[Federal Register Volume 62, Number 136 (Wednesday, July 16, 1997)]
[Notices]
[Pages 38130-38146]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10716]


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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

[[Page 38131]]

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 June 23, 1997, through July 3, 1997. The 
last biweekly notice was published on July 2, 1997 (62 FR 35846).

Notice Of Consideration of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Harards Consideration 
Determination, And Opportunith For A Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By August 15, 1997, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no

[[Page 38132]]

significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: May 30, 1997, identified as CY-97-006
    Description of amendment request: Changes to the Operating License, 
DPR-61, and facility Technical Specifications (TS) that reflect the 
permanently shut down and defueled status of the plant.
    CY-97-006 contains the proposed changes to the license conditions 
in DPR-61 on Fire Protection, Power Level and Fuel Movement; and 
submittal of a new set of TS referred to by the licensee as the 
Defueled TS (DTS). The DTS contain a revised Definitions section, 
removal of the sections on Safety Limits and Limiting Safety System 
Settings, Limiting Conditions for Operation and Surveillance 
Requirements were modified extensively, the Design Features section was 
revised, and the Administrative Controls section was modified to 
reflect all the preceding changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Connecticut Yankee Atomic Power Company (CYAPCO) has reviewed 
the proposed changes to the Operating License and the Technical 
Specifications in accordance with 10 CFR 50.92 and concluded that 
the changes do not involve a significant hazards consideration 
(SHC). The basis for this conclusion is that the three criteria of 
10 CFR 50.92(c) are not compromised. The proposed changes do not 
involve an SHC because the changes would not:1. Involve a 
significant increase in the probability of consequences of an 
accident previously evaluated.
    Because of the present plant configuration, many of the 
postulated accidents previously evaluated (i.e., loss or coolant 
accident, main steam line break, etc.) are no longer possible. The 
accidents previously evaluated that are still applicable to the 
plant are fuel handling accidents and gaseous and liquid radioactive 
releases.
    There is no significant increase in the probability of a fuel 
handling accident since refueling operations have ceased. In fact, 
there is more likely a decrease in probability of a fuel handling 
accident since the need to move/rearrange fuel assemblies is minimal 
until they are removed from the spent fuel pool (i.e., for dry cask 
storage or for transferring to U.S. Department of Energy 
possession).
    The radiological consequences of a gaseous or liquid radioactive 
release are bounded by the fuel handling accident. With the plant 
defueled and permanently shutdown, the demands on the radwaste 
systems is lessened since no new radioisotopes are being generated 
by irradiation or fission. Therefore, there is no increase in the 
probability or consequences of a gaseous or liquid radioactive 
release.
    The changes to the Operating License reflect the permanently 
defueled condition for power level and fuel movement restrictions 
and the fire protection regulation which is applicable for a 
permanently defueled plant.
    With respect to the Service Water System (Specification 3/
4.7.3), Electrical Power Systems (Specification 3/4.8) and spent 
fuel pool makeup, the basis for placing appropriate requirements in 
the Technical Requirements Manual is due to the reduced heat load in 
the spent fuel pool.
    The plant was shutdown on July 22, 1996 and more than 280 days 
have passed since the shutdown, thus the heat load on the spent fuel 
pool cooling system is greatly reduced. Present cooling performance 
data as well as calculations demonstrate that either the plate or 
the shell and tube heat exchanger has more than adequate heat 
removal capacity. In the event of a loss of forced cooling, 
calculations indicate that the spent fuel pool time to boil is 
greater than 40 hours based on an initial pool temperature of 
150 deg.F. The initial pool temperature of 150 deg.F is based on 
Technical Specification 3/4.9.15 which has a pool temperature limit 
of 150 deg.F. Even during boiling, the fuel is adequately cooled. 
Once boiling commences, the operators have in excess of 18 days to 
provide forced cooling and/or makeup before there is inadequate 
shielding provided by the water in the pool. This allows sufficient 
time to provide for alternate forced cooling or makeup to the spent 
fuel pool in the event of a service water system failure. Therefore, 
operability of spent fuel pool cooling does not require service 
water, electrical power, or makeup water to be immediately 
available.
    Should failure to restore operation of the spent fuel pool 
cooling system occur before boiling takes place, cooling of the 
spent fuel can be accomplished by allowing the spent fuel pool to 
boil and adding makeup water at a rate equal to or greater than the 
boil-off rate.
    CYAPCO has in place procedures to establish onsite power in the 
event of a Loss of Normal Power (LNP) and in the event of a loss of 
cooling to the Spent Fuel Pool. For a LNP, power can be made 
available within approximately one hour. If onsite power cannot be 
reestablished, due to equipment failure, at approximately 2 hours 
into the LNP, limited makeup water could be provided by gravity feed 
from a tank (available in approximately 30 minutes) or an unlimited 
supply of water could be provided via the diesel fire pump from the 
Connecticut River (available in approximately 30 minutes). 
Therefore, within approximately 2 1/2 hours of the event start, 
cooling and/or makeup would be reestablished to the spent fuel pool. 
Historically, the longest LNP the HNP has experienced has been less 
than 30 minutes.
    The changes to Technical Specification 3.3.3.8, ``Radioactive 
Gaseous Effluent Monitoring Instrumentation'' and Table 3.3.-10 
delete the trip function from the main stack noble gas activity 
monitor. The changes to Technical Specifications 3.11.2.1, Dose 
Rate, and 3.11.2.3, Dose, delete the requirement to include the 
radioiodine isotopes in the dose calculations. These changes are 
based on the following:
    There is no significant increase in the consequences of a fuel 
handling accident since the accident scenarios assume an assembly 
with significant amounts of radioactive iodine or noble gas. The 
plant was shutdown on July 22, 1996. Except for I-125 (half-life 
=59.5 days), I-129 (half-life = 1.6E7 years), and Kr-85 (half-life 
=10.8 years), the spent fuel inventory of the dose contributing 
radioactive iodine and noble gas isotopes has decayed more than 20 
half-lives since shutdown (i.e., less than 0.0001% of the original 
amount remains). In addition, the definition for ``Dose Equivalent 
I-131'' (<gr-thn-eq>Standard Technical Specifications, Westinghouse 
Plants,'' NUREG-1431) does not include I-125 and I-129 in the dose 
assessment due to their negligible inventory in the spent fuel. 
Except for Kr-85, the other noble gas nuclides that contribute to a 
whole

[[Page 38133]]

body dose have also decayed to a negligible amount. CYAPCO has 
performed fuel handling and cask drop accident dose calculations 
which conclude that doses (i.e., whole body and thyroid) at the 
Exclusion Area Boundary are a small fraction of the 1O CFR 100 dose 
limits and the EPA PAGS. In fact, due to this decreased radioactive 
inventory, there is a significant decrease in the consequences of a 
fuel handling accident.
    Based on the above, the proposed changes to the Operating 
License and the Technical Specifications do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There is no change in how spent fuel is stored or moved in the 
spent fuel pool. Therefore, the postulated fuel handling accidents 
are still bounding and are still considered as credible postulated 
accidents. The bases provided in the CYAPCO analysis of previously 
evaluated accidents in Section 1, above, also applies to the 
possibility of new or different accidents herein.
    Based on the analysis in Section 1, above, the changes to 
Technical Specification related to radioactive iodine and noble gas 
isotopes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Based on these considerations, the proposed changes to the 
Operating License and the Technical Specifications do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    With respect to the Service Water System (Specification 3/
4.7.3), Electrical Power Systems (Specification 3/4.8) and spent 
fuel pool makeup, the basis for placing appropriate requirements in 
the Technical Requirements Manual is due to the reduced heat load in 
the spent fuel pool.
    The Technical Specification basis states that the time to spent 
fuel pool boiling after a loss of forced cooling following a full 
core offload is 7 hours.
    In accordance with the analysis set forth above under No. 1, 
there is no change in how spent fuel is stored or moved in the spent 
fuel pool.
    Based on the above, the proposed changes to the Operating 
License and the Technical Specifications 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270
    NRC Project Director: Marvin M. Mendonca, Acting Director

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam 
Neck Plant, Middlesex County, Connecticut

