance: April 18, 1994.
    Effective date: April 18, 1994.
    Amendment No.: 69.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12384).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 18, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: December 23, 1992, as 
supplemented August 12, 1993 and January 21, 1994 (TS 328).
    Brief description of amendments: The amendments modify the 
operability requirements for the low pressure coolant injection (LPCI) 
mode of the residual heat removal (RHR) system while the reactor is 
shut down. The amendments permit the RHR system to be considered 
operable for LPCI when aligned for shutdown cooling if it can be 
manually realigned and is not otherwise inoperable.
    Date of issuance: April 19, 1994.
    Effective date: April 19, 1994.
    Amendment Nos.: 204, 223, and 177.
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
    Date of initial notice in Federal Register: March 31, 1993 (58 FR 
16873).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611.

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

    Date of application for amendments: February 7, 1994 (TS 93-11).
    Brief description of amendments: The amendments replace the wording 
in Surveillance Requirement 4.7.9.i, ``Snubber Service Life Program,'' 
with that from the Westinghouse Electric Corporation Standard TS, 
Revision 4a. In addition, the amendments delete the wording in SR 
4.7.9.c, ``Snubber Visual Inspection Performance and Evaluation,'' that 
is inconsistent with Generic Letter 90-09.
    Date of issuance: April 18, 1994.
    Effective date: April 18, 1994.
    Amendment Nos.: 179--Unit 1 171--Unit 2.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: February 7, 1994.
    The Commission's related evaluation of the amendments are contained 
in a Safety Evaluation dated April 18, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendments: February 7, 1994 (TS 93-19).
    Brief description of amendments: The amendments revise Technical 
Specification 5.3.1 to allow the substitution of filler rods for fuel 
rods in fuel assemblies by incorporating the guidance in Generic Letter 
90-02, Supplement 1.
    Date of issuance: April 18, 1994.
    Effective date: April 18, 1994.
    Amendment Nos.: 180--Unit 1 172--Unit 2.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12367) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated April 18, 1994.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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: September 28, 1992.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 2.2, Limiting Safety System Settings, TS 3.3.1, 
Reactor Protection System Instrumentation, and TS 3.3.2, Isolation 
Actuation Instrumentation by removing the functions associated with the 
main steam line radiation monitors.
    Date of issuance: April 22, 1994.
    Effective date: April 22, 1994.
    Amendment No. 58.
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 6, 1993 (58 FR 
598) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 22, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment: July 28, 1992, as supplemented 
on February 17, 1993.
    Brief description of amendment: The proposed amendment would delete 
Technical Specification (TS) 3/4.9.9, ``Refueling Operations--
Containment Purge and Exhaust Isolation System,'' and its bases, 
because of its redundancy to other TSs that address the operability 
requirements of the containment purge and exhaust isolation system. 
Also, the proposed amendment would revise TS 3/4.3.2, ``Safety System 
Instrumentation--Safety Features Actuation System Instrumentation,'' 
and TS 3.4.9.4, ``Refueling Operations--Containment Penetrations,'' and 
its bases. The effect of this proposed change would be to allow the 
bypass of the safety features actuation system in Mode 6, 
``Refueling,'' by the use of the containment purge and exhaust system 
noble gas monitor in conjunction with manual closure of the containment 
purge and exhaust isolation valves instead of automatic closure.
    Date of issuance: April 15, 1994.
    Effective date: April 15, 1994.
    Amendment No. 186.
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 6, 1993 (58 FR 
599) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 1994.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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: October 8, 1993.
    Brief description of amendments: The amendments revise the 
technical specifications (TS) by deleting tables listing certain 
components from the TS and relocating the lists to plant procedures in 
accordance with the guidance provided in NRC Generic Letter 91-08, 
``Removal of Component Lists from Technical Specifications.''
    Date of issuance: April 22, 1994.
    Effective date: April 22, 1994.
    Amendment Nos.: 181 and 162.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57860) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 22, 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.

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

    Date of application for amendment: July 29, 1993, as supplemented 
by letters dated March 11 and 17, 1994.
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TS) to reflect a new refueling mast. 
Specifically, the amendment adds new values for protective features in 
the TS to reflect the new refueling mast. Values for the old refueling 
mast are retained in the TS.
    Date of issuance: April 29, 1994.
    Effective date: April 29, 1994.
    Amendment No.: 121.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14900) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 29, 1994.
    Public comments on proposed no significant hazards consideration 
comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

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.
    Brief description of amendments: The amendments adding operating 
conditions and limiting conditions for operation for the atmospheric 
steam dump valves, the crossover steam dump system, the turbine stop 
and governor valves, and the various turbine overspeed protection 
features installed at the Point Beach Nuclear Plant. Additionally, the 
amendments revised the surveillance requirements for the auxiliary 
feedwater system, and added explanatory text to the bases for Sections 
15.3.4 and 15.4.8.
    Date of issuance: April 20, 1994.
    Effective date: April 20, 1994.
    Amendment Nos.: 147 and 151.
    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 amendments is 
contained in a Safety Evaluation dated April 20, 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: October 21, 1993, as supplemented by 
letters dated March 14, 1994, and April 18, 1994.
    Brief description of amendment: The amendment revises Technical 
Specification Sections 6.5.1, Plant Safety Review Committee (PSRC) and 
6.8, Procedures and Programs, in order to allow implementation of a 
Qualified Reviewer Program for the review and approval of new 
procedures and procedure changes. Technical Specification 6.5.1.6, PSRC 
Responsibilities, has also been revised in accordance with Generic 
Letter 93-07, ``Modification of Technical Specification Administrative 
Control Requirements for Emergency and Security Plans,'' to delete 
requirements for PSRC review of the Emergency Plan and Security Plan 
and related implementing procedures.
    Date of Issuance: April 28, 1994.
    Effective date: April 28, 1994, to be implemented within 120 days 
of issuance.
    Amendment No.: Amendment No. 73.
    Facility Operating License No. NPF-42. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62159) The March 14, 1994, and April 18, 1994, supplemental letters 
provided additional clarifying information and revised the 
implementation period and did not change the initial no significant 
hazards consideration. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated April 28, 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 4th day of May 1994.

    For the Nuclear Regulatory Commission.
Jack W. Roe,
Director Division of Reactor Projects--III/IV Office of Nuclear Reactor 
Regulation.
[FR Doc. 94-11226 Filed 5-11-94; 8:45 am]
BILLING CODE 7590-01-P