[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'' (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'' (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 MCPRf [minimum critical
power ratio] and MAPLHGRf [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 MCPRf and
MAPLHGRf 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