    Date of amendment request: May 30, 1997, identified as CY-97-024
    Description of amendment request: CY-97-024 provided the proposed 
technical specifications (TS) needed to implement the Certified Fuel 
Handler (CFH) program at the plant. This new position will replace the 
former licensed operator positions. A copy of the CFH Training Program, 
``Nuclear Training Manual NTM-7.083'' was enclosed with the license 
amendment request for NRC review and approval. However, this manual 
will be reviewed separately from the proposed TS changes and when the 
NRC review of the manual is completed a letter of approval will be sent 
to the licensee.
    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:
    Connecticut Yankee Atomic Power Company (CYAPCO) has reviewed 
the proposed changes to the Technical Specifications in accordance 
with 10 CFR 50.92 and concluded that the changes do not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10 CFR 50.92(c) are not 
compromised. The proposed changes do not involve an SHC because the 
changes would not.
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed qualification, staffing and training requirements 
are appropriate for the present plant conditions.
    The plant has permanently ceased operations, the reactor has 
been permanently defueled, and the spent fuel stored in the spent 
fuel pool.
    Because the present plant conditions, many of the postulated 
accidents previously evaluated (i.e., loss-of-coolant accident, main 
steam line break, etc.) are no longer possible. The accidents 
previously evaluated that are still applicable to the plant are fuel 
handling accidents and gaseous and liquid radioactive releases.
    There is no significant increase in the probability of a fuel 
handling accident since refueling operations have ceased. In fact, 
there is more likely a decrease in probability of a fuel handling 
accident since the need to move/rearrange fuel assemblies is minimal 
until they are removed from the spent fuel pool (i.e., for dry cask 
storage or for transferring to U.S. Department of Energy 
possession).
    There is no significant increase in the consequences of a fuel 
handling accident since the accident scenarios assume an assembly 
with significant amounts of radioactive iodine or noble gas. The 
plant was shutdown on July 22, 1996. Except for I-125 (half-
life=59.5 days), I-129 (half-life=1.6E7 years), and Kr-85 (half-
life-10.8 years), the spent fuel inventory of the dose-contributing 
radioactive iodine and noble gas isotopes has decayed more than 20 
half-lives since shutdown (i.e., less than 0.0001% of the original 
amount remains). In addition, the definition for ``Dose Equivalent 
I-131'' (<gr-thn-eq>Standard Technical Specifications, Westinghouse 
Plants,'' NUREG-1431) does not include I-125 and I-129 in the dose 
assessment due to their negligible spent fuel inventory. Except for 
Kr-85, the other noble gas nuclides that contribute to a whole body 
dose have also decayed to a negligible amount. CYAPCO has performed 
fuel handling and cask drop accident dose calculations which 
conclude that doses (i.e., whole body and thyroid) at the Exclusion 
Area Boundary and the Low Population Zone are a small fraction of 
the 10 CFR 100 dose limits. In fact, due to this decreased 
radioactive inventory, there is a significant decrease in the 
consequences of a fuel handling accident.
    The radiological consequences of a gaseous or liquid radioactive 
release are bounded by the fuel handling accident. With the plant 
defueled and permanently shutdown, the demands on the radwaste 
systems are lessened since no new radioisotopes are being generated 
by irradiation. Therefore, there is no increase in the consequences 
of a gaseous or liquid radioactive release.
    Based on the above, the proposed changes to the Technical 
Specifications do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There is no change in how spent fuel is stored or moved in the 
spent fuel pool. Therefore, the postulated fuel handling accidents 
are still bounding and are still considered as credible postulated 
accidents.
    Based on the above, the proposed changes to the Technical 
Specifications do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    There is no change in how spent fuel is stored or moved in the 
spent fuel pool.
    The plant was shutdown on July 22, 1996. Except for I-125 (half-
life=59.5 days), I-129 (half-life=1.6E7 years), and Kr-85 (Half-
life=10.8 years), the spent fuel inventory of the dose-contributing 
radioactive iodine and noble gas isotopes has decayed more than 20 
half-lives since shutdown (i.e., less than 0.0001% of the original 
amount remains). Except for Kr-85, the other noble gas nuclides that 
contribute to a whole body dose have also decayed to a negligible 
amount. CYAPCO has performed fuel handling and cask drop accident 
dose calculations which conclude that doses (i.e, whole body and

[[Page 38134]]

thyroid) at the Exclusion Area Boundary and the Low Population Zone 
are a small fraction of the 10 CFR 100 dose limits.
    Therefore, there is no significant reduction the margin of 
safety. In fact, due to this decreased radioactive iodine inventory, 
there is more likely an increase in the margin of safety.
    Based on the above, the proposed changes to the Technical 
Specifications 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270
    NRC Project Director: Marvin M. Mendonca

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

    Date of amendment request: June 20, 1997 (NRC-97-0037), as 
supplemented by letter dated July 3, 1997
    Description of amendment request: The proposed amendment would 
relocate technical specification surveillance requirement 4.4.1.1.2 for 
the reactor recirculation system motor-generator (MG) set scoop tube 
stop setpoints to the Updated Final Safety Analysis Report. In 
addition, the proposed amendment includes the following changes to the 
surveillance testing methodology: (1) eliminating any licensing basis 
requirement for the electrical stops, and (2) revising the periodicity 
from a calendar basis to a situational basis (i.e., plant conditions 
that would dictate a change in stop positions).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change removes from the Fermi 2 Technical 
Specifications (TS) a Surveillance Requirement (SR 4.4.1.1.2) that 
is an implementation detail and relocates it to the Updated Final 
Safety Analysis Report (UFSAR), where it is more adequately and more 
appropriately controlled in accordance with 10 CFR 50.59. In 
addition, this proposed change revises the test methodology by: (1) 
eliminating the requirement for the electrical stops because they 
are not credited for mitigating any transients or accidents, and (2) 
revising the periodicity from a calendar basis to a situational 
basis to coincide with the beginning of each operating cycle or 
post-maintenance. These changes do not eliminate the necessary 
testing of the MG set mechanical stops. The MG set mechanical stops 
will continue to remain operable because the recirculation pump MG 
set mechanical speed stop settings will continue to be maintained at 
or below the required limits. The MCPR<INF>f</INF> [minimum critical 
power ratio] and MAPLHGR<INF>f</INF> [maximum average planar linear 
heat-generation rate] limits, along with the recirculation pump MG 
set mechanical speed stop settings on which they are based, are 
specified in the Core Operating Limits Report and operation within 
these limits is required by Technical Specifications 3.2.1 and 
3.2.3. The changes described will therefore have no impact on the 
probability or consequences of an accident previously evaluated.
    2. The changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed Technical Specification change does not result in 
any changes to the design (equipment/configuration) or operation of 
the plant and will thus not create a new failure mode or common mode 
failure. The MG set mechanical stops will continue to operate as 
intended and as designed. These changes will therefore not create 
the possibility of a new or different kind of accident, from any 
accident previously evaluated.
    3. The changes do not involve a significant reduction in the 
margin of safety.
    Changes in the methodology and frequency of testing will not 
involve a significant reduction in the margin of safety because the 
testing necessary to ensure the stops are set correctly will 
continue to be performed. Additionally, the MCPR<INF>f</INF> and 
MAPLHGR<INF>f</INF> limits, along with the recirculation pump MG set 
mechanical speed stop setting that they are based on, are specified 
in the Core Operating Limits Report, and operation within these 
limits is still required by Technical Specifications 3.2.1 and 
3.2.3. Therefore, the margin of safety as defined in the bases of 
any Technical Specification is not reduced by relocating the 
surveillance requirement from the TS to the UFSAR. In addition to 
the above, relocation of the TS is consistent with the BWR Improved 
Standard Technical Specification, NUREG-1433, Rev. 1.
    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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: John N. Hannon

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

    Date of amendment request: April 24, 1997
    Description of amendment request: The requested amendment revises 
the inservice inspection requirements associated with steam generator 
tube sleeves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does Not Involve a Significant Increase in the Probability or 
Consequences of an Accident Previously Evaluated.
    This change implements a more stringent surveillance requirement 
than currently exists. It incorporates a requirement to inspect a 
minimum of 20% of each type of installed sleeve in each steam 
generator. The 20% inspection criterion is conservative with respect 
to the existing requirement of a 3% initial inspection of all steam 
generator tubes. Additionally, since the process for inspections has 
not changed, the probability or consequences of accidents previously 
analyzed are not increased as a result of inspection activities. The 
proposed changes have no impact on any previously analyzed accident 
in the safety analysis report.
    The administrative changes made to update the technical 
specifications or to correct inconsistencies introduced in previous 
amendments do not affect reactor operations or accidental analyses 
and have no radiological consequences.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Does Not Create the Possibility of a New or Different Kind of 
Accident from any Previously Evaluated.
    The changes made to increase the initial sample of sleeved tubes 
inspected during a surveillance, to update the technical 
specifications and to correct inconsistencies introduced in previous 
amendments are administrative and do not change the design, 
configuration or method of operation of the plant nor does it 
introduce any new possibility for an accident.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. Does Not Involve a Significant Reduction in the Margin of 
Safety.
    As previously discussed, this change implements a more stringent 
surveillance requirement than currently exists. The existing 
technical specifications require an initial inspection of 3% of the 
tubes in each steam generator while the proposed change

[[Page 38135]]

requires inspection of a minimum of 20% of each type of installed 
sleeve. The 20% inspection criterion is conservative with respect to 
the existing technical specification. Existing technical 
specification operability and surveillance requirements are not 
reduced by the proposed change, thus no margins of safety are 
reduced.
    The other administrative changes do not reduce technical 
specification operability and surveillance requirements, and 
therefore, do not reduce any margin of safety.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: June 26, 1997
    Description of amendment request: The proposed amendment will 
modify Technical Specification (TS) Tables 3.7-1 and 3.7-2. Table 3.7-1 
will be revised to change the Main Steam Safety Valves (MSSVs) orifice 
size from 26 square inches to 28.27 square inches and to relocate the 
orifice size from the TS Table to the TS Bases. The change to correct 
the orifice size is an editorial change to make the TS consistent with 
plant design. Table 3.7-2 will be revised by deleting the provision 
that allows continued plant operation with three MSSVs inoperable. The 
proposed amendment will also revise TS Bases 3/4.7.1.1 to remove the 
equation used for determining the reduced maximum allowable linear 
power level-high reactor trip settings of TS Table 3.7-2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No
    In response to the ABB/CE report pursuant to 10CFR21 regarding 
the omission of Main Steam Safety Valve (MSSV) piping pressure loss 
in safety analyses, the proposed change will eliminate the ability 
to operate the plant in accordance with Technical Specification 
3.7.1.1 Action a with three MSSVs inoperable. The Bases to this 
Technical Specification will also be revised to state that the 
acceptability for operation at lower power levels with one or two 
MSSVs inoperable will be determined from results obtained from a 
loss of condenser vacuum accident analysis under these conditions. 
Deleting the allowance for continued operation with three MSSVs 
inoperable does not increase the probability of an accident. The 
consequences of an accident will not be increased by these changes. 
These changes are more restrictive and ensure that the MSSVs 
maintain their safety function of removing adequate heat from the 
steam generator in order to maintain peak steam generator pressure 
and peak pressurizer pressure well below their respective acceptance 
criteria during normal operation and all anticipated operational 
occurrences.
    Changing the MSSVs orifice size listed in TS to their actual 
size and the orifice size utilized in the safety analysis, and 
relocating the MSSVs orifice size to the Technical Specification 
Bases does not affect the probability or consequences of an 
accident. The correct orifice size was used in the safety analysis 
and it is not subject to change unless a station modification is 
performed which will require a 10CFR50.59 evaluation and revision of 
the safety analysis. The MSSVs orifice size can be adequately 
controlled in the TS Bases which will also require a 10CFR50.59 to 
be changed.
    Therefore, operation of Waterford 3 in accordance with this 
proposed change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No
    The proposed change will eliminate the ability to operate the 
plant in compliance with Technical Specification 3.7.1.1 Action a 
with three MSSVs inoperable. The Bases for this Technical
    Specification will also be revised to state that the 
acceptability for operation at lower power levels with one or two 
MSSVs inoperable will be determined from results obtained from a 
loss of condenser vacuum accident under these conditions. The 
proposed change also revises the MSSVs orifice size to reflect the 
actual orifice size and the orifice size utilized in the safety 
analysis, and relocates the orifice size from Technical 
Specifications to the Technical Specification Bases. The proposed 
change does not involve any new equipment, components, or 
modifications and does not create any new system interactions or 
connections. Therefore, operation of Waterford 3 in accordance with 
this proposed change will not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No
    The proposed change will ensure that all appropriate acceptance 
criteria for the MSSVs are met during normal operation and all 
anticipated operational occurrences. The Technical Specification 
Bases 3/4.7.1.1 will be updated to state that the acceptance 
criteria for operation in accordance with Technical Specification 
3.7.1.1 Action a will be determined from the results of the limiting 
loss of condenser vacuum accident. This change ensures that the 
transient and dynamic effects which occur during accident scenarios 
are fully evaluated. These changes also ensure that the MSSVs will 
maintain peak steam generator pressure and peak pressurizer pressure 
well below their respective acceptance criteria during normal 
operation, design basis accidents and anticipated operational 
occurrences.
    The proposed change also revises the MSSVs orifice size to 
reflect the actual orifice size and the orifice size utilized in the 
safety analysis, and relocates the orifice size from Technical 
Specifications to the Technical Specification Bases. This change 
corrects an editorial error in the Technical Specifications and 
relocates unsurveilled design details from the Technical 
Specifications. Adequate control of the orifice size will remain 
adequate because any changes to the orifice size or the orifice size 
listed in the Bases will require a station modification and a TS 
Bases change. Station Modifications and TS Bases changes requires 
evaluation in accordance with 10CFR50.59.
    Therefore, operation of Waterford 3 in accordance with this 
proposed change will not involve a significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: James W. Clifford, Acting

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

    Date of amendment request: May 5, 1997
    Description of amendment request: The proposed amendment to 
Technical Specifications 3.9.1.2 and 3.9.13 and

[[Page 38136]]

their Bases would allow crediting soluble boron for maintaining k-
effective at less than or equal to 0.95 within the spent fuel pool 
(SFP) rack matrix following a seismic event of a magnitude greater than 
or equal to an operating basis earthquake (OBE).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed change in accordance with 
10CFR50.92 and has concluded that the change does not involve a 
Significant Hazards Consideration (SHC). The bases for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
satisfied. The proposed change does not involve [an] SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    There is one Spent Fuel Pool accident condition discussed in 
Chapter 15 of the FSAR [Final Safety Analysis Report]. The FSAR 
discusses a fuel handling accident which drops a fuel assembly onto 
the fuel racks during fuel movement. Degradation of the Boraflex 
panels in a post-seismic condition will have no effect on the 
probability of a fuel assembly drop onto the stored fuel, or the 
fuel racks. Changing the way Boraflex responds to a seismic event 
will have no impact on the probability of a seismic event. A 
misplaced fuel assembly can be postulated in the MP3 [Millstone Unit 
3] fuel pool as a result of either equipment malfunction or operator 
error. Degradation of the Boraflex panels will have no effect on the 
probability of a fuel misplacement event. Therefore, the degradation 
of Boraflex in a post-seismic condition does not involve an increase 
in the probability of an accident previously evaluated.
    A fuel handling accident could cause a radioactive release of 
fission gases, resulting in dose consequences. This radioactive 
release of fission gases is due to the failure of a certain number 
of fuel pins which are postulated to fail during the fuel handling 
accident. The number of fuel pins which are postulated to fail in 
this event is not affected by the degradation of the Boraflex panels 
in a post-seismic condition. There are no criticality issues with 
this fuel handling accident for the reasons described next. Should a 
fuel handling accident occur prior to a seismic event, the existing 
fuel handling accident/misloading criticality analysis is still 
valid, such that 800 ppm [parts per million] of soluble boron is 
sufficient to ensure that K-effective of the SFP is maintained at 
less than 0.95. Although overly conservative, should a fuel handling 
accident occur during or after a seismic event, even with no 
Boraflex credit, the proposed 1750 ppm of soluble boron is 
sufficient to ensure that K-effective of the SFP is maintained at 
less than 0.95. Therefore, this proposed change does not involve an 
increase in the probability or consequences of an accident 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The change in the way Boraflex in conjunction with the addition 
of 1750 ppm boron responds to a seismic event does not create a new 
accident. The use of soluble boron in the Spent Fuel Pool is safe 
during and immediately following a seismic event, because the 
balance of the equipment in the fuel building not connected to the 
fuel pool which could cause a dilution (firewater, hot water 
heating, and demineralized water, CCP [component cooling-plant]) are 
seismic or mounted in such a fashion as to not direct unborated 
water into the fuel pool should a line rupture. Non borated water 
sources that are connected to the SFP will be isolated following a 
seismic event of greater than or equal to [an] OBE to prevent 
dilution. Therefore there is no possibility of [an] SFP boron 
dilution accident coincident with a seismic event, and credit for 
soluble boron is acceptable to meet the K-effective limit of 0.95 
for the SFP. The crediting of soluble boron in the Spent Fuel Pool 
to control K-effective following a seismic event does not create a 
new accident as boron dilution of the pool can be prevented by 
closing and administratively controlling the opening of dilution 
paths to the pool and initiating routine sampling requirements on 
SFP boron. At present the crediting of soluble boron following a 
fuel misplacement event is allowed for the Millstone 3 Spent Fuel 
Pool. Analysis has shown that a seismic event of greater than an OBE 
level earthquake can be more limiting than a fuel misplacement 
event. As such the minimum boron requirement in the fuel pool will 
be increased from 800 ppm to 1750 ppm. As such, no new accident has 
been created because the crediting of boron following a malfunction/
accident has always been an allowed event.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety, as defined by MP3 Technical 
Specifications, is to ensure that the K-effective of the MP3 SFP is 
maintained less than or equal to 0.95 at all times. There is no 
reduction in the margin of safety as the result of the degradation 
of Boraflex following a greater than an OBE seismic event, because 
soluble boron can be used to compensate for the loss of Boraflex. A 
value of 1750 ppm of soluble boron in the SFP at all times ensures 
that K-effective of the MP3 SFP is maintained less than or equal to 
0.95 at all times, including this new malfunction of degraded 
Boraflex following a greater than an OBE seismic event.
    Eliminating the credit for the negative reactivity effect of 
Boraflex panels in conjunction with the addition of 1750 ppm boron 
will have no effect on the probability of a seismic event. As the 
probability of a seismic event has not changed there is no increase 
in the probability of an accident or malfunction due to a seismic 
event. Following a seismic event operators are presently required to 
make inspections of the plant to determine post seismic event plant 
conditions. As a result of this change, inspections will be required 
to post seismic event evaluations to review the status of the Spent 
Fuel Pool and isolate potential dilution paths. These action are 
consistent with present guidance in the seismic response procedure 
and do not create an undue burden on the operator. To compensate for 
the potential
    loss of Boraflex after a seismic event, the SFP is now required 
to be borated at all times to 1750 ppm to maintain the proper post 
seismic [K-effective] condition. As such there is no mitigation 
equipment that has to operate in the Spent Fuel Pool following a 
seismic event.
    Although the Boraflex in the fuel racks is assumed to fail in a 
greater than an OBE seismic event, the presence of soluble boron in 
the fuel pool water will compensate for the loss of Boraflex. 
Surveillance requirements on SFP boron will ensure that there will 
be boron present in the SFP and ensure that the SFP is not diluted 
below the minimum required boron concentration during normal 
operation.
    As the presence of SFP soluble boron during and after a seismic 
event maintains [K-effective] less than 0.95 there is no effect on 
the consequences of any malfunctions evaluated. As there are no new 
accidents created and there are no changes in the probability or 
consequences of previously analyzed accidents there is no effect on 
the consequences of any accident. There is no reduction in the 
margin of safety as the result of the degradation of Boraflex 
following a greater than an OBE seismic event, because soluble boron 
can be used to compensate for the loss of Boraflex to maintain K-
effective less than 0.95.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    In conclusion, bases on the information provided, it is 
determined that the proposed change does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

[[Page 38137]]

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

    Date of amendment request: March 26, 1997
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to incorporate additional 
restrictions on the operation of the main steam safety valves (MSSVs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Omaha Public Power District (OPPD) proposes to revise the 
Fort Calhoun Station (FCS) Unit No. 1 Technical Specifications (TS) 
2.1.6, ``Pressurizer and Main Steam Safety Valves,'' to incorporate 
additional restrictions on the Main Steam Safety Valves (MSSVs) as a 
result of recent engineering analyses.
    FCS has two Steam Generators (SG), each with one 2 1/2-inch MSSV 
and four 6-inch MSSVs. The purpose of the MSSVs is to limit the 
secondary system pressure to less than or equal to 110% of the 
design pressure of 1000 lbs. per square inch absolute (psia) when 
passing 100% of design steam flow.
    The pressure drops in the main steam lines were calculated. The 
total losses (line losses and valve losses) of 30.5 psid (2 1/2 inch 
valves) and 33.5 psid (6 inch valves) were compared to the valve 
blowdown which is adjusted/checked each refueling outage as part of 
the required surveillance test. The pressure losses are less than 
the 39 psid and 40 psid blowdown for the 2 1/2 inch and 6 inch valve 
with the lowest setpoint (respectively). Therefore, the 
recommendation from the Part 21 to review blowdown settings to 
preclude valve chatter was conducted and there is no concern at FCS. 
A review of existing calculations for line losses in the primary 
system was conducted and was determined to be 39 psid for the inlets 
to the primary safety valves.
    Analyses were then conducted to determine the impact of the 
total line losses on previously analyzed accidents documented in the 
Updated Safety Analysis Report (USAR). The scope of the analyses was 
to evaluate the pressure drops in the piping run for both the 
primary and MSSVs to determine the impact on the peak primary and 
secondary system pressures. The applicable transient for peak 
primary system pressure is the Loss of Load, and for maximum 
secondary system pressure is the Loss of Feedwater. All analyses 
were performed using the NRC-approved CESEC-III transient analysis 
methodology and computer code.
    The assumptions of the analyses were that the plant is operating 
at 1535.6 MWt, (100% power + 2% uncertainty + reactor coolant pump 
heat), the MSSVs lifted at +3% of their nominal setpoints, the 
primary safety valve setpoints were adjusted to account for line 
losses and lifting at +1% of their setpoints, and the pressure 
losses in the main steam line to the SG were added to obtain the 
maximum secondary system pressure within the SG. Additional cases 
were evaluated with a +6% primary safety valve drift since this 
possibility is described in the Bases to TS 2.1.6.
    The results from these analyses confirm that the effective 
increase in MSSV set pressure caused by the piping pressure losses 
leading to the primary safeties and MSSVs is below the 1100 psia 
design limit for the secondary system, and below the 2750 psia 
design limit for the primary system. This is predicated on the fact 
that only one (1) MSSV may be inoperable per SG.
    Failure of a MSSV is not an initiator of any previously analyzed 
accident, and therefore the proposed changes do not increase the 
probability of an accident previously analyzed. The proposed change 
to revise TS 2.1.6 to allow only one MSSV per SG to be inoperable 
has been shown, utilizing NRC approved methodology, to
    limit the design pressure to values below the design limits. An 
administrative change to revise the TS setpoint value for both the 
primary safety valves and MSSVs from pounds absolute to pounds gauge 
is proposed to be consistent with the nameplate values of the valves 
and has no effect on any analyses. Therefore the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There will be no physical alterations to the plant 
configuration, changes in operating modes, setpoints, or testing 
methods. The additional restrictions being incorporated into the TS 
on MSSV operation will ensure that the design basis limits of 110% 
of design pressure will be met for the primary and secondary systems 
for analyzed accidents when considering inlet pipe pressure drops. 
The possibility of valve chatter being caused by the additional 
pressure losses identified in the Main Steam lines and MSSVs was 
reviewed and is not a concern. This is due to the valve blowdown 
(the difference between a valve's opening pressure and closing 
pressure) being greater than the pressure losses. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change results in a peak primary pressure of 2649 
psia (with 1% primary safety valve drift as allowed by TS 2.1.6) and 
peak secondary pressure of 1081 psia for the loss of load event 
compared to 2632 psia and 1075 psia documented in USAR Section 14.9. 
The proposed change results in a peak primary pressure of 2562 psia 
and peak secondary pressure of 1090 psia for the loss of feedwater 
event compared to 2487 psia and 1052 psia documented in USAR Section 
14.10. The analyses confirm that the primary and secondary systems 
will continue to be below their respective design limits of 2750 
psia and 1100 psia. 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William H. Bateman

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: May 31, 1996
    Description of amendment request: This change deletes Technical 
Specification 4.7.2.d.2, ``Control Room Emergency Outside Air Supply 
System Surveillance Requirement,'' related to the detection of 
chlorine.
    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 proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Review of the various design basis accidents identified in 
Chapter 15 of the Susquehanna SES [Steam Electric Station] Final 
Safety Analyses Report (FSAR) concluded that none of these accidents 
are affected by deletion of the chlorine detection surveillance 
requirement from Technical Specifications. With the elimination of 
bulk quantities of gaseous chlorine from use at Susquehanna SES the 
probability of control room inhabitability due to a gaseous chlorine 
release has actually decreased. Therefore, this proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change involves only the deletion of the chlorine 
detection system Technical Specifications based upon a plant

[[Page 38138]]

modification to remove gaseous chlorine as a biocide from 
Susquehanna SES and replace it with an oxidizing biocide with non-
gaseous/non-volatile properties. The release of chlorine from an 
off-site source is bounded by Reg. [Regulatory] Guide 1.95 in that 
manual isolation capability for the control room ventilation system 
is acceptable. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. This change does not involve a significant reduction in a 
margin of safety.
    The proposed change would not alter the margins of safety 
provided in the existing FSAR analysis (Sections 2.2.3.1.3 and 6.4) 
for chlorine release events since the basis for the existing margin 
of safety, which are the Reg. Guide 1.95 requirements, are not 
altered by the change. As stated above, since gaseous chlorine is no 
longer used for open cooling water treatment at Susquehanna SES and 
since the biocide currently used does not pose the same personnel 
inhalation threat as gaseous chlorine, safety margin has actually 
increased. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama

    Date of amendments request: June 13, 1997
    Description of amendments request: The proposed amendments would 
change Technical Specification (TS) 3/4.9.13, ``Storage Pool 
Ventilation (Fuel Movement),'' by adding a note in the TSs to 
specifically indicate that the normal emergency power source may be 
inoperable in MODE 5 or 6 provided that the requirements of TS 3.8.1.2 
are satisfied and extend the TS 3.9.13 completion time allowed for 
returning one out-of-service penetration room filtration system from 48 
hours to 7 days. The Bases will also be modified to provide additional 
detail concerning these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) The proposed changes do not significantly increase the 
probability or consequences of an accident previously evaluated in 
the FSAR [Final Safety Analysis Report]. The proposed changes have 
no impact on the probability of an accident. The storage pool 
ventilation system will continue to ensure that radioactive material 
released as a result of a fuel handling accident in the spent fuel 
pool room will be filtered through the HEPA [high efficiency 
particulate air] filters and charcoal absorbers prior to discharge 
to the atmosphere. There is no change in the FNP [Farley Nuclear 
Plant] design basis as a result of this change and, as a result, 
does not involve a significant increase in the consequences of an 
accident previously evaluated.
    (2) The proposed changes to the TSs do not increase the 
possibility of a new or different kind of accident than any accident 
already evaluated in the FSAR. No new limiting single failure or 
accident scenario has been created or identified due to the proposed 
changes. Safety-related systems will continue to perform as 
designed. The proposed changes do 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 the margin of safety. As a result of these proposed changes, the 
penetration room filtration system, when it is aligned to the spent 
fuel pool room, will continue to require verification of 
operability. There is no impact in the accident analyses. These 
proposed changes are technically consistent with the requirements of 
NUREG-1431, Revision 1 which has already received the requisite 
review and approval of the NRC staff. Thus the proposed changes do 
not involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch 
Nuclear Plant, Unit 1, Appling County, Georgia

    Date of amendment request: April 29, 1997, as supplemented by 
letter dated May 28, 1997
    Description of amendment request: The amendment would revise the 
Unit 1 reactor vessel pressure and temperature limits to reflect data 
collected from the material sample recovered during the March 1996 Unit 
1 outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Pressure and Temperature (P/T) limits for the reactor pressure 
vessel are established to the requirements of 10 CFR [Part] 50, 
Appendix G to ensure brittle fracture of the vessel does not occur.
    This revision changes the P/T curves in the Unit 1 Technical 
Specifications to reflect the material capsule surveillance results 
from the sample removed during the [s]pring outage of 1996.
    The RPV [reactor pressure vessel] surveillance capsule contained 
flux wires for neutron flux monitoring and Charpy V notch impact and 
tensile test specimens. The irradiated material properties were 
compared to available unirradiated properties to determine the 
effect of irradiation on material toughness for the base and weld 
materials through Charpy testing. Irradiated tensile testing results 
are compared with unirradiated data to determine the effect of 
irradiation on the stress-strain relationship of the materials.
    The P/T curves are modified to reflect the results of the above 
examination. These curves and their operating limits were evaluated 
using the approved methodologies of 10 CFR [Part] 50 Appendix G and 
ASME [American Society of Mechanical Engineers] Code Appendix G. The 
new curves therefore represent the latest information available on 
the state of the reactor vessel materials. The P/T curves are 
generated for reactor vessel protection against brittle fracture, 
they do not affect the recirculation piping. Accordingly, the 
probability of occurrence of a design basis Loss of Coolant Accident 
(LOCA) is not increased. Likewise, no other previously evaluated 
accident and transients, as defined in Chapter 14 of the Final 
Safety Analysis Report (FSAR) are affected by this proposed change 
to the Unit 1 P/T curves. Additionally, this proposed revision does 
not affect the design, operation, or maintenance of any safety 
related system designed for the mitigation or prevention of 
previously analyzed events.

[[Page 38139]]

    Since no previously evaluated accidents or transients are being 
affected by this change, their probability of occurrence is not 
increased and their consequences are not made worse.
    2. Do the proposed changes create the possibility of a new or 
different type of accident from any previously evaluated?
    Implementing the proposed P/T curves into the Unit 1 Technical 
Specifications does not alter the design or operation of any system 
or piece of equipment designed for the prevention or mitigation of 
accidents and transients. As a result, no new operating modes are 
introduced from which a new type accident becomes possible. Existing 
systems will continue to be operated per present design basis 
assumptions.
    The proposed P/T limits were generated from the evaluation of 
the material capsule removed during the [s]pring Unit 1 outage of 
1996. As a result, these limits include the latest available 
information on the reactor vessel materials. Furthermore, they will 
continue to be monitored per the requirements of the Technical 
Specifications and 10 CFR [Part] 50 Appendices G and H. For the 
above reasons, the changes do not create the possibility of a new 
type of accident.
    3. Do the proposed changes involve a significant reduction in 
the margin of safety?
    The purpose of the P/T limits is to avoid a brittle fracture of 
the reactor vessel. As such, material capsules are removed 
periodically to determine the effects of neutron irradiation on 
reactor vessel materials. This change to the Unit 1 P/T curves is 
proposed to incorporate the evaluation results of the latest capsule 
removed during the [s]pring Unit 1 outage of 1996. Accordingly, 
these curves represent the latest information available on the 
reactor vessel materials. Also, the curves were generated using the 
approved methodologies of 10 CFR [Part] 50 Appendix G.
    The pressure test curve (Figure 3.4.9-1) is also being revised 
to reflect exposure dependencies. These curves were generated for 
exposures of 16, 18, 20, 24, 28, and 32 EFPY [effective full-power 
year]. As previously described, each of these curves were generated 
using approved methodologies and all reflect the results of this 
latest material capsule report.
    The proposed change does not affect the evaluation of any FSAR 
Unit 1 Chapter 14 transient and accident. Furthermore, the proposed 
change does not affect the operation of systems or equipment 
important to safety.
    The Limiting Condition for Operation of Specification 3.4.9 will 
not change. Also, no Technical Specification surveillances or 
surveillance frequencies are revised as a result of this Technical 
Specification submittal, besides the fact that the P/T surveillances 
will now refer to the revised curves. Procedures regarding the 
monitoring of the P/T limits during reactor startup, cooldown, and 
leakage testing will not change as a result of this proposed 
Technical Specification change with respect to frequency of the 
surveillance or the methods used to perform the surveillances. Thus, 
the P/T limits will continue to be surveilled as before per the same 
procedures and the same frequencies.
    No other Technical Specifications are affected by the proposed 
revision. The margin of safety to any Technical Specifications 
safety limit therefore is not reduced.
    For the above reasons the new curves do not represent 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: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, 
Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, 
Georgia

    Date of amendment request: May 30, 1997
    Description of amendment request: The proposed amendments would 
revise power sources to valves associated with low pressure coolant 
injection (LPCI) mode of residual heat removal (RHR) system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The LPCI valves operate to establish and maintain adequate core 
cooling following a LOCA [loss-of-coolant accident]. The proposed 
changes do not alter the function or mode of operation of the LPCI 
valves. Therefore, the probability of the LOCA accident is not 
increased. An analysis which considered the consequences of the 
various transients and accidents with the proposed change in power 
supply of the LPCI valves indicates the consequences are not 
increased.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
The change in power supply to the LPCI valves maintains the original 
design criteria that a power supply independent of the remaining RHR 
subsystem be utilized for single-failure criteria. The function of 
the LPCI valves and any other existing equipment is not altered. 
Operation of the valves in the proposed configuration was analyzed, 
and no new failure modes exist. An analysis of the impact on the 
operation and design of other systems and components indicates no 
new failure modes are introduced. Therefore, these changes do not 
contribute to a new or different type of accident.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety. The change in power supply to the LPCI 
valves was evaluated relative to RHR and electrical distribution 
system function during normal and accident conditions. The proposed 
change does not alter the performance of any system safety 
functions. The results of the SAFER-GESTR LOCA analysis reconfirm 
the large margins existing in fuel peak cladding temperature under 
the proposed configuration. Therefore, there is no 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: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: June 13, 1997
    Description of amendment request: The proposed amendments would 
revise the Technical Specification Limiting Condition for Operation 
3.4.10 Pressurizer Safety Valves. Specifically, the change would reduce 
the nominal set pressure by 1 percent to 2460 pounds per square inch 
gauge (psig) and increase the tolerance to plus or minus 2 percent.
    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:

[[Page 38140]]

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The increase in the PSV [pressurizer safety valve] tolerance 
from [plus or minus] 1% with a setpoint of 2485 psig to [plus or 
minus] 2% and reduction in the nominal setpoint from 2485 psig to 
2460 psig has the net effect of reducing the minimum lift setting 
allowed by the TS [technical specifications] from 2460 psig to 2410 
psig. The effects of this change have been evaluated for its impact 
on the assumed frequency of safety valve challenges and failures to 
reclose, and the proposed change was found to have a negligible 
impact. In other words, reducing the minimum lift setting does not 
significantly increase the probability of an inadvertent actuation 
of a safety valve during normal operation. Reducing the minimum lift 
setting does increase the potential that the PSVs may open during an 
event, but this change has been evaluated and does not adversely 
impact the consequences of any accident previously evaluated. No 
change to any equipment response or accident mitigation scenario has 
resulted, and there are no additional challenges to fission product 
barrier integrity. Therefore, the proposed change does not 
significantly increase the probability or consequences of any 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The increase in the PSV tolerance from [plus or minus] 1% with a 
setpoint of 2485 psig to [plus or minus] 2% and reduction in the 
nominal setpoint from 2485 psig to 2460 psig does not create the 
possibility of a new or different kind of accident than any accident 
previously evaluated. No new accident scenarios, failure mechanisms, 
or limiting single failures are introduced as a result of this 
proposed change. The proposed revision to Technical Specification 
3.4.10 does not challenge the performance or integrity of any 
safety-related systems. Therefore, the possibility of a new or 
different kind of accident is not created.
    3. Does the proposed change involve a significant reduction in a 
margin of safety.
    The proposed change to Technical Specification 3.4.10 does not 
involve a significant reduction in a margin of safety. The 
modification will have no affect on the availability, operability or 
performance of the safety-related systems and components. The 
increased PSV set pressure tolerance has been reviewed with respect 
to the accident analysis assumptions and requirements and evaluated 
or analyzed, as required. These evaluations and analyses determined 
that all applicable acceptance criteria continue to be met, thus the 
proposed increase in the PSV set pressure tolerance will not result 
in a significant reduction in the margin of safety associated with 
the acceptance criteria for the accident analyses.
    The Bases of the Technical Specifications rely in part on the 
ability of the regulatory criteria being satisfied assuming the 
limiting conditions for operation for various systems. Conformance 
to the regulatory criteria for operation with the increased PSV set 
pressure tolerance is demonstrated, and the regulatory limits are 
not exceeded. Hence, the margin of safety as defined in the Bases 
for the Technical Specifications is not significantly reduced.
    Therefore, there is no significant reduction in any 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: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: May 16, 1997 (TXX-97119)
    Brief description of amendments: The licensee has proposed revised 
core safety limit curves and Overtemperature N-16 reactor trip 
setpoints based on analyses of the core configuration for CPSES Unit 2, 
Cycle 4. These changes apply equally to CPSES Units 1 and 2 licenses 
since the Technical Specifications are combined.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    A. Revision to the Unit 2 Core Safety Limits
    Analyses of reactor core safety limits are required as part of 
reload calculations for each cycle. TU Electric has performed the 
analyses of the Unit 2, Cycle 4 core configuration to determine the 
reactor core safety limits. The methodologies and safety analysis 
values result in new operating curves which, in general, permit 
plant operation over a similar range of acceptable conditions. This 
change means that if a transient were to occur with the plant 
operating at the limits of the new curve, a different temperature 
and power level might be attained
    than if the plant were operating within the bounds of the old 
curves. However, since the new curves were developed using NRC 
approved methodologies which are wholly consistent with and do not 
represent a change in the Technical Specification BASES for safety 
limits, all applicable postulated transients will continue to be 
properly mitigated. As a result, there will be no significant 
increase in the consequences, as determined by accident analyses, of 
any accident previously evaluated.
    B. Revision to Unit 2 Overtemperature N-16 Reactor Trip 
Setpoints
    As a result of changes discussed, the Overtemperature reactor 
trip setpoint has been recalculated. These trip setpoints help 
ensure that the core safety limits are protected and that all 
applicable limits of the safety analysis are met.
    Based on the calculations performed, no significant changes to 
the safety analysis values for Overtemperature reactor trip setpoint 
were required. The f(delta I) trip reset function was revised due to 
more top-skewed axial power distributions predicted for this cycle. 
The analyses performed show that, using the TU Electric 
methodologies, all applicable limits of the safety analysis are met. 
This setpoint provides a trip function which allows the mitigation 
of postulated accidents and has no impact on accident initiation. 
Therefore, the changes in safety analysis values do not involve an 
increase in the probability of an accident and, based on satisfying 
all applicable safety analysis limits, there is no significant 
increase in the consequences of any accident previously evaluated.
    In addition, sufficient operating margin has been maintained in 
the overtemperature setpoint such that the risk of turbine runbacks 
or reactor trips due to upper plenum flow anomalies or other 
operational transients will be minimized, thus reducing potential 
challenges to the plant safety systems.
    SUMMARY
    The changes in the amendment request applies NRC approved 
methodologies to changes in safety analysis values, new core safety 
limits and new N-16 setpoint and parameter values to assure that all 
applicable safety analysis limits have been met. The potential for 
an operational transient to occur has not been affected and there 
has been no significant impact on the consequences of any accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed changes involve the calculation of new reactor core 
safety limits and overtemperature reactor trip setpoint resets. As 
such, the changes play an important role in the analysis of 
postulated accidents but none of the changes effect plant hardware 
or the operation of plant systems in a way that could initiate an 
accident. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?

[[Page 38141]]

    In reviewing and approving the methods used for safety analyses 
and calculations, the NRC has approved the safety analysis limits 
which establish the margin of safety to be maintained. While the 
actual impact on safety is discussed in response to question 1, the 
impact on margin of safety is discussed below:
    A. Revision to the Unit 2 Reactor Core Safety Limits
    The TU Electric reload analysis methods have been used to 
determine new reactor core safety limits. All applicable safety 
analysis limits have been met. The methods used are wholly 
consistent with Technical Specification BASES 2.1 which is the bases 
for the safety limits. In particular, the curves assure that for 
Unit 2, Cycle 4, the calculated DNBR is no less than the safety 
analysis limit and the average enthalpy at the vessel exit is less 
than the enthalpy of saturated liquid. The acceptance criteria 
remains valid and continues to be satisfied; therefore, no change in 
a margin of safety occurs.
    B. Revision to Unit 2 Overtemperature N-16 Reactor Trip 
Setpoints
    Because the reactor core safety limits for CPSES Unit 2, Cycle 4 
are recalculated, the Reactor Trip System instrumentation setpoint 
values for the Overtemperature N-16 reactor trip setpoint which 
protect the reactor core safety limits must also be recalculated. 
The Overtemperature N-16 reactor trip setpoint helps prevent the 
core and Reactor Coolant System from exceeding their safety limits 
during normal operation and design basis anticipated operational 
occurrences. However, it was shown in these calculations that the 
current Unit 2 overtemperature reactor trip setpoint (presented in 
the current Technical Specifications and excluding the f(delta I) 
trip reset function) remains valid. The most relevant design basis 
analysis in Chapter 15 of the CPSES Final Safety Analysis Report 
(FSAR) which is affected by the Overtemperature reactor trip 
setpoint is the Uncontrolled Rod Cluster Control Assembly Bank 
Withdrawal at Power (FSAR Section 15.4.2). This event has been 
analyzed with the new safety analysis value for the Overtemperature 
reactor trip setpoint to demonstrate compliance with event specific 
acceptance criteria. Because all event acceptance criteria are 
satisfied, there is no degradation in a margin of safety.
    The nominal Reactor Trip System instrumentation setpoints values 
for the Overtemperature N-16 reactor trip setpoint (Technical 
Specification Table 2.2-1) are determined based on a statistical 
combination of all of the uncertainties in the channels to arrive at 
a total uncertainty. The total uncertainty plus additional margin is 
applied in a conservative direction to the safety analysis trip 
setpoint value to arrive at the nominal and allowable values 
presented in Technical Specification Table 2.2-1. Meeting the 
requirements of Technical Specification Table 2.2-1 assures that the 
Overtemperature reactor trip setpoint assumed in the safety analyses 
remains valid. The CPSES Unit 2, Cycle 4 Overtemperature reactor 
trip setpoint is not significantly different from the previous 
cycle, and thus provides operational flexibility to withstand mild 
transients without initiating automatic protective actions. Although 
the value of the f(delta I) trip reset function setpoint is 
different, the Reactor Trip System instrumentation setpoint values 
for the Overtemperature N-16 reactor trip setpoint are consistent 
with the safety analysis assumptions which have been analytically 
demonstrated to be adequate to meet the applicable event acceptance 
criteria. Thus, there is no reduction in a margin of safety.
    Using the NRC approved TU Electric methods, the reactor core 
safety limits are determined such that all applicable limits of the 
safety analyses are met. Because the applicable event acceptance 
criteria continue to be met, there is no significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: James W. Clifford, Acting

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

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

Commonwealth Edison Company, Docket No. STN 50-455, Byron Station, 
Unit No. 2, Ogle County, Illinois Docket No. STN 50-457, Braidwood 
Station, Unit No. 2, Will County, Illinois

    Date of amendment request: May 24, 1997
    Description of amendment request: The amendments revise the 
technical specifications related to venting of the emergency core 
cooling system pumps and associated piping. The application originally 
included Byron, Unit 1. However, on May 31, 1997, ComEd supplemented 
the application to request an emergency license amendment for Byron, 
Unit 1. Amendment No. 90 was issued on June 1, 1997.
    Date of publication of individual notice in Federal Register: June 
10, 1997 (62 FR 31633)
    Expiration date of individual notice: July 10, 1997
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481

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

    Date of application for amendment: May 16, 1997
    Brief description of amendment: The proposed amendment would make 
an administrative change to add a supervisory position to the list of 
personnel who may be required to hold a senior reactor operator 
license. Date of publication of individual notice in Federal Register: 
June 4, 1997 (62 FR 30625)
    Expiration date of individual notice: July 7, 1997
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126

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.

[[Page 38142]]

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

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

    Date of application for amendment: January 20, 1997, with the 
proposed no significant hazards consideration submitted by letter dated 
January 30, 1997, as supplemented February 27, April 11, May 14, and 
June 20 (2 letters), 1997
    Brief description of amendment: The amendment authorizes Boston 
Edison Company (BECo) to change the UHS administrative limit from 
68 deg.F to 75  deg.F, and change the Updated Final Safety Analysis 
Report (UFSAR) to reflect the use of containment pressure to compensate 
for the deficiency in NPSH following a design basis accident and 
increase the accident analysis design UHS temperature from 65 deg.F to 
75 deg.F. As part of this amendment, BECo has proposed to submit a 
Technical Specification amendment for the UHS temperature by the first 
quarter of 1998. In addition, within 180 days of issuance of this 
amendment, BECo has committed to complete the containment analysis 
using the ANS 5.1-1979 Decay Heat Curve with a 2-sigma uncertainty 
added. The staff considers BECo's commitments acceptable and has 
conditioned the amendment accordingly.
    Date of issuance: July 3, 1997
    Effective date: July 3, 1997
    Amendment No.: 173
    Facility Operating License No. DPR-35: Amendment revised the 
Updated Final Safety Analysis Report.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8792) The February 27, April 11, May 14, and June 20 (2 letters), 
1997, letters provided clarifying information that did not change the 
initial proposed no significant hazards consideration determination as 
submitted by letter dated January 30, 1997. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 3, 1997. No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360

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: March 14, 1997, as supplemented 
May 16, and June 17, 1997
    Brief description of amendment: The amendment approves changes to 
the Final Safety Analysis Report to reflect new analysis of the 
radiological consequences of dropping a fuel cask.
    Date of issuance: June 26, 1997
    Effective date: June 26, 1997
    Amendment No. 73
    Facility Operating License No. NPF-63. Amendment revises the Final 
Safety Analysis Report.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17226). The May 16, and June 17, 1997 supplemental information did not 
change the original no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated June 26, 1997. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

    Date of amendment request: April 11, 1997
    Brief description of amendment: The amendment changes the Waterford 
steam Electric Station, Unit 3, Technical Specifications (TSs) by 
revising TS 3.6.2.2 and Surveillance Requirement 4.6.2.2 for the 
Containment Cooling System. Also, a Surveillance Requirement is added 
to verify that valves actuate on a Safety Injection Actuation Signal. 
To support this addition, Technical Specification Bases 3/4.3.6.2.2 is 
also included.
    Date of issuance: July 3, 1997
    Effective date: July 3, 1997, to be implemented within 60 days.
    Amendment No.: 131
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1997 (62 FR 
19626) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 3, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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

    Date of application for amendment: April 17, 1997
    Brief description of amendment: The amendment modifies Technical 
Specification 3.7.14 by clarifying the actions to be taken when an area 
temperature exceeds its temperature limit.
    Date of issuance: June 24, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 141
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: (62 FR 27798 May 21, 
1997) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 24, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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

    Date of application for amendment: April 15, 1997
    Brief description of amendment: The amendment makes changes to 
Technical Specification (TS) Sections 4.3.3.6 and 4.6.4.1, which 
require that the hydrogen monitors be periodically tested. 
Specifically, the changes increase the testing interval of the 
monitor's hydrogen sensor, correct inconsistencies

[[Page 38143]]

between the TS surveillances, and make changes to the Bases of the 
surveillances.
    Date of issuance: June 24, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 142
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27797) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 24, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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

    Date of application for amendments: April 11, 1997
    Brief description of amendments: These amendments revise Technical 
Specification (TS) 3/4.6.2.3, ``Containment Cooling System,'' and its 
associated Bases section to ensure that the TSs properly test the 
containment fan cooling units' post-accident mode of operation.
    Date of issuance: June 24, 1997
    Effective date: Both units, as of the date of issuance, to be 
implemented within 60 days.
    Amendment Nos. 197 and 180
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27799) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 24, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079

Tennessee Valley Authority, Docket Nos. 50-327 and 50328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: March 13, 1997, as supplemented 
on June 26, 1997 (TS 97-01)
    Brief description of amendments: The amendments change the 
Technical Specifications by raising the allowable U-235 enrichment, as 
specified in Section 5.6.1.2, of fuel stored in the new fuel pit 
storage racks from 4.5 to 5.0 weight percent.
    Date of issuance: July 1, 1997
    Effective date: July 1, 1997
    Amendment Nos.: 225 and 216
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27802). The June 26, 1997 supplement provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in an environmental assessment dated June 16, 
1997, and a Safety Evaluation dated July 1, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendment: August 27, 1993, as supplemented 
by letters dated November 9, 1993, April 26, 1996, and September 25, 
1996
    Brief description of amendment: The amendment revises the Technical 
Specifications to incorporate the revised 10 CFR Part 20, Standards for 
Protection Against Radiation.
    Date of issuance: June 19, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 151
    Facility Operating License No. DPR-28. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
507) The November 9, 1993, April 26, 1996, and September 25, 1996, 
submittals did not change the initial proposed no significant hazards 
consideration. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 19, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301

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: June 4, 1996 (TSCR 188 and 
189), as supplemented August 5, September 26, October 21, November 13, 
November 20, and December 2, 1996, and January 16, March 20, and April 
2, 1997
    Brief description of amendments: These amendments revise Technical 
Specifications (TS) 15.1, ``Definitions;'' TS 15.2.1, ``Safety Limit, 
Reactor Core;'' TS 15.2.3, ``Limiting Safety System Settings, 
Protective Instrumentation;'' TS 15.3.1, ``Reactor Coolant System,'' 
Section C, ``Maximum Coolant Activity,'' and Section G, ``Operational 
Limitations;'' TS 15.3.4, ``Steam and Power Conversion System;'' TS 
15.3.5, ``Instrumentation System;'' TS 15.4.1, ``Operational Safety 
Review;'' TS 15.5.3, ``Design Features-Reactor;'' and TS 15.6.9, 
``Plant Reporting Requirements'' to reflect parameters associated with 
new steam generators in Unit 2 and changes in analyses that affect both 
Units 1 and 2.
    Date of issuance: July 1, 1997
    Effective date: July 1, 1997. The TS shall be implemented within 45 
days from the date of issuance and the Final Safety Analysis Report 
changes shall be implemented by June 30, 1998. Implementation of these 
amendments includes incorporation of accident analyses submitted in 
support of this amendment into the Final Safety Analysis Report in 
sufficient detail to support future evaluations performed in accordance 
with 10 CFR 50.59 and as described in the licensee's applications dated 
June 4, 1996, as supplemented on August 5, September 26, October 21, 
November 13, November 20, and December 2, 1996, and January 16, March 
20, and April 2, 1997, and evaluated in the staff's safety evaluation 
dated July 1, 1997.
    Amendment Nos.: 173, 177
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34903 and 61 FR 34904) and April 9, 1997 (62 FR 17243) The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated July 1, 1997. No significant hazards consideration 
comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241

[[Page 38144]]

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

    Date of amendment request: March 21, 1997, as supplemented by 
letter dated April 15, 1997
    Brief description of amendment: The amendment revises Technical 
Specification 6.8.5.b to provide an exception to the examination 
requirements of Regulatory Guide 1.14, Revision 1, ``Reactor Coolant 
Pump Flywheel Integrity'' and delays the inspection of the ``D'' 
reactor coolant pump flywheel to the Fall 1997 refueling outage. A 
typographical error in TS 6.8.5.c is corrected.
    Date of issuance: June 24, 1997
    Effective date: June 24, 1997, to be implemented within 30 days of 
issuance.
    Amendment No.: 106
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27803) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 24, 1997. 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

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration and 
opportunity for a hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    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 for the 
amendment 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.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective 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 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By August 15, 1997, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be

[[Page 38145]]

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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    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 20555-001, 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, by the above 
date. A copy of the petition should also be sent to the Office of the 
General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-001, 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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: December 11, 1996, as 
supplemented March 27, 1997, April 17, 1997, and June 17, 1997
    Brief description of amendment: The amendment revises Technical 
Specifications to allow extended rod position indicator deviation 
limits, on-line calibration of the rod position indication and to 
clarify the operability requirements during calibration.
    Date of issuance: June 27, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 194
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. The NRC published a public 
notice of the proposed amendment, issued a proposed finding of no 
significant hazards consideration and requested that any comments on 
the proposed no significant hazards consideration be provided to the 
staff by the close of business on June 25, 1997. The notice was 
published in the Peekskill Evening Star on June 20-25, 1997.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the State of New York and 
final no significant hazards consideration determination are contained 
in a Safety Evaluation dated June 27, 1997.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610

North Atlantic Energy Service Corporation, Dockets Nos. 50-443, 
Seabrook Station, Unit 1, Seabrook, Massachusetts

    Date of amendment request: June 19, 1997
    Brief description of amendment: The amendment revised Technical 
Specification 6.8.1.6.b. to include a reference to the NRC-approved 
Westinghouse Topical Report WCAP-12610-P-A, ``VANTAGE+ Fuel Assembly 
Reference Core Report,'' dated April 1995.
    Date of issuance: June 24, 1997
    Effective date: As of the date of issuance, and to be implemented 
before transition into Operational Mode 2 during startup from Refueling 
Outage 5.
    Amendment No.: 52
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. The Commission's related 
evaluation of the amendment, finding of emergency circumstances, 
consultation with the States of New Hampshire and Massachusetts, and 
final no significant hazards considerations determination are contained 
in the safety evaluation dated June 24, 1997.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, New Hampshire 03833
    Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270 
Acting
    NRC Project Director: Patrick D. Milano

North Atlantic Energy Service Corporation, Dockets Nos. 50-443, 
Seabrook Station, Unit 1, Seabrook, Massachusetts

    Date of amendment request: May 29, 1997
    Brief description of amendment: The amendment modifies Technical 
Specification 5.3.1 by replacing the current term ``zircaloy'' with 
terminology that explicitly identifies the NRC-approved Westinghouse 
fuel assembly design in use at the Seabrook Station consisting of 
assemblies with either ZIRLO or Zircaloy-4 fuel cladding material.
    Date of issuance: June 24, 1997
    Effective date: As of the date of issuance, and to be implemented 
before transition into Operational Mode 2 during startup from Refueling 
Outage 5.

[[Page 38146]]

    Amendment No.: 53
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes. The NRC published a public 
notice of the proposed amendment, issued a proposed finding of no 
significant hazards consideration, and requested that any comments on 
the proposed no significant hazards consideration be provided to the 
staff by the close of business on June 10, 1997. The notice was 
published in Foster's Daily Democrat and in the Portsmouth Herald on 
June 4, 1997. Public comments were received, and they have been 
addressed in the staff's safety evaluation.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the States of New Hampshire 
and Massachusetts, and final no significant hazards determination are 
contained in a safety evaluation dated June 24, 1997.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, New Hampshire 03833
    Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270 
Acting
    NRC Project Director: Patrick D. Milano

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

    Date of application for amendment: March 22, 1997, as supplemented 
by letters dated April 2, April 3, April 9, April 15, and May 14, 1997. 
Additional information was also received by telefax on May 19, 1997.
    Brief description of amendment: The amendment revises Surveillance 
Requirement (SR) 3.3.1.1.15, Reactor Protection System (RPS) Response 
Time functions 3 and 4 and SR 3.3.6.1.7, Primary Containment Isolation 
System Response Time, functions 1.a, 1.b, and 1.c, adding a note to 
indicate that the sensor is excluded from response time testing when 
verifying that the response time is within limits. The amendment also 
revises SR 3.3.5.1.7, Emergency Core Cooling System (ECCS) Response 
Time by relocating the requirements to SR 3.5.1.8, ECCS Operating, and 
adding a note to SR 3.5.1.8 to indicate that no actuation 
instrumentation response time measurement is required. Additionally, SR 
3.5.1.8 requires that the SR be met in MODES 1, 2, and 3, whereas the 
previous SR 3.3.5.1.7 was required to be met in MODES 1, 2, 3, 4, and 
5.
    Date of Issuance: June 11, 1997
    Effective date: June 11, 1997
    Amendment No.: 150
    Facility Operating License No. NPF-21. The amendment revised the 
Technical Specifications. Press release issued requesting comments as 
to proposed no significant hazards consideration: Yes. April 11, 1997. 
Tri-City Herald (Washington). Comments received: No. The Commission's 
related evaluation of the amendments, finding of exigent circumstances, 
consultation with the State of Washington and final determination of no 
significant hazards consideration are contained in a Safety Evaluation 
dated June 11, 1997.
    Attorney for licensee: Perry D. Robinson, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    Local Public Document Room location:  Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    NRC Project Director: William H. Bateman
    Dated at Rockville, Maryland, this 9th day of July 1997.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 97-18513 Filed 7-15-97; 8:45 am]
BILLING CODE 7590-01-F